INTERN AtlONAkS YM POSIU M
RADIATION CREjyU (PLUTONR^^tfVEARS)
PREPRINTS VOLUME
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EARC BOMBAY ; FEBRUARY 4'-7, 1991
! ()R(;ANISEI) BY BOARD OFIRESEARCH )N NUGLBAR|SCIENCES DEPARTMENTJOF ATOMIC ENERGY IN *,A oo^oA —^ ®J °®
INTERNATIONAL SYMPOSIUM ON RADIOCHEMISTRY & RADIATION CHEMISTRY (PLUTONIUM - 50 YEARS)
PREPRINTS VOLUME
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BARC, BOMBAY FEBRUARY 4-7, 1991.
ORGANISED BY BOARD OF RESEARCH IN NUCLEAR SCIENCES DEPARTMENT OF ATOMIC ENERGY GOVERNMENT OF INDIA FOREWORD
The Department of Atomic Energy (DAE) has been arranging symposia in the field of Nuclear Chemistry & Radiation ChemiBtry since 1964. After 1980 these symposia have become an annual feature, the present being 10 in succession. The importance of the current symposium is that it is being held as INTERNATIONAL SYMPOSIUM TO COMMEMORATE 50 years of the discovery of PLUTONIUM by Prof. CJ.T. Seaborg and jothers in the year 1941 and memorable events thereafter that lead to the discovery of the entire group of trans-plutonium elements. Befitting *o the occasion, the symposium is being held at Bhabha Atomic Research Centre during Feb. 4-7, 1991. Major time of this symposium will be devoted to the topics of research and development in the field of actinides. Fourteen leading scientists from France, Germany, Japan, USA & USSR will deliver talks on the current international status of Science and Technology of Actinides and their impact on environment. Twelve senior Indian scientists will review the current actinide research in the country. Besides these, 76 research papers will be presented. Four invited talks and 109 research papers will be presented in the field of Nuclear Chemistry & Radiation Chemistry. As it has been a practice in the previous years, th© invited talks and research papers are published in Preprints volume which is made available to all the participants on the opening day. He hope this will enable the participants to involve them in indepth discussion during the delioerations of the symposium. The Symposium Committeo is grateful to Prof.Virendra Singh, Chairman, Board of Research in Nuclear Sciences (BRNS) and other members of the BRNS for their financial support. In organizational matters, I am greatly indebted to Dr.R.M.Iyer, Chairman, Symposium Organising Committee and Director, Chemical & Isotope Group, Chri P.R.Roy, Chairman, International Advisory Committee and Director, Materials Group, Shri A.N.Prasad, Director, Reprocessing Group and all other members of the International Advisory Committee for their guidance in the organization of this symposium. Our sincere thanks are due to Dr.P.E.Natarajan, Chairman, Technical Coniinittee, Dr. H.C.Jain, Chairman, Local Organising Committee, Shri.A.V.Jadhav, Secretary, Symposium Organizing Committee and all members of these committees for their untiring efforts in the successful organization of this symposium. I have also received profound help from a large number of my colleagues in Fuel Chemistry Division and other Divisions of BARC. I express ipy sincere thanks to them. The symposium committee is grateful to all the invited speakers, authors of all research papers for their active participation & for timely submission of manuscript and Shri.M.R.Balakrishnan, Head, Library & Information Services and his colleagues for bringing out this Preprint volume in time. I have a special word of thanks to our friends from abroad for accopting our invitation and travelling long distances to be amongst us and giving us the opportunity to share their rich experiences and thoughts with us.
(D.Ek-Sood) Convener International Advisory Committee
P.R.Roy - Chairman ( BARC, India) S.N.Bhattacharyya (SINP, India) G.R£hoppln (Florida, USA) K.C.Dash (Utkal Univ., India) F.David (Orsay, France) R.Guillaumont (Orsay, France) E.P.Horwitz (ANL, USA) C.Keller (Karlsruhe, FRG) V'.Krishnan (Inst.ot Sc. Bangalore )
R.A.Mashelkar (NCL. India) C.K.Mathews (IGCARr India) H.Matzke (Karlsruhe, FRG) S.P.Mishra (BHU, India) B.F.Mysoedov (Moscow, USSR) K.Naito (Nagoya, Japan) R.H.Pembles (BNFL, UK) A.N.Prasad (BARC. India) A.R.Rsddy (Def.Lab., Jodhpur) H.Roepenack (Alkem, GmbH, FRG) S.B.Soman (HERB, India) Z.R.Threl (Inst. of Sc., Bombay) M ,S .Wadia (Poona Univ., India) M.Yamawaki (Tokyo, Japan)
Symposium ©rganising Committee
R.M.Iyer - Chairman ^.D.Sood - Convener fi.V.Jadhav - Secretary M.R.Balakrishnan K.R.Balasubramanian LS.Bhatt S.Gangadharan C.Ganguly Keshav Chander S.B.Manohar R.Narayanan P.R.Natarajan D.S.C.Purushottam N .Ramamoorthy A. Ramanujam P.R.U. Rao R.K.Singh P.K.Mattel Technical Committee P.R.Natarajart - Chairman S.K.Aggarbial A.U.Jadhav Keshav Chandor V,V.S. Man/ J.N.Mathur R.Parthasarathij R.Prosad K.L.N.Rao R.D.Saini B.SJbmar G.l/enkattfswarem
Isocal 6rganising Committee
H.C.Jnin - Chairman fi.R.Joshi - Reception & Transport Keshav Chander - Registration K. Raghuraman - Accommodation Ziley Singh - Catering
MEMBERS T.R.Bangia S.P.Heisilkar R.I/.Kamat N.B.Khedekar H.R.Mhatre S.K.MukherjBS Q.Ramaswami R.K.Rastogi A.V.R.Reddy M.K.Saxena I/,HMaidya CS.Yadav TABLE OF CONTENTS
No. TITLS OF THE PAPER NAME OF THE AUTHOR IT - INVITED LECTORES
IT-01 HISTORICAL ACHIEVEMENTS OF PLUTONIUM Westrun E.F-Jr. ISOLATION,'DRY CHEMISTRY', AND THERMODYNAMICS AT CHICAGO MET.LAB. IT-02 STUDIES ON URANIUN AND PLUTONIUM FUELS Gurin Y. AT CADARACHE 1T-G3 PLUTONIUM TECHNOLOGY IN GERMANY : Stoll «. HISTORY AND PRESENT STATUS IT-B'l SOLID STATE CHEMISTRY OF Naito K. URANIUM PLUTONIUM MIXED OXIDE IT-05 RESEARCH ON CHEMICAL PROPERTIES OF Zvara I. TRANS-ACTINIDE ELEMENTS IT-06 STUDIES ON PLUTONIUM AND NEPTUNIUM Beauvy M. , CHEMISTRY AT CADARACHE Larroque J. n-»7 ttyasoedov B-F- ELEMENTS IN SOVIET UNION IT-08 RECENT ADVANCES IN THE THERMODYNAMICS David F. OF TRIVALENT ACTINIDES IT-09 THERMODYNAMICS OF ACTINIDES Sood D.D. IT-18 NUCLEAR PROPERTIES OF PLUTONIUM Satya Prakash IT-11 SOME ASPECTS OF SOLUTION CHEMISTRY Patil S.K. OF PLUTONIUM If-12 STABILITY OF PENTAVALENT PLUTONIUM Capdevila H., Vitorge P. IT-13 PLUTONIUM FUEL FABRICATION AT BARC Ganguly C IT-14 THE INPILE BEHAVIOUR OF PLUTONIUM FUELS Mathews C.K- IT-15 CHEMICAL BEHAVIOUR OF FISSION PRODUCT Yamawaki M., Konashi K. IODINE IN IRRADIATED OXJDE FUEL PINS FOR NUCLEAR REACTORS IT-16 ACTINiDE EXTRACTION CHEMISTRY lilTH Nusikas C.,Conda»inc« C., AMIDE TYPE OF EXTRACTANTS Cuillerdier C.Nigend L. U-17 SPEUT FUtL RtPRfcCeS/SiHS -.ft PfctSPECTWE IT-18 PLUTONIUM CHEMISTRY AND FAST REACTOR Vasudeva Rat> P.R., FUEL REPROCESSING Sri ni vaean T-G. IT-19 REPROCESSING OF HIGH CONTENT Pu FUELS Balasubra«ar NC- NUCLEAR CHEMISTRY NC-01 EFFECT OF MASS ASYMMEJRY ON ANGULAR Datta T., Dange S.P., Naitc H., DISTRIBUTION IN THE 2:>8U ia40MeV,F) Satya Prafcash SYSTEM NC-02 HALF-LIVES OF PLUTONIUM ISOTOPES EY Aggarwal S.K., Parab A.R., MASS SPECTROMETRY AND ALPHA Chitambar S.A., Jain H.C. SPECTROMETRY NC-03 POSITRON ANNIHILATION STUDIES IN Datta T., Pujari P.K. , URANIUM OXIDES Chaddha A.K., Satya Prakash NC -04 INITIAL RETENTION, ISOCHRONAL Dedgaonkar V.G., Apte R.6., ANNEALING AND THERMAL ANALYSIS - A Bhagwat D.A. CORRELATION NC-05 PREPARATION OF A MULTITRACER SOLUTION Ambe F. , Ambe S. , Iwaatoto M. , FROM GOLD FOIL BOMBARDED BY 95 MeV/ Ohfcubo Y., Chen S.Y., Garg A.N. NUCLEON 48Ar IONS IN RING CYCLOTRON NC-06 CHARGE DISTRIBUTION IN ALPHA PARTICLE Rattan S.S., Ramaswami A., INDUCED FISSION OF 209Bi Singh R.J., Satya Prakash NC-07 HELIUM ION INDUCED FISSION EXCITATION Iyer R.H. , Shari«a R.C. , FUNCTION OF ERBIUM Kalsi P.C., Pandey A.K. NC-08 ISOMERIC CROSS SECTION RATIOS IN 12, Tomar B.S., Goswani A., Dak S.K. INDUCED REACTIONS ON 89Y Reddy A.V.R.,Manohar S.B., Satya Prakash NC-09 THE LOW-LYING EXCITED STATES OF 197Hg Chakravarty N., Rattan S.S., (64.2H) AND THE ISOMERIC TRANSITION Singh R.J., Ramaswami A., AND ELECTRON CAPTURE DECAY OF197mHg<23 ,8H) Satya Prahash RC - RADIATION CHEMISTRY RC-01 SOLVATION OF EXCESS ELECTRON IN POLAR Bandyopadhyay T. LIQUID : AN OVERVIEW RC-02 THE EPR PARADOX, THE HYDRATED Gopinathan C. tSOLVATED) ELECTRON, AND THE REACTIONS OF THE HYDRATED ELECTRON RC-03 POLARITY EFFECT ON SOLVATED ELECTRON Kapoor S.K., Gopinathan C. REACTIONS RC-04 DIMER ANION FORMATION IN THE PULSE Nai k D.B., Hoorthy P.N. RADIOLYSIS OF 4-PYRIDINOL IN AQUEOUS SOLUTIONS RC-05 REACTION RATE CONSTANTS OF e~ AND Mahal H.S, Nanohar Lai OH. RADICALS WITH ALKYLBROMIDES AH - APPLICATIONS OF RADIQISOTOPES & RADIAT.TOHS AR-Bi EXTRACTION OF SILVERU) WITH AMBERLITE Chatterjee A. Basu S. LA-2 AR-U2 THE INFLUENCE OF DILUENTS ON SYNERGISM Raghupathy S., Sudersanan M. IN THE EXTRACTION OF EUROPIUM AR-03 EXTRACTION & SEPARATION OF Zr, Nb t Hi Mishra P.K., Chakravortty V. , BY ALIQUAT 336 AND ITS MIXTURES WITH Dash K.C. NEUTRAL DONORS FROM AO.HC1 I THIOCYANATE IV AR-04 RPEC SEPARATION OF ZIRCONIUM, NIOBIUM Das N.R. , Lahi ri S. AND HAFNIUM WITH HDEHP AR-05 RAPID EXTRACTION AND RADIOCHEMICAL Jadhav S.D. , Amiani A.M. , SEPARATION OF Cr ( III ) AND Eu(III) WITH Turel Z.R. ALIZARIN INTO DIFFERENT ORGANIC SOLVENTS AR-06 SOLVENT EXTRACTION AND RADIOCHEMICAL Bhatia D.S., Mariam S., SEPARATION OF Mo(VI) AND W(VI) WITH Turel Z.R. " ORGANIC DYES INTO NITROBENZENE AR-B7 SOLVENT EXTRACTION OF Tl(III) AND Kapadia J. , Patil V.B., SnUI) WITH 1,2,3-BENZO-TRIAZOLE INTO Turel Z.R. ORGANIC SOLVENTS AR-08 EXTRACTION MECHANISM OF Tc(VII> IN Sastry E.A.P.S., Sujatha S., TBP-HDEHP-MONOBASIC ACID SYSTEM Mishra G.K., Singh R.N. AR-89 SYNERGISM IN THE EXTRACTION OF Zr(IV) Sujatha S., Mi shra G.K , BY DIFFERENT SOLVENTS FROM VARIOUS Singh R.N. MINERAL ACID SOLUTIONS AR-1B LIQUID-LIQUID EXTRACTION OF Mo AR-12 REMOVAL OF Cr(VI) FROM ABUEOUS SOLU- Bhutani M.M.; Ramesh Kursari., TIONS BY ZrO2 : A KINETIC BASED STUDY Mitra A.K. AR-13 ELECTROLYTE-DIFFUSION OF Co(N03>2 IN Patil S.F., Rajurkar N.S., AGAR SEL MEDIUM : CONCENTRATION Borhade A.V. DEPENDENCE, OBSTRUCTION EFFECT AND ACTIVATION ENERGY AR-14 MULTIELEMENTAL NONDESTRUCTIVE NEUTRON Shanbhag B.S., Turel Z.R., ACTIVATION ANALYSIS OF THE CANCEROUS Turel K.E. TISSUES OF THE HUMAN BRAIN EMPLOYNG 252Cf NEUTRON SOURCE AR-15 STUDIES ON THE HETEROGENITY OF MICRO- Nurty G.S., Padmanabhan P.K. RETICULAR CARBOXYLATE ION EXCHANGERS-I1 AR-16 CARRIER FREE SEPARATION OF 99MD FROM Das S.K., Nair A.G.C., FISSION PRODUCTS Deshmukh S.M., Satya Prakash 4R-17 STUDIES ON THE DECONTAMINATION OF Vaidya N.D., Unni P.R., FISSION PRODUCED 9Mo FROM 131I USING Mathakar H.R., Subrananian H. CHELEX 100 CHELATING ION EXCHANGE RESIN AR-18 SEPARATION OF CARRIER FREE 99mTc FROM Bhattacharyya D.K., Dutta N.C., 99Mo OVER CERIA COLUMN Das N.R. AR-19 DEVELOPMENT OF A COLUMN GENERATOR FOR Kothari K.K., Pillai M.R.A. 99lnTc USING L0W SPECIFIC ACTIVITY 99Mo FOR RESEARCH IN TECHNETIUM CHEMISTRV AR-20 RAglOLABELING OF TESTOSTERONE WITH Samuel Grace, Venkatesh Meera 12jI FOR USE IN RIA 125 AR-2 I LABELING OF ESTRADIOL WITH I AT Korde Aruna , Vent:atesh Meera, DIFFERENT POSITIONS AND THEIR Ramji Lai USABILITY IN RADIOIMMUNOASSAYS AR-22 LABELING OF INSULIN WITH 57Co FOR Samuel Grace, Pillai K.R.A. RADIOIMMUNOASSAY AR-23 STUDIES ON PREPARATION AND ASSAY OF Ramanurthy T.V., Jayakuaar V., CARBON-14 LABELLED OXALIC ACID Harish Chander AR-24 HOFMANN ELIMINATION OF p-NITROPHENYL Raaamurthy T.V., Aurthur Fry ETHYL-1-C-14-TRIMETHYLAMM0NIUM BROMIDE-A CARBON-14 ISOTOPE EFFECT STUDY AR-25 ION-EXCHANGE SEPARATION OVER THORIUM Sarkar B., Basu S. TUNGSTATE COLUHN AR-26 IMMOBILIZATION OF BARIUM, ANTIMONY Bhattacharyya D.K., Dutta N.C., dHD CADMIUM FROM ABUEDUS SOLUTION BY Das N-R- ZIRCONIUM OXIDE + AR-27 ADSORPTION OF ZrO2 IONS ON Ti0o Mishra G.K., Sujatha S. , POWDER Singh R.N. AR-28 SOME ASPECTS OF IR STUDY ON FIXATION Bhutani M.M. , Reddy P.N. , OF CHRDMATE IONS ON HYDROUS ANTIMONY Mitra A.K., Ramesh Kuaari TRIO SIDE AR-29 CONCENTRATION OF CHROMATE IONS ON V2O5 Bhutani M.M. , Mitra A.K. , - A RADIGINDICATOR SORPTION STUDY Ranesh Kumari AR-30 DESORPTION OF I- IONS CHEMISDRBED ON MOLYBDENUM Daniels E.A., Reddy P.G. AR-31 IOD1NE-IODATE ISOTOPIC EXCHANGE REACT- ION STUDIED BY RA0IOTRACER TECHNIQUE Ram K.D., Tripathi R. AR-32 SORPTIQN OF t'ANGAN'ESE AND CHROMIUM ON INDIAN SOILS Khan I.A., Satya Brat AR-53 FLOW RATE MEASUREMENT IN A CRYSTAI.Ll ZER DRAFT TUBE BY MEANS OF A NEUTRALLY Pant H.J. BUOYANT SEALED RADIOACTIVE FLOW FOLLOWER AP-34 RADIOTRACER STUDY OF SOME n- MOLECULAR Mishra S.P., Upadhyaya V. COMPLEXES AR-35 HOMOGENEOUS ISOTOPIC EXCHANGE BETWEEN Gangaiah T., Ramadevi P., NICKEL*It) AND BIS(RESACETOPMENONE Naidu G.R.K. PHENVLHYDRAZONE) NICKEL(II) COMPLEX AR-3£ A BINARY COCKTAIL FOR THE DETERMINA- Subba Rao R.V., Rafi Ahmed TION OF THORIUM AND NATURAL URANIUM A.G., Singh N.S.B., USING LIQUID SCINTILLATION COUNTING Balasubramanian G.R. AR-37 EXTRACTION OF IRON FROM LUBRICATING OIL Chacharkar M.P., Tak E.B. AR-38 RADIOCHEMICAL DETERMINATION OF SELECTED Krishnamoorthy K.R., Iyer R.K. TRACE ELEMENTS IN BIOLOGICAL MATERIALS AR-39 -SUBSTOICHIOMETRIC ISOTOPE DILUTION Satpute P.D. Garg A.N. ANALYSIS OF MERCURY USING POTASSIUM n-BUTYL XANTHATE AR-AB SIMULTANEOUS DETERMINATION OF Sb AND Lanjeuar R.B., Meginwar R.G., Se BY RADIOCHEMICAL NEUTRON ACTIVATION Chut(;e N.L. , Garg A.N. ANALYSIS AR-41 RAD1OCHEMICAL NEUTRON ACTIVATION Weginwar R.G., Garg A.N. ANALYSIS OF Fe, Zn, Co, Sb, Se AND P IN CANCEROUS BREAST TISSUE AR-42 DETERMINATION OF TRACE AMOUNTS OF Rajesh N., Subramanian M.S. INDIUM BY RADIGCHEMICAL DISPLACEMENT AR-43 MJLTIELEMENTAL NEUTRON ACTIVATION Pawar M.B., Arabulkar M.N., ANALYSIS OF WATER SAMPLES AND Garg A.N. ENVIRONMENTAL STANDARDS AR-44 INVESTIGATION ON THE EFFECT OF Das A., Changdar S.N. SOLVENTS ON THE DIFFUSION OF IONS BY RADIOACTIVE TRACER TECHNIQUE AR-45 RADIOMETRIC PROSPECTING IN THE APATITE Kshirasagar T.V.S.R., HTNES OF SITARAMAPURAM, VISAKHAPATNAM Ranakrishna M.1., Nagaaai1esKara DISTRICT, ANDHRA PRADESH, INDIA Rao B. , Naid'i R. T. AR-46 HULTlftODE ACTIVATION ANALYSIS OF Das N.R., Basu D., NIOBIUM IN GEOLOGICAL MATERIALS Bhattacharyya S.N. VI AR-47 RARE EARTH POTENTIAL OF THE Kshirasagar T.V.S.R. , CARBONATITES OF ViSAKHAPATNAM Nagamalleswara Rao 8., DISTRICT, ANDHRfi PRADESH, INDIA AND Parthasarathy R. , THEIR GEOLOGICAL SIGNIFICANCE Pawaskar O.B. , Ha P. AR-48 IN VIVO MEASUREMENT OF TOTAL BODY Mitra S., Plank L.D. , CHLORINE BY PROMPT GAMMA NEUTRON Knight R.S. , Hi J1 G.L. ACTIVATION ANALYSIS AR-49 INSTRUMENTAL ACTIVATION ANALYSIS OF Rajurkar N.S., Kanade K.G., MILLETS BY CALIFORNIUM-252 NEUTRON SOURCE AR-50 USE OF MAGNETISED CELLULOSE IN RADIO- Thorat R.B., Jyotsna N. IMMUNOASSAY FOR TRI IODCTHYRONINE(T3) AR-51 PREPARATION AND CHARACTERIZATION OF Rathi rtaswamy A. RADIOLABELLED MYOGLOBIN NT - NOCLEAR TECHNIQUES AHD IHSTBOMENTATION NT-01 IRON-BEARING MINERALS IN SOME INDIAN Ram L.C., Tripathi P.S.M., COALS CHARACTERISED BY MOSSBAUER Mishra. S.P. SPECTROSCOPY NT-02 TRACK LENGTHS OF 268Pb IN DIELECTRIC Ghosh S., Raju J., Dniv SOLIDS NT-03 PERFORMANCE STUDY OF MODIFIED GAS Rafi A.A.G.,Singh N.S.B., FLOW PROPORTIONAL DETECTOR Subbarao R.V, Ravisankar Af Ba]asubrasanian G.R. SSC - SOLOTION AND SOLID STATE CHEMISTRY OF ACTINIDES SSC-01 EXTRACTION OF U(VI),Pu(IV),Am1111>, Mathur J.N. , Murali M.S. , ?i»(in),ZriIV),Ru(IIl) AND Pd '. 11) Natarajan P.R., Badheka L.P., FROM NITRIC ACID MEDIUM BY MIXTURE Banerji A. OF OCTYL(PHENYL)-N,N-DIISOBUTYL CARBAMOYL METHYL PHOSPHINEOXIDE AND TRIBUTYLPHOSPHATE IN DODECANE SSC-02 STUDIES ON THE EXTRACTION OF SOME Lohithakshan K.V., Nair P.S., ACTINIDES FROM ACIDIC MEDIUM BY Raghuraffian K., Jadhav A.V., D.IHEXYL- N,N-DIETHYL CARBOMYL Jain H.C. METHYLENE PHGSPHONATE (DHDECMP) SSC-83 EXTRACTION OF PLUTONIUM FROM PHOSPHATE Sagar V.B. , Pawar S.M.. , CONTAINING NITRIC ACID SOLUTIONS USING Joshi A.R., Kasar U.M., DHDECMP AS EXTRACTANT Sivaranakrishnan C.K. SSC-B4 STUDIES ON THE EXTRACTION OF P>j SSC 29 CRYSTAL STRUCTURE OF K,Pu SSC 2 OXIDATION STATE OF URANIUM IN U30? Jayadevan N.C., Keskar M. FROM ST^ID STATE REACTIONS SSC -33 POSITION OF ACTINIDES IN THE Newton Nathaniel T. PERIODIC TABLE : EVOLUTION OF A GENERALISED FORM ssc- 34 SOLVENT EXTRACTION OF URANIUM SST--01 RECYCLING OF PLUTONIUM FROM FUEL Joshi rt.R.Charyulu M.M,Ghadse F),R, FABRICATION SCRAP Kadam A.V, Kssar U.M, Pawar S.M., Naronha D.M., Pius I.C., Ray M., Sager 7.B., Si var amal-.r i shnan C.K. SST-02 REMOVAL OF LOW LEVELS OF URANLUM FROM Sudersanan M., Iyer R.K. AQUEOUS SOLUTIONS BY COPRECIPITATION AND ION EXCHANGE SST-03 SEPARATION OF URANIUM AND IRON BY Mary Xavier, Mathew K.A., ANION EXCHANGE IN HYDROCHLORIC ACID Nair P.R., Patil B.N., Jain H.C. SBT-04 MASS TRANSFER CHARACTERISTICS OF Hareendran K.N., Pushparaja, URANIUM IN DM2 ETHYL HEXYL) Subramanian M.S. PHOSPHORIC ACID -TRIOCTYLPHOSPHINE SST-05 SEPARATION OF URANIUM AND PLUTONIUM Chetty K.V., Godbole A.G., USING MACROPOROUS ANION EXCHANGE Mapara P.M., Rajendra Snsrup RESINS FROM MIXED SOLVENT MEDIA : COLUMN STUDIES SST-06 RECOVERY OF PLUTONIUM FROM PHOSPHATE Pius I.C. , Charyulu K.M., CONTAINING AQUEOUS ANALYTICAL WASTE Si varanaliri shpan C.K. SOLUTIONS USING MACROPOROUS ANION EXCHANGE RESIN SST-87 STUDIES ON THE DEGRADATION OF ANION Dhani P.S., Gopalakrishnan V., EXCHANGER EMPLOYED FOR PLUTONIUM Raoanujaa A., Dhu*M«,d R.K., PURIFICATION Sudersanan M. IX SST-08 STUDIES ON POLISHING OF DEGRADED PURE* Singh R.K., Nair M.K.T., Bajpai SOLVENT FOR PARTIAL RECYCLE D.D., Varadarajan N., Gurba P.B., Gupta K.K., Rajandra Kumar SST-B9 CARRIER-MEDIATED TRANSPORT OF TRACE Anil Kumar, Singh R.K., QUANTITY OF PLUTONIUM ACROSS Nair M.K.T., Shufcla J.P. DICYCL0HEXAN0-18-CR0WN-6 ' TOLUENE I. IQUiD MEMBRANE FROM LOW ALPHA WASTE SOLUTIONS SST-10 A MODIFIED METHOD FOR THE SYNTHESIS OF Varadarajan N., Gabriel J., PO1Y CONDENSATE PHENOLiC RESIN WITH Pintu Sen, Nair M.K.T. IMINODIACETIC ACID (IDA) FUNCTIONAL GROUP FOR TREATMENT OF ALKALINE WASTE FROM REPROCESSING PLANTS SST-il FEASIBILITY STUDY OF VACUUM DISTILLA- Yeotikar R.6., Kaushik C.P., TION AS A TECHNIQUE FOR DECONTAMINATION Kanwar Raj OF KEROSENE RADIOACTIVE WASTE SST-12 ELECTRO-OXIDATIVE DECOMPOSITION OF Gubra P.B., Singh R.K., Singh R.R, ALPHA CONTAMINATED iON EXCHANGE RESIN Nair M.K.T., Venugopal A.K., Bajpai D.D., Dharkas S.P. SST-13 ELECTROLYTIC DESTRUCTION OF NITRIC Palamalai A, Rajan S K, Sanpath M, ACID IN REPROCESSING STREAM Chinnusamy A,Govindan P,Mohan S.V, Raman V.R, Balasubramanian G.R. SST-14 ELECTROOXIDATIVE DESTRUCTION OF Palamalai A, Mohan S.V, ORGANIC WASTES GENERATED IN A Govindan P,Chinnusamy A,Sanipath M, REPROCESSING PLANT Raman V.R, Balasubramanian G.R. SST-15 EVALUATION OF STABILITY OF EPOXY Rao S.V.S., La) K.B., Amalraj R.V. MATRICES FOR WASTE IMMOBILISATION USING i37Cs AS A TRACER SST-16 EFFECT OF DILUENT ON EXTRACTION OF Suresh A., Srinivasan T.G., URANYL NITRATE BY TRI-N-BUTYL VasudevraoP.R. PHOSPHATE SST-17 RECOVERY OF URANIUM USING Sabharwal K.N., Vasudeva Rao P.R., Bi-FUNCTIONAL RESINS Srinivasan M. SST-18 LABORATORY STUDIES ON REMOVAL OF Venkatesan M, Ravi T.N, Devarsjan TRIBUTVL PHOSPHATE FROM AQUEOUS M.C.. Natarajan R., Raman V.R., SOLUTIONS IN PUREX PROCESS Bal ?.5ubramani a^ G.R. AK - ACTINIDES AND ENVIRONMENT AE-01 Geochemical Association of Awericium Matkar V.M., Narayanan U, in Troffiba/ Sediment Bhat I.S, Pillai K.C. AE-02 URANYL UPTAKE IN CEMENT HYDRATION Labha=etwar N.K., Shrivastava DP. PHASE : llBTOBET^ AE-03 PREPARATION OF "- BY 235U (a,3n) Singh R.J, Deshmuk.h S.M, REACTION A., Rattan S.S., Satya Prakash AC ANALYTICAL CHEMISTRY AC-01 KADIOCHEMICAL STUDIES IN CHEMICAL Dhawale B.A., Raje«>wari B. , SEPARATION OF RARE EARTHS IN PLUTONIUM Bangia T.R., Ssstry M.D., AND THEIR ESTIMATION BY EMISSiON Natarajan P.R. SPECTROGRAPHIC METHOD AC-02 A NEW METHOD OF NON-DESTRUCTIVE ASSAY Gubbi G.K, RamasMami A, OF PLUTONIUM IN SOLUTION USING A Singh R.J, Satya Prakash, SINGLE ISOTOPIC GAMMA-RAY SOURCE Natarajan P.R. AC-03 RCLE OF URANIUM AND PLUTONIUM MATRICES Goyal N, Purohit P.3. , IN THE ATOMIZATION OF Be, Sn AND Zn Page A.G., Sastry M.D. FROM C GRAPHITE FURNACE AC-04 DIRECT SPECTROGRAPH1C DETERMINATION OF Venkatasubramanian R. SUB PFM LEVELS OF Ba, Li & Sr AND PPM LEVELS OF Cs, V. h Ua IN NUCLEAR GRADE THORIUM OXIDE i T h 02 ) AC-05 BACK EXTRACTION OF^ThdV) FROM ITS TTA Rastogi R.K., Mahajan N.A., COMPLEX IN BENZENE BY AQUEOUS FLUORIDE Chaudhuri N.K. AND ITS APPLICATION IN THE ANALYSIS DF FLUORIDE IN NUCLEAR FUEL SAMPLES AC-06 DIRECT AND SIMULTANEOUS SPECTROPHOTO- Moorthy A.D., Gurba P.B., METRIC DETERMINATION OF URANIUM AND Chaugule G.A., Varadarajan N., ACIDITY USING A FLOW CELL Singh R.K., Nair M.K.T. AC-07 SIMULTANEOUS DETERMINATION OF URANIUM Kaushik C.P., Yeotikar P.G., AND IRON IN HIGH LEVEL RADIOACTIVE Kanwar Raj WASTE STREAMS USING DUAL WAVE LENGTH TECHNIQUE OF SPECTROPHOTOMETER AC-08 SIMULTANEOUS SPECTROPHOTOMETRIC George Thomas, Patil B.M., DETERMINATION DF URANIUM(VI) AND VaradarajanN., Singh R.K., IRON! Ill) IN PUREX PROCESS STREAMS Bajpai D.D., Mair M.K.T. AC-09 EMPLOYMENT OF GRAPHITE ELECTRODE IN Sharma H.S., Manolkar R.B., THE COULOMETRIC DETERMINATION OF Marathe S.G. URANIUM AND PLUTONIUM AC-lfl EXPERIMENTAL DESIGN AND STATISTICAL Yadav M.8., Hari Singh, ANALYSIS OF DATA TO ASSIGN A VALUE TO Vaidyanathan S., Sood D.D. URANIUM CONTENT IN RUBIDIUM URANIUM(IV) TRISULPHATE - A POSSIBLE STANDARD REFERENCE MATERIAL (SRM) FOR URANIUM AC-ll A TITRIMETRIC METHOD FOR THE Keshav Chander, Hasilkar S.P., SEQUENTIAL DETERMINATION OF THORIUM Jadhav A.V., Jain H.C. AND URANIUM AC-12 QUANTITATIVE ESTIMATION OF PLUTONIUM IN Dhas A.J.A., Rakshe P.R., Michael DEGRADED ANION EXCHANGE RESIN AND TRI- K.M, Yadav R.U, Singh V.P, Vijayan BUTYL-PHOSPHATE BY NEUTRON COUNTING K, Ramamoorthy N., Kapoor S.C. AC-13 EXTRACTIVE RADIOMETRIC DETERMINATION Sreenivasa Rao K., Inamdar G.A., OF UPANIUH-233 WITH 1 -<2-PYPIDYLAZO)-2 Kulkarni R.T., Mukherji A., NAPHTHOL (PAN) IN THOREX PR0CF3S Ramanujam A., Dhumwad R.K. SAMPLES AC-14 STUDIES ON ALPHA LIQUID SCINTILLATION Manolkar R.B., Keshav Chander, COUNTING FOR THE DETERMINATION OF Marathe S.G. PLUTONIUM IN SOLUTIONS AC-15 EXTRACTION CHROMATOGRAPHIC STUDIES OF Ray U.S., Mitra K. Th(IV) AND ITS ANALYTICAL APPLICATIONS AC-16 SIMULTANEOUS MASS SPECTRONETRIC Chitambar S.A., Khodade P.S., ANALYSIS OF URANIUM AND PLUTONIUM FOR Parab A.R., Jain H.C. DETERMINATION OF CONCENTRATON OF U & Pu IN DISSOLVER SOLUTION OF SPENT FUEL AC-17 DETERMINATION OF NITROGEN, FLUORINE Ramakunar K.L., Raman V.A., AND CHLORINE IN U-jOg FUEL MATERIALS Sant V.L., Kavimandan V.D. , USING SPARK SOURCE MASS SPECTROMETRY Jain H.C. AC-IB QUANTITATIVE DETERMINATION OF UO2 IN Khan K.B., Jain G.C., Ganguly C. UN-UO2 MIXTURE BY X-RAY DIFFRACTION METHOD XI IT - INVITED TALKS IT - 01 to IT - 30 HISTORICAL ACHIEVEMENT OF PLUTONIUM ISOLATION/DRY CHEMISTRY," AND THERMODYNAMICS AT THE CHICAGO MET LAB Edgar F. Westrum, Jr. Department of Chemistry, University of Michigan Ann Arbor, MI 48109 USA SUMMARY — paper deals wilh the micro-metallurgy of plutonium metal and the isolation of clement 9f as a pure metallic phase. The physical and chemical properties, the development of thermodynamics and some anecdotal experiences are noted. (Key Words: plutonium—metallurgy and properties, thermodynamics L INTRODUCTION The challenge of being involved in the "microgram-scale" exploration of the properties of plutonium during the wartime Metallurgical Laboratory was an opportunity indeed; the honor of being selected as a participant in the historical aspecis of this Indian symposium to Commemorate the 50th Anniversary of the Discovery of trie Transuranium Elements is a pleasure. I am delighted to have the chance 10 present halfway around the world from my usual scene of activities some historical aspects from the early days of these important developments in which we labored beneath a deep veil of secrecy and compartmentalization of knowledge. Indeed, many other survivors of this era learned the "whole" truth only at the "Symposium to Comniemorate the 50th Anniversary of the Discovery of the Transuranium Elements" in Washington, D.C., in 1990 III or even more recently in a good historical summary volume HI. The classic words of Lewis and Rancall's 1921 Preface to Thermodynamics seem strangely appropriate 131. "There are ancient cathedrals which, apart from their consecrated purpose, inspire solemnity and awe. Even the curious visitor speaks of serious things, with hushed voice, and as each whisper reverberates through the vaulted nave, the returning echo seems to bear a message of mystery. The labor of generations of architects and artisans has been forgotten, the scaffolding erected for their toil has long since been removed, their mistakes have been erased, or have become hidden by the dust of centuries. Seeing only the perfection of the completed whole, we are impressed as by some superhuman agency. But sometimes we enter such an edifice that is still partly under construction; then the sound of hammers, the reek of tobacco, the trivial jests bandied from workman to workman, enable us to realize that these great structures are but the result of giving to ordinary human effort a direction and a purpose." The extraordinarily complex research endeavor at the wartime 'Met Lab1 at the University of Chicago) in the 1944 era was an intense period of projjress in inorganic uranium and transuranium element chemistry. I leave it to others to present the "view from on high." But perhaps the time is ripe for a somewhat lighter fare—-"the view from the trenches"—as it might be called Although earlier research on plutonium was done on the tracer scale, the "microgram- scale" research was the order of the day when this author arrived at Chicago in March of 1944 n - ()i.i fresh from the halls of academe at Berkeley. Here in the Met Lab aqueous chemistry was performed at more or less ordinary chemical concentrations by creating ultra-miniature balances, specially adapted microscopes, etc., and employing these with ultrasmall capillary containers, pipcts, and burcts with only microliter volumes. When extreme precision was needed, e.g., in specific activities studies, a microgravimetric approach was utilized /4/. "Dry" and "amphibious" chemistry has its special hazards, and micro-metallurgy with highly radioactive materials posed health, handling, and "recovery" problems that by the presem time have been solved, but the facilities then can only be described as "primitively rustic " In fact, one does not exaggerate in noting that the absence of precautionary safety measures could be described as little short of criminal. Awareness of some hazards had not been developed, and appropriate facilities arid/or equipment were in many instances simply not available. The wartime urgency of getting on with the tasks encouraged some risk taking. Everything considered, it is a little short of miraculous that the scientists in the "New Chemistry" and related campus structures survived nearly unscathed. The "cubicles" for metal production in Room 41 of "New Chem" may have offered token protection to other laboratory workers, but no' Ac slightest to the operator of the high temperature microfumace. Despite such conditions—metallic plutonium was produced and in high purity, its metallurgical production studied in some detail, its physical properties, and chemical reactivity were explored. Before early 1944 the total amount of plutonium in the laboratory never exceeded a few hundred micrograms, but gradually thereafter milligram amounts began to become available and we graduated to the so-called "semimicro" scale! The isolation of the first compound of 2 9 plutonium using cyclotron produced 94Pu ^ by Cunningham and Wemer /5/ represents a real milestone after the earlier discovery of the isotope itself by Kennedy, Seaborg, Segre, and Wahl /6/. A similar outstanding achievement was the production of the element itself in metallic form. Section C-J of the broader Chemistry Division under the direct supervision of Professor Glenn T. Seaborg was concerned with the development of chemical procedures for the extraction of plutonium, for the purification of plutonium—and in its later phases—with research on the isotopes of other heavy elements including other transuranium elements. The fast pace of its history and its development—compiled from some 60 categories of documentation including progress reports, notes, personnel records, patent files, health and administrative bulletins, and laboratory notebooks—provides a rich record not only of the scientific day to day achievements, but of the social life embellished with additional information on the then curreni, newsworthy events. Four large tomes of this monumental undertaking by Professor Seaborg permit even today a very comprehensive review of the progress of the research endeavor /7-10/. That tiiis was assembled from letters, reports, diaries, and miscellaneous documents is a real achievement. II. ISOLATION AND MICROMETALLURGY OF PLUTONIUM The investigations summarized in this paper comprise microchemical and micromctallurgical studies of plutonium made during 1943 and 1944 under the pressure of wartime conditions at the Manhattan Project Metallurgical Laboratory operated by the University of Chicago for the United States War Department. These studies led to the first successful preparation of plutonium metal and provided data essential to the macroscale technology of plutoniuin production that followed very soon thereafter at Los Alamos. A major chemical problem in the development of the release of nuclear energy was the preparation on the microgram scale of globules of plutonium metal suitable for the investigation of its physical and chemical properties. MicrometaJlurgy on this scale was unprecedented; an entirely original approach was necessary. The uncertainty inherent in the prediction of the physical and chemical properties of a new clement (in a series the very nature of which was still under speculation), the exceedingly small amounts of plutonium available, the doubt as to the IT - 01.2 identity and composition of compounds to be reduced, and the extreme health hazard of traces of the finely divided solids aggravated the problems. The complex allotropy of plutonium which was not suspected at the outset also contributed to the difficulty of these investigations. Preliminary experiments were made on "stand-in" compounds which it was hoped would simulate the behavior of plutonium compounds in reduction to the element The objective was reduction on the microgram scale, with high yields of metal iji a single globule. A variety of methods was tested, including reduction with atomic hydrogen, electrolytic reduction in gaseous and fuscd-salt systems, and thermal reduction of halides with alkali and alkaline-earth metals; only the last-mentioned proved successful. On the basis of the limited data available, uranium compounds were selected as probably representative of the chemical reactivity and physical properties of the analogous plutonium compounds. UF4 was chosen because of its stability, non- hygroscopicity, and completive non-volatility. It was anticipated that P11F4 would have a similar set of desirable properties. As noted, one of the initial goals of Section C-I in 1944 was the isolation of the element plutonium and the study of its preparation in high purity as well as its physical and chemical properties /11/. Such information would prove crucial in the subsequent large scale developments at Los Alamos Scientific Laboratory in New Mexico where the bomb itself would be fabricated. But let me digress to comment on some of the problems involved in working on either the miciogram scale or even on the milligram scale. By good fortune—or by design—one of the team members was a watchmaker and another a skilled instrument m?ker. By buying our own chests of machinists tools, some of us worked evenings in the machine shop to produce some of the special equipment we urgently needed for daytime research in supplement to our already heavy demands on the instrument shop personnel. The first unequivocal production of plutonium metal was made in November, 1943 /ll/. About 35 jig of P11F4 were transferred on the tip of a sharp needle to a small thoria crucible. About ten times the sjtoichiometric amount of barium metal freshly cut under xylene was placed in a larger crucible, the small curcible was inserted on top of the reductant, the outer crucible cap inserted, and the entire assembly placed in a microfurnace and fired from 1,200 to 1,400 *C in a high vacuum. A number of 3-jJ.g plutonium-metal globules were formed. The metal had a silvery luster, a density of about 16 gm/cm3, and rapidly absorbed hydrogen at about 210 *C to form a black powder subsequently identified as PuH3 /11/. A typical Goldschmidt reduction using an alkali or an alkaline metal was one of the early techniques used to produce high quality plutonium metal. The reaction between condensed phases of reactants was generally violent and tended to yield dispersions of metal and slag. Techniques were therefore developed whereby only the reductant vapor came in contact with the halide to be reduced. This involved fabrication and careful degassing of a covered inner refractory crucible of beryllia with an internal volume of a fraction of a microliter. A fraction of a milligram of a plutonium halide (often in the form of a single cylindrical pellet) was placed in the inner crucible. A small freshly cut cube of metal—usually barium—was placed over a cover plate and the system enclosed within a refractory outer crucible with a conical, snugly-fit plug. The crucible system (qf. Fig. 1) was enclosed within an electrically energized tantalum microfumacc coil surrounded by a tantalum radiation shield. After attainment of a high vacuum, the run was fired for several minutes at temperatures in the vicinity of 1200 "C. The reaction is: 3Ba(g) + 2PuF3 -> 2Pu(l) •• 3BaF2 The flux, BaF2» is largely absorbed on the microscale by the rather porous crucible. IT - 01.3 Figure 1. The microdouble-crucible arrangement iri the tantalum-shielded resistance furnace for the reduction of plutonium compounds. On "good" days a small sphere of plutonium was obtained. How big was this sphere? Well, let's say even for 100 Jig of metal one could expect a diameter of about 0.004 mm (or about 0.1 of an inch mil—or about a few percent the thickness of a sheet of paper! Before glove boxes were manufactured, a wooden box with a glass top was our workstation. But that very dense little bead was tricky to handle with jeweler's forceps and bounced very well! It was not unknown for it to get on the floor. Today, its radioactivity would have enabled it to be quickly found with a hand-held portable counter, but although such indispensible devices were indeed in design stages, they were just becoming available. So with their derrieres high and noses low, the group were obliged to use their eyes to recover from the laboratory floor that not only hazardous bit of matter—but what at that time represented a significant fraction of the laboratory's supply (really the world's supply!) of plutonium in all forms. About that time the one predictable, virtually routine daily ceremony would occur. Five Nobel laureates would enter the laboratory and be introduced to each member of the team and vice versa—despite the fact that that ceremony had probably already occurred several times that week! As soon as the distinguished visitors departed we returned noses to the floor until our product was retrieved. It should be appreciated that i;i early periods even the oxidation state (and formulas of the compounds being reduced were not known) and fluorides, for instance, were often designated only by color—such descriptions were refined gradually as the chemistry X-ray diffraction studies helped to enable a more precise identification. Even reference to the better known chemistry of uranium was only moderately instructive as handbooks and texts of that time clearly confused the metal with the monoxide. A series of studies on the reduction of PuF4 with Ba in BeO crucibles indicated that optimum results were obtained at about I,150*C. Although a single globule was desired, the results were considered acceptable if the over-all yield of compact globules was high and each globule represented at least 20 per cent yield. Reduction of PUCI3 to metallic plutonium was achieved in fair yield by the action of Li, Na, K, Ba, and Ca. Plutonium trifiuoride is easily prepared, has desirable physical characteristics, and has been more thoroughly investigated from the standpoint of metil production on the milligram scale in this laboratory than have the other halides. Yields were excellent, and the few failures could generally be traced to faulty technique or secondary reactions with crucible material. Almost 200 microscaJe reductions of P11F3 were made in beryllia. With the very rapid progress of basic chemistry in the C-I Section, production of metal of higher den ,ity, greater hardness, and higher purity was soon achieved. These samples were promptly snu ied for physical properties (X-ray diffraction pattern, malleability, melting point, phase interc< nversion, and volatility) and hydrogen uptake in forming hydrides. Although PuOoi PUCI3, PuF , and PuF4 were reduced on occasion, frequently with Ba metal as the reductant metal (Li and Ca were also used succssfully on occasion), the goal iyas to get a single sphere of IT - 01 .4 product in high purity. The purity of our product was 99.75 per cent by spectroscopy and chemical analysis. Moreover, typical yields varied between 86 and 92 per cent- Various refractories were tried as crucible materials: these included ThOo, La2Q3, Ta, nitrided tantalum, CaF2, ^ CaO- ThO2 crucibles contaminated the product with thorium metal, and—in general—none seemed as desirable as BeO2 in producing high purity, high density metal—in high yield- The extensive line of cerium sulfides was produced at the Department of Chemistry of the University of California at Berkeley by Leo Brewer. These and binary sulfides for ultimate large scale plutonium production were also individually proved to be of value in subsequent larger stale experiments at Los Alamos. Our work—despite the disparities of scale—proved valuable in the later, larger scale production of weapons-grade plutonium at Los Alamos.. Two attempted reductions of PuOCl with potassium yielded only the phase identified both by its characteristic sernimetallic appearance and by X-ray diffraction analysis as PuO. Attempted reductions of PuC>2 with na and Mg vapor at about 1,200* C did not show evidence of reduction; Ca and Li vapor at 1,200" C yielded only partial reduction to a phase tentatively identified as Pu2O3. With the passage of time, glove boxes to better confine the pelleting of the halides and the loading of the crucibles became available. (The insertion of the crucible into the furnace continued to be done in the room or in the cubicle.) Principals in the metal production process of this era, in addition to the author, were H. L. Baumbach, Group Leader (recruited from the Hollywood film industry chemistry), Zene Jasaitis (a chemist and—by good fortune—an amateur watchmaker), Sherman Fried (a resourceful chemist), Herman Robinson (a chemist with mechanical, electronic, and ingenious inventiveness to boot). It is interesting to learn how plutonium metal is produced today on the kilogram scale. li involves calcium metal and PuF4 in the presence of iodine so that both the Cu +I2 -* Cal2 and PuF4 + 2Ca -» Pu(l) + 2CaF2 in CaO refractory crucibles represents current technology and the processes seem to be still under improvement/12/. Usually the metal is purified by subsequent refining processes. IIL PREPARATION OF OTHER PLUTONIUM COMPOUNDS One attempt at metal production using OaSi2 as a reactant resulted instead in creation of beautiful, brightly metallic in lustre crystals of PuSi2 /13/. The reaction is simply: PuF3 + 1.1 CaSi2 -* PuSi2 + 1.1 CaF2 + 0.2 SiF4. Other studies included those on plutoniurn nitrides /14/, oxides /15/, oxychloridcs /It/, sulfides and oxysulfides /17/, etc. The nitride was readily produced from the trichloride will! ammonia: PuCI3 + 2NH3(g) -> PuN + 3HCl(g). M • 01 •> Time does not permit further elaboration of the chemistry which has been extensively treated in the literature. Other compounds were, of course, studied by other groups simultaneously in the Section C-l. The cooperation of Professor W. H. Zachariasen was of tremendous help by his use of X-ray diffraction in identifying structures. For a variety of reasons (including the isotopic composition of the pile-produced plutonium-238) it became desirable to standardize—after some enhancement of the precision—the assay technique. To get higher precision than could be obtained with micropipettes, a weighing technique of the droplets delivered was developed /4/ and the specific activity determined l\ 8/. Amusingly, this Westrum value was designated by the Board of Senior Responsible Reviewers as the only value of the specific activity of pile produced plutonium to be used in the open literature for a period of several years to avoid revelation of the extent of fission in the pile material. The health and safety aspects as already noted were obviously other areas of considerable concern to all of us. The radioactive and chemical aspects of working with plutonium were under active study by the Health Physics Group and many interesting eras were endured. A favorite study of the health physics group was the microscopic examination of fingernail ridges. In trying to displace plutoniuru froiTi bones by calcium preparations "Calcium' pills of such size that they must have been designed for use by horses rather than humans *\irc prescribed. Probably berylliosis—akin to silicosis—represented a more serious threat th^ii any of us in the Metal Production Group realized and yet, so far as I know, none of us wcrt afflicted. Individuals at other sites suffered. Yes, the record shows that overly zealous cleaning women on several occasions dumped piutonium solutions from labelled flasks and beakers into the sink drains. Bunsen burners connected by mistake to high-pressure air hoses Lnsiead of gas lines whipped around breaking glassware containing solutions; centrifuge vials occasionally fractured. Somehow we survived! A largely hairless baby mouse discovered in a Coca Cola bottle from the vending machine caused die collapse of the young woman who had drunk the coke before discovering the mouse! This was one of the more memorable crises. Several year? ago, I received letters from medical personnel associated with our national nuclear program offering me free burial if I would make my (presumably) radioactive body available for study upon my death. It seemed like a sporting proposition, but after submission of appropriate samples, they declined on the grounds there was insufficient radioactivity embedded to make me worthy of post mortem study. I am grateful to have avoided the complications of being radioactive, but must provide burial at my expense. IV. THERMODYNAMIC ASPECTS OF PLUTONIUM CHEMISTRY When in the 1945 era pressure on metal production slacked because the Los Alamos production needed research on a scale so large that only they could provide it, we initiated some thermochemical studies to enhance our understanding of the chemistry of plutonium. Such studies were greatly encouraged by Professor Wendell M. I -atirner on his visits as a consultant from the University of Califoma (at Berkeley). Initiatory studies were made in an enthalpy of solution calorimeter constructed for use with radioactive materials /18,19/. The earliest studies involved measurements of TI1CI4 /20/, P11F3 IT - 01.6 fl\f P11CI3 1221 (including the effect of concentration of the aqueous hydrochloric acid /23/), and fPB^4/ My concern for tiic thermodynamics of materials involved in the production of nuclcaz energy has not flagged—despite other overwhelming interests—through the decades of the intervening almost 50 years. In the late 40's and the early 50's the studies begun by Westrum et al. at Chicago were continued at the University of California's Radiation Laboratory at Berkeley. Daring that time and subsequently I continued to collaborate at the Argonne National Laboratory (Argonne, IL) as a visiting scientist. These thrrmochermcal and thermophysical experiments involved interactions with Dr. Darrell W. Osborne, Dr. John B. Hatcher, Dr. Harold R, Lohr, and to a lesser extent, others. In the process we designed adiabatic calorimeters capable of measuring at heat capacities at equilibrium from 5 to about 350 K, etc. By the mid 50's we had adiabatic cryogenic calorimetric facilities at the University of Michigan as wel! as superambient adiabatic caloriraetry to 600 K and continued to study uranium and thorium compositions since these could be readily handled in the academic milieu with but minor radioactivity safety complications. In recent years a collaboration with Dr E.H.P. Conifunfce of Reactor Centrum Nederlands (Petten> The Netherlands) has resulted in the exploration of the thermodynamics of high-yield fission products and of their reactions with the nuclear fuel, the cladding—and for more serious accidents—with the concrete of the containment pits /25/. For example, complete thermophysical and thermochemical studies have been made over a considerable temperature range (from about 5 K to high temperatures) on a-, (}-, and f UQ3, UPd3, URui, URh3, CsBO2, RuO,, In2O3, p~ Ba(OH)2, Sr(OH)2, CsOH, CS2R11O2, Cs2MnO2, etc., so that Jie rearfivmes—and volatilities—-of the product compounds can be assessed. The first critical compilation of therrnochemicai data for plutonium and its compounds was that of Brewer et al. published in 1949 /26/. This was based on the early work carried out mainly in the USA, shortly after plutonium compounds were available in macro amounts. All the investigators involved in this early research deserve the greatest credit since, working often with only scant knowledge of the full properties of the systems involved, they produced results which have—to a large extent—been substantiated by later work. The 1963 book on the thermodynamics of uranium compounds by Rand and Kubaschewski 121! and the review paper by Getting in 1967 represent further updating, assessing, and evaluation of the thermodyanmics of the actinides /28/. Other surveys of the thermodynamic properties of certain actinide materials have indeed appeared since 1967, but they are either of limited breadth, or have not been critically evaluated. Even taken as a whole, these data are not in a format ideally suited for cither science or technology. Over the sixties and seventies the IAEA (International Atomic Energy Agency) under the direction of Dr. Reinosuke Hara initiated a series of "panel" discussion involving small groups of experts as well as much larger symposia (1962, 1967, 1979) which deal with the important aspecis of the thermodynamics of nuclear reactor materials. About 1975 the International Atomic Energy Agency provided a series of monograph? on the physico-chemical properties of special reactor materials. One treats plutonium /2% but others do not necessarily deal with transuranium substances. Consequently, the IAEA initiated in 1976 a new series of monographs on The Chemical Thermodynamics of the Actinides /30/. Part 4, for example, deals with the actinide chalcogenides (excluding oxides) /31/ and presents among other critical evaluation an interesting analysis of heat-capacity contributions to the population of excited electronic levels ir. die actinides (Schortky contributions). Unfortunately, several key volumes have not been completed. These include the oxides and pnictides among others. IT - 01 .7 The endeavors of COD AT A in the critical evaluation of thcrmodynamic dsia and the generation of self-consistent tables and evaluation schemes will influence the production of subsequent actinide tables greatly in the next decade /32,33/. Cordfunke and Konings an: to be comr-'imented for the production of a compilation on Thermochemical Data for Reactor Materials and Fission Products /34/—excellent in all respects and particularly noteworthy in that the importance of rnclar volumes has not been neglected. This summarizes much relevant work of ihe type described in the previous section and other relevant studies to give a complete thermochemical and thernjophysical analysis. The data provided arc relevant to safety studies for al> reactor systems, and particularly to light water and fast breeder reactors. It is to be hoped that these activities will help 10 abate the irrational fear of nuclear energy in the United States and usher in the era that was foreseen from the initiation of the reactor era. In any event, it can help to ameliorate the severity of such accidents as do occur by judicious selection of reactor materials. The evaluation work of the NEA-TPB (Nuclear Energy Agency—Thermochemical Data Bank of the OECD, Gif-sur-Yvette, France) on the uranium volume /35/ and perhaps also on subsequent volumes critically reviewing iir. thermodynamics of elements of particular importance in the performance assessment of radioactive waste disposal systems is also noted. This endeavor should also serve as a basis (or reassurance to calm the now nearly-catastrophic parallel concerns for the planet's safety in the dispoal of nuclear wastes in the face of increasing needs for nuclear energy. In conclusion, the team endeavors that produced high quality plutonium metal involved a number of Section C-I scientists and were "...hui the result of giving to ordinary human effort a direction and a purpose." V. REFERENCES 1. To be presented in the ACS Non-Series Book, '50th Anniversary of the Discovery of the Transuranium Elements,' American Cnemical Society, Washington, DC. [Expected 1991] 2. F.G. Gosling, The Manhattan Project: Science in the Second World War,' Energy History Series, U. S. Dept. of Energy, Office of Administration and Human Resources Management, Executive Secretariat, History Division, Washington, DC. (August 1990). 3. G.N. Lewis, M. Randall, Them odyamics.,1 McGraw Hill, New York (1921). 4. E.F. Westrum, Jr. 'National Nuclear Energy Ser., Div. IV, Plutonium Project Rec. Vo1. 14B. The Transuranium Elements: Research Papers, Part II, Paper 6.2,' Edited by G.T. Seaborg, J.J. Katz, W.M. Manning, McGraw Hill: New York (1949), p 1185. 5. B.B. Cunningham, L.B. Werner, Ibid , 'Paper 1.8,' p. 510. 6. J.W. Kennedy, G.T. Seaborg, E. Segre, A.C. Wahl, Ibid., Paper 1.2,' p 50. 7. G.T Seaborg, 'History of Met Lab S.xdon C-I for the period April 1942 - April 1943,' Lawrence Berkeley Laboratory Pub. 312, Vol. 1 (February 1977). 8. G.T. Seaborg, Ibid... (or the period May 1943 - April 1944, Pub. 112, Vol.II, (May 1978). 9. G.T. Seaborg, Lbia., for the period May 1944 - April 1945, Pub. 112, Vol. Ill (May 1979). 10. G.T. Seaborg, !bi-i.: for the period May 1945 - May 1946, Pub. 112, Vol. IV (June 1980). IT - 01.8 11. S. Fried, E.F. Westrum, Jr., H.L. Baumbach, P.L. Kirk, "The Microscale Preparation and Micrometallurgy of Plutonium Metal," in The Metal Plutonium,' Edited by A.S. Coffinberry and W.N. Miner, University of Chicago Press, Chicago (1961). 12. N.M. Edels;tein, 'Actinides in Perspective', Pergamon Press, (1982). 13. EJ\ Westrum, Jr., 'National Nuclear Energy Ser., Div. IV, Plutonium Project Rec. Vol., 14B. The Transuranium Elements: Research Papers, Part I, Paper 6.5,' Edited by G.T. Seaborg, JJ. Katz, W.M. Manning, McGraw Hill, New York (1949), p 729. 14. 3.M. Abraham, N.R. Davidson, E.F. Westrum, Jr., Ibid., 'Part II, Paper 6.60,' p 945. 15. Ei\ Westrum, Jr., Ibid.,' Part n, Paper 6.57,' p 936. 16. E-F. Westrum, Jr., H.P. Robinson, Ibid., 'Part n. Paper 6.56,' p 930. 17. B.M. Abraham, N.R. Davidson, E.F. Westrum, Jr.. Ibid., "Part I, Paper 6.18', p. 814. 18. E.F. Westrurn, Jr., H.P. Robinson, Ibid., 'Part U. Paper 6.51,' p 889. 19. E.F. Westrum, Jr. Thermochemical Measurements on the Transuranium Elements.' U. S. Atomic Energy Commission Declassified Document, AECD-1903: Oak Ridge, Tennessee (1948). 20. E.F. Westrum, Jr., H.P. Robinson, 'National Nuclear Energy Ser., Div. IV, Plutonium Project Rec. Vol., 14B. The Transuranium Elements: Research Papers, Part n. Paper 6.50,' Edited by G. T. Seaborg, JJ. Katz, W.M. Manning, McGraw Hill, New York (1949), p 887. 21. E.F. Westrum, Jr., L. Eyring, Ibid., 'Part II, Paper 6.52,' p 908. 22. E.F. Westrum, Jr., H.P. Robinson, Ibid., 'Part D, Paper 6.53,' p 914. 23. H.P. Robinson, E.F. Westrum, Jr. Ibid., 'Part II, Paper 6.54,' p 922. 24. E.F. Westrum, Jr., Ibid., 'Part It, Paper 6.55,' p 926. 25. E.H.P. Cordfunke, R.J.M. Konings, E.F. Westrum, Jr., J. Nucl. Materids, 167, 205-12 (1989). 26. L. Brewer, L.A. Bromley, P.W. Gillcs, N.L. Lofgren, in The Transuranium Elements,' National Nuclear Eneigy Series, Div. IV-14B, McGraw-Hill Book Co. Inc., New York,(1949), p. 861. 27. M.H. Rand, O. Kubaschewski, The Thermodynamic Properties of Uranium Compounds,' Oliver and Boyd, London (1963). 28. F.L. Oetting, Chem. Rev. 67,261 (1967). 29. M.H. Rand, D T. Livey, P. Feschotte, H. Nowotny, K. Seifert, R. Ferro, "Plutonium: Physico-Chemical Properties of its Compounds and Alloys," Edited by O. Kubaschewski, 'Atomic Energy Review,' Vol. 4, (1975), Special Issue No. 1. 30. F.L. Oetting, V.A. Mcdvedev, M.H. Rand, E.F. Westrum, Jr., Editors, The Chemical Thermodynamics of Aetinide Elements and Compounds,' International Atomic Energy Agency, Vienna (1972 et seq.). IT - 01 .9 31. F. Gr0nvold, J. Drowart, EP. Westrum, Jr. The Chemical Thermodynamics of Actinide Elements and Compounds, Pan 4, The Actinide Chalcogenides (Excluding Oxides),' International Atomic Energy Agency, Vienna, (1984). 32. J.D. Cox, D. D. Wagman, V. A. Medvedev, Editors, 'CODATA Key Values for Thennodynamics,' Hemisphere Publishing Corporation, New York (1989). 33. D. Garvin, V.B. Parker, HJ. White, Jr., Eidtors, 'CODATA Thennodynamic Tables. Selections for Some Compounds of Calcium and Related Mixtures: A Prototype Set of Tables/ Hemisphere Publishing Corporation, New York (1987). 34. I. Grcnthe, J.Fuger, RJ. Lemire, A.B. Muller, C. Nguyen-Trung, H. Wanner, 'NEA-TDB, Chemical Thermodynamics of Uranium,' OECD Nuclear Energy Agency (March, 1990), Final Dr;tft for Peer Review. 35. E.H.P. Cordfunke, R.J.M. Konings, Editors, Thermochemical Data for Reaaor Materials and Fission Products,' North-Holland, Amsterdam (1990). IT - 01 .10 STUDIES OF PROPERTIES OF PLUTONIUM BASED FUELS AT CADARACHE Yannick GUERIN, Yannick PHILIPPONNEAU Commissariat a I'Enc-rgie Atomique Cadarache, 13J08 Sasnt Paul lez -Durance, France SUMMARY : Some of the studies performed at Cadarache on (U,Pu)O2 are presented and discussed. We consider principally the thermal conductivity and the oxygen potential and draw attention on the influence of irradiation in these properties (Key Words : Uranium -plutmiiutn oxide, contiurlivity, oxygen potential). INTRODUCTION Uranium oxide is the most widely used nuclear fuel ; hut uranium plutonium oxide has also been largely experimented as fuel for Liquid Metal Fast Breeder Reactor and more recently for Pressurised Water Reactor (MOX fuel). For this reason, properties of uranium piulonium oxide have been extensively studied mainly during the sixties and seventies. Nevertheless in some Ujpics, fuel properties are still 1'ar from being perfectly known : t!-.is is particularly true in the high temperature region and for all that concerns the evolution of properties under irradiation. For LMFBR several alternative fuels are also considered : for example carbide in India, metal in the United States, or nitride in France. Research and development of these fuels include improvement of fabrication routes, basic research on fuel properties, irradiation experiments, post-irradiation examination and modelling of in pile behaviour. It is outside the scope of the present paper to make a complete review of the studies of fuel performed in France. Our purpose is just to highlight some aspects of the studies concerning two important properties of mixed oxide fuel : thermal conductivity and oxygen notential, trying to point out the few that is known about the influence of irradiation on these properties. IT - 02.1 I. THERMAL CONDUCTIVITY OF (U,Pu)O2-x Thermal conductivity is of course of primary importance for the designer as it allows to calculate the fuel temperatures ; but it is also a property very difficult to establish with high accuracy because of the great number of parameters playing a role : temperature, O/M, Pu/M, amount and type of porosity, burn-up, ... The last published recommendation of thermal conductivity is the Martin's one 111 which was a re-appraisal of the Washington's review 121. Since that time, new measurements have been performed /e.g. 3, 4, 5/ and it appeared necessary to make a new and complete review of the existing data in order to update the recommendation. This work is currently under achievement. Thermal conductivity A is generally deduced from thermal diffusivity a, thermal capacity Cp and specific mass p : \ = a. p .Cp 1.1. The specific mass, calculated from lattice parameter and thermal expansion 16! is known with good accuracy. 1.2. The thermal capacity of mixed oxide fuel may be regarded as well known up to about 2000 K ; but at very high temperature this property raises some questions. In UO2, a Bredig transition is likely to occur at 2610 K 17, 8, 9/ ; and it can 'oe seen on figure 1 that at this temperature, depending upon the author, the recommended Cp value may vary up to 35 %. On plutonium oxide and on urania-pluionia, enthalpy measurements at high temperature are too scarce to determine whether such a Bredig transition exists. Therefore in the different recommendation this possibility is discarded, which might be the source of an important error. To clarify this open question of thermal capacity at high temperature is one of our objectives. In that purpose we are setting up a drop calorimeter device in order to perform enthalpy measurements on uranium-plutonium oxide and other types of plutonium base'J fuel up to 3000°C. This apparatus will be operative in 1991 and has been designed to carry out in the same furnace enthalpy and diffusivity measurements. IT - 02.2 1.3. All published results of thermal diffusivity measured on uranium-plutonium oxide for every vslue of plutonium content, O/M ratio and porosity have been collected and transformed into conductivity data by using the same expressions of specific mass and thermal capacity 1.3.1. For plutonium contents of about 20 % (which have been the most extensively studied), it appears that : - Conductivity of stoichiomelric fuel is characterized by a-large scatter (figure 2). This is due to the fact that for this fuel, slight deviation from stoichiometry or small impurity contents may play an important role. - But for O/M = 1.98, as already notictd /10/, the scatter in the published data is lower and has been still reduced by using a single law of Cp ani)r, to deal wilh all diffusivity data (figure 3). This data set may be chosen as a good basis for a new recommendation. Our recommendation will be very near the Martin's one for stoichionietric fuel (figure 2) and for very low O/M ratios ; but a significant difference appears for O/M = 1.98 (figure 3) which is near the as-fabricated fuel. 1.3.2. The influence of plutonium content on thermal conductivity is not clear enough to be able to make a recommendation on this parameter. Nevertheless at stoichiometric state, for Pu/M range of interest in LMFBR, A of (U,Pu)O2 is lower than A of U(>2 in the low temperature region It is therefore necessary to know the Pu/M dependance of thermal conductivity for the low plutonium contents of MOX fuels and also to investigate the influence of plutonium homogeneity, i. e. the influence of the fabrication routes. This is a program we are starting now. 1.4. Pat the highest uncertainty on fuel thermal conductivity arises when we consider the influence of *> irn~up. During irradiation, from the beginning of life, fuel suffers a lot of changes which influence the thermal conductivity integral : - geometry : cracking, gap closure, - porosity : restructuration, gas bubbles, • O/M : radial redistribution of oxygen at beginning of life*, and global evolution afterwards, - Pu/M : Pu enrichment near central hole. All these variations of parameters are taken into account in fuel modelling of in-pile behaviour and the resulting shift of thermal conductivity is then automatically deduced. IT - 02.3 Defe< ts produced in the lattice by the displacement of atoms in the fission spikes might decrease the conductivity. These displacement damage effects are expected to play a role at low temperature and are generally discarded at irradiation temperatures. But the effect of fission products certainly requires consideration, especially when we know that in LMFBR fuel elements, target burn-up is now as high as 20 at %. There are three ways to obtain experimental information allowing to assess the evolution with burn-up of thermal conductivity and to try to isolate the fission product contribution to this effect: a- In-pile temperature measurements either by thermocouple in the central hole /ll, 12/ or by power-to-melt experiments /13/ give direct access to the fuel temperature. Such experiments are necessary for endorsement of fuel modelling codes but they are difficult to perform on highly burnt fuel and the results depend not only upon fuel conductivity integral but also upon gap conductance. As a consequence the influence of conductivity alone cannot be deduced with high accuracy. b- Our Japanese colleagues have attempted to perform diffusivity measurements directly on irradiated fuel /14/ : their results suggest an important decrease of X with burn-up. But to be sure to deduce from such measurements the influence of the fission products alone, a very complete characterization of burnt fuel (O/M, porosity, m'~rostructure, absence of cracking, etc..) is necessary, which is a quite difficult task in hot cell. c- For the above reasons, in order to make a recommendation of the influence of fission products on thermal conductivity, we considered mainly the few diffusivity measurements performed on oxides doped with appropriate impurities to simulate irradiation /3, 15, 16, 17/. The results again show a decrease of \ which is sometimes very important. But the raw results must be corrected to take into account the fact that in these oxides with simulants the O/M (M means here U + Pu + fission products in solid solution) is lower than in pure (U,Pu)02 sintered in the same conditions. After making the different corrections, it may be deduced that, after 10 at % buirn-up, the decrease of X due to fission products effect alone is of the order of 9 % at 1000 K and 4 % at 2000 K. But this is clearly a topic where uncertainties are important and additional experimental information is needed. IT - 02.A II OXYGEN POTENTIAL AND O/M OF IRRADIATED FUEL II. 1. Oxygen potential measurements Oxygen potential of unirradinted UO2 and (U,Pu)O2 has been determined as a function of temperature, Pu/M content and 0/M ratio by several techniques (thermogravimetry, transpiration, galvanic cell). And different models although empirical give reasonable agreement wih experimental results /e.g. 18, 19/ But the evolution of AG(C>2) during irradiation is still a pending question. And in LMFBR fuel elements, oxygen potential is the basis for understanding and predicting important phenomena such as clad corrosion and reaction with sodium. Until 1988 the most consistent experimental result on this topic was the measurements of w n Woodley on UQ 75?*U0.25^2-X ^ chemically added fission products simulating burn-ups of up to 10 at % 1201. It was deduced that oxygen potential increases with burn-iup, the lower the temperature and the higher the O/M ratio, the higher the increase. But some doubt was remaining whether such measurements were really representative of what occurs in pile. Thanks to a collaboration with our colleagues of Trans-Uranium Institute, measurements of oxygen potential were performed on irradiated fuel /21/. The oxide with 20 % plutonium and initial O/M ratio of 1 98 was taken from fuel pins irradiated in the Phenix reactor up to 3.8, 7.0, and 11.2 at %. The measurements were performed at Karlsruhe in a galvanic cell consisting of an yttria- (ioped thoria crucible which contains the specimens and of a Fe/FeO reference system. The results show a continuous and significant increase of Z1G(O2) with burnup (figure 4) : at 1200 K, this increase is about 63, 100 and 146 kJ/mole for 3.8, 7.0, and 11.2 at % burn-up respectively, that is to say an increase significantly higher than the one observed by ^oodley. 11.2. O/M of irradiated fuel The burn up dependence of oxygen potential is the consequence of three effects : - O/M increase due to the oxidative nature of Pu fission. - AG(O2) increase at constant O/M ratio due to the influence of fission products in solid solution. IT - 02.5 Buffering action by some Fission products compounds and cladding corrosion. Therefore, to be able to determine what is the shift of AG(C>2) due to the fission product effect alone, it is necessary to know the O/M value of the irradiated fuel. This O/M value had been previously deduced from lattice parameter measurements performed ten years ago at Fontenay-aux-Roses on fuel irradiated in Rapsodie and Phenix /22/. O/M of burnt oxide was calculated from the decrease of lattice parameter after an oxido- reduction treatment putting the oxide in the stoichiometric state. The results on Phenbi fuel irradiated up to 7 at % burn-up indicated an O/M comprised between 1.98 and 1.985 . It was difficult to explain this plateau of O/M at about 1.98 while /\G(O2> was continuously increasing. But we recently found an explanation : Tt has been known for a long time that self irradiation of plutonium based fuel (and other actinide compounds) causes an increase of lattice parameter with time /23/. We have » quantified this effect by performing over a long period (about 4 years) measurements < .«..ice parameter on (U,Pu)O2 pellets taken from the same batch. We observed a continuous increase of lattice parameter : about 60 pm in 3 years. But we also verified that it was possible to anneal these lattice defects induced by self- irradiation by maintaining the oxide specimen during some hours at 700"C. It became obvious therefore that in the measurements performed at Fontenay-aux- roses, the decrease (about 60 pm) of the lattice parameter after the oxido-reduction treatment (8 hours at 900°C) was due not only to O/M change but also (and may-be essentially) to annealing of the self-irradiation effect. We think now that the mean O/M of burnt fuel is probably near the stoichiometric state, but we have not yet a proof for that and an important experimental effort is being done in the hot cells of Cadarache in order to obtain a clear experimental determination of the O/M of irradiated oxide. The increase of A^K^) that appears on figure 4 is probably due to the additional effects of O/M increase during the first at % burn-up and AG(O2> increase afterwards. When we shall be able to isolate the effect of the O/M change, a good consistency is likely to appear between the measurements on irradiated fuel /21/ and the measurements on oxide with simulated burn-up /207. IT - 02.6 Th-,n measurements of oxygen potential on highly irradiated oxide will bring information on the buffering effects in fuel elements in relationship with cladding corrosion and radial and cxial migration of fission products. This is one of the objectives of a new series of ACKO2) measurements which is currently carried out. CONCLUSION Despite the large amount of research on fuel properties that have been done all over the world for many years, it still remains some specific but important topics in which the knowledge is not yet satisfactory. Our program is aiming at reducing these lacks of knowledge. Nevertheless it is interesting to note that for the two properties that have been treated here, thermal conductivity and oxygen potential of irradiated fuel, a review of old results with the help of new measurements allows to explain apparent inconsistency and to reduce the scatters. II - 02.7 REFERENCES 1. D.G. MARTIN, J. Nucl. Mat., 110, 73 (1982). 2. A.B.G. WASHINGTON, UKAEA Report, TRG Report 2236(D) (1973). 3. S. FUKUSHIMA, T. OHMICHI, A. MAEDA and M.IIANDA, J. Nucl. Mat, 116, 287 (19S3). 4. J.T.I. BONNEROT, CEA R 5450 (1988). 5. H.ELBEL and D. VOLLATH, J. NUCL. Mat., 153, 50 (1988). 6. D.G. MARTIN, J. Nucl. Mat., 152, 94 (1988). 7. M.A. BREDIG, Coll. Int. CNRS, 205, 183 (1971). 8. J.K. FINK, Int. J. Thermoph., 3, 2 (1982). 9. G.J. HYLAND, R.W.OHSE, J. Nucl. Mat., 140, 149 (1986) 10. M. BEAUVY, 11^ ECTP, Umea (1988). 11. M. CONTE, J.P. GATESOUPE, M. TROTABAS, J.C. BOIVINEAU and G. COSOLI, Int. Conf. on F.B.R. Fuel Performance, Monterey, 301 (1979). 12. L. BRUNEL, M. BOIDRON, A. LANGUILLE and M. PELLETIER, Nuclear fuel performance, Stradford-upon-Avon, BNES (1985). 13. J. ROUAULT, M. TOURASSE and D. GEITHOFF, Science and Techn. of Fast Reactor Safety, Guernsey, BNES (1986). 14. T.NAMEKAWA, T.MITSUGI, T. TACHIBANA and S. YAMANOUCHI, 35th Conf on remote techn. (1987). 15. F. SCHMITZ, G. DEAN, M. HOUSSEAU, F. de KEROULAS o.nd J.C. VAN CRAEYNEST, Conf on Fast Reactor Fuel and Fuel Elements, Karlsruhe (1970). 16. H.E. SCHMIDT, TUI Progress Report 11 (1971). 17. S.M. HARTLIB, A. HOUGH, M.P. WAITE and A.R. HALL, AERE, R 7325 (1973). 18. M. de FPANCO and J.P. GATESOUPE, Plutonium and other actinides, 133 (1976). 19. R.G.J. BALL, AERE-R 13395 (1989). 20. R.E. WOODLEY, J. Nucl. Mat., 74, 290 (1978). 21. Hj. MATZKE, J. OTTAVIANI, D. PELLOTTIERO and J. ROUAULT, J. Nucl. Mat., 160, 142 (1988). 22. M. TROTABAS, BNES meet, Grange-over-Sands, 165 (1980). 23. U. BENEDICT, M. COQUERELLE, J.DE BUEGER and C. DUFOUR, J. Nucl. Mat., 45, 217 (1972). IT - 02.1 iboo M) 2800 T(K) FTGUHE Thermal capacity of UO , derived from enthalpy measurements O V»Nt»»n»S • I.IBBV ZU « HM1D1 • WF.ll BACHtR # HET7.LER A FIKI'SIIIMA i^".'. .'. FIGURE 2 -jj = ?.00 - Thermal rondur t.i v i ty versus temperature IT - 02.9 PLUTONIUM TECHNOLOGY IN GERMANY HISTORY AND PrtESENT STATUS Wolfgang Stoll Ret. Gen . Mgr . Allcem, Hanau Ameliastr.25, D-645o Hanau-1 SUMMARY Plutonium Technology started in Germany 1964 from Lab scale by fabricating units containing ,(U,Pu)0o for SNEAK-cri tical assembly in Karlsruhe. Until now a variety of MOX- and FBR-fuel elements have been produced in the Hanau-plant, containing altogether more than 5 tons Pu. KEY-WORDS: Automation,FBR-fuelelements,MOX- fuelelements,Technology,Garmany INTRODUCTION As a consequence of the Fast Breeder Program started in Germany back in 196o it was decided, that Plutonium- fuel fabrication technology was essential to supply fuel for future German nuclear power plants.Alkem was formed by private industry to operate first a Plutonium Laboratory and afterwards fabrication facilities located at the Karlsruhe Nuclear Research Center.Fabrication started with squared steel boxes containing (U,Pu)O_-pellets to be used for the fast critical facility SNEAK. After chemical industry and reactor vendors had agreed to pool their nuclear activities, it was decided to collocate U0_ and CU,Pu)O? fuel fabrication facilities at the Hanau industrial site near NUKEM in 1971. Starting material for MOX- fuel fabrication was both Plutonium nitrate and oxide, coming from reprocessing plants. TECHNOLOGY DEVELOPMENT Like other facilities Alkem started from laboratory scale with 3 lineup of fabrication steps in Interconnected gloveboxes. Initially most operations werde manually done by trained and skilful operators.Initial Plutonium had less than 1o4 higher isotopes and could be fabricated Into fuel within less than 3 years after separation. A first major gamma dose increase was found in Plutonium separated from high burnup LWR-fuel, which had to be fabricated into FBR- fuel containing more than 254 Pu.Ever increasing throughput required and more restricive dose limitations for operators led to an increasing number of fabrication steps to be mechanized and finally automized.Whlle this was easiest in Bteps, where one had to deal with flowable matter, like in chemical conversion from nitrate to oxide, or with discrete geometries, like sintered pellets and cladding tubes, it would have been needed from exposure considerations most in thse steps, where loose powder is handeled.Technically those steps however are the most difficult ones to be reliably automated. IT - 03.1 In the late 7o-es the acid-solubility requirements for MOX fuel were reinforced by reproceeeors.So processes had to be developed, which did allow formation of soluble mixed .cristals during sintering by interdiffusion.For mixed powders this was achieved by an intensive milling operation, whereas in conversion coprecipitation of hexavalent nitrates, containing at least 60* Uranium via the Ammonia-carbonato-complexes did result in very good solubility. Meanwhile both processes have been automated to a degree, where only repair and maintenance requires hands-on operation. Sintering and pellet grinding had been mechanized already in the early 7o-es, whereas drying of pellets and rod-filling was automized since 1973. More severe licensing requirements coming up in the 80-es led to a new building structure, first for Plutonium storage, later for the whole fabrication plant, which has now to withstand high externa' forces, as they are connected with earthquake, planecrash explosion and fire. In accordance with these new buildings had to be constructed and built, which are soon about to be ready for ongoing fabrication.They will contain two new highly automated fabrication lines with an annual capacity of 5o to 60 tons of MOX-LWR-fuel per year each.As soon as this plant is operable, the elder plant, fabricating now around 35 tons/year, will be shut down and decontaminated. In spite of the fact, that the alpha-tight encasings in the new plant look like gloveboxes, thay are double-walled for neutron-arid gammashieiding. Their gloves will only be used for repair and maintenance. The new plant allows fabrication from aged oxide containing up to 3% Pu-238 and still keeps the maximum annual operators exposure below 1o raSv. Up to now the ,existing facilities have supplied fuel to 16. nuclear power stations containing alltogether more than 5 tons of fissile Plutonium. IT - 03.2 SOLID STATE CHEMISTRY OF URANIUM PLUTONIUM MIXED OXIDE Keiji Naito Department of Nuclear Engineering, Nagoya University Furo-cho, Chikusa-ku, Nagoya 464 01, Nagoya SUMMARY-The defect structures and related properties of uranium-plutonium mixed oxide, (U,Pu)O2±x» at ni9h temperatures are reviewed mainly based on the results by the present authors by means of electrical conductivity measurement and thermogravimetry, in comparison with doped U (Key words:Uranium-plutonium mixed oxide, Doped UOo? Defect structure, Electric conductivity, Thermogravivuetry) I. INTRODUCTION Uranium-plutonium mixed oxide is promised to be used as> a fuel in modern thermal fission reactors and in fast reactors. The mixed oxide, as well as uranium and plutonium oxides, are characterised by their nonstoichiomety, due to which their properties tend to alter significantly. Since the nonstoichiometry is closely related to the defect structure, in this paper, the defect structure and the related properties of uranium-plutonium mixed oxide at high temperatures are reviewed mainly based on the results by the authors, in comparison with doped UO2±X. II. OXYGEN POTENTIAL The oxygen potential is a key parameter in predicting the oxidation-reduction behavior of the oxide at high temperatures. The oxygen T=1273K potential of uranium- AU]-yPuy)0 .x plutonium oxide indicates 2 some tendencies with the variation of plutonium content : The oxygen potential increases with increasing plutonium content in both hypo- and hyper- stoichiometric regions, which is explainable,to a first approximation, by the so-called valence control rule as well as other doped U02±x' -10 -5 log(P / Pa) III. DEFECT STRUCTURE O2 The defect structures of Eig. 1 Oxygen partial pressure the oxides can be discussed dependences of electrical based on the oxygen partial u PuuP conductivity of ( 1-y( y^°2-'^ x pressure dependences of the and PuO departure from stoichiometric 2±x. IT - 04.1 composition and the electrical conductivity. Systematic measurement on the electrical conductivity of the mixed oxides, (Un_yPUy)O2+x» ky the authors /I/ TFig. 1) enable us to conclude that the predomi- nant defect in hyper- stoichiometric (Ui-yPUy)02+x is thought to be a multiply ionized complex defect /2/, of which ionization number increases with increasing plutonium content. The defect structure in nearly stoichio- metric region is known only as Fig. The variation of the electric conductivity of a neutral defect from the in electrical conductivity data. (U1_yPuy)02+x near The electrical conductivity of stoichiometric region the mixed oxide in this region against plutonium content. varies with plutonium content as shown in Fig. 2. This behavior can be explained by considering the electronic hopping between plutonium ions (Pu3+_Pu4+) in addition to the hole hopping between uranium ions (U4+-IF+) . The predominant defects in hypostoichiometric (Ui_yPuy)02-x is thought to be an ionized oxygen vacancy or its cluster. Arrhenius plots of the electrical conductivity for some mixed oxides /I/ are shown in Fig. 3. The activation energies obtained were classi- fied into two groups : One is lower activation energy with larger deviation from stoichio- metry, and the other is higher activation energy with smaller Fig. 3 Temperature dependences of deviation. The electric the electrical conductivity conduction in the former case P 3nd U can be explained by assuming of (Ui-y V°2+x °' the hole hopping mechanism 4+ 5+ between U and U sites in (U,Pu)O2+x» similarly to thecase of U02+x- Tne latter may be explained by considering the hopping between Pu3+ and Pu4+ in addition to that between U4+ and U5+. IV. REFERENCES 1. K. Naito, T. Tsuji, T. Fujino and T. Yamashita, J. Nucl. Mater. 169, 329 (1989) 2. T7~"Matsui and K. Naito, J. Nucl. Mater. 1_3_2, 212 (1985) IT - 04.2 RESEARCH ON CHEMICAL PROPERTIES OF TRANS-ACTINIDE ELEMENTS Zvara I. IT - 05 STUDIES ON PLUTONIUM AI^'.> NEPTUNIUM CHEMISTRY AT CADARACHE. BEAUVYM., LARROQUEJ. CEA, DRN, DEC, SPU, Centre d'E'tudea Nuc!6aires de Cadarache, 13108 St Paul lez Durance, Cedex, France. Abstract After an historical review on the chemistry of the transuranium elements developed in our laboratory since 1949, we present briefly the chemical field of our fundamental research on the 5f elements during the last five years. Mom broad aspect has been discussed for the actinide borides, with focus on the chemical bonding and correlation with atom radii. PLUTONIUM AND NEPTUNIUM HISTORY The neptunium isotope Np^39 jg the first transuranium element artificially synthetized which has been identified (Me Millan and Abelson, 1940 [1]). The plutonium was discovered later in the USA at the end of the same year, in the form of Pu*** by Seaborg, Me Millan, Kennedy and Wahl [2]. Fifty years after, these two actinides which are the eons of the nuclear technology, are available in very large quantities (several tons produced in France for some, isotopes). However, the physical and the chemical properties of this "5F elements remain difficult to predict from their place in the Periodic Chart. The neptunium and the plutonium are on tae boundary between two behaviours, the one of the transition elements (3d), or, for the trsnsplutonium elements, the one of the Ianthanides (40- It in wollknown now that these two behavioum can be associated, with the electronic structure of the atoms, respectively the delocalization and the localization of the 6f electrons, but the conditions necessary for this delocalization or this localisation are not well established. Therefore, the properties of the plutonium or neptunium compounds cannot be extrapolated e nd they must be measured. IT - 06.1 The first isolation of plutonium in France had been done in 1949. The preparation of metal on the gramme scale had been realized in 1956, in our laboratory in Fonienay-aux Roses, by Anselin, Faugeras and Grison. This wide scale production of plutonium metal was the first reported in the literature [3], After this work, many studies on plutonium metal and plutonium fuels for the Fast Breeder Reactors, had been undertaken in CJT laboratory. We remember the investigations on the allotropic modifications of plutonium described by Pascard in 1960 [4], and during this period of the beginning of the nuclear age, many investigations on the different potential nuclear fuels (metal, oxides, carbides) have been developed before to choice the material for the reactor Rapsodie [5]. This experimental Light Metal Fast Breeder Reactor (40MW) had run with the mixed oxide fuel (U,Pu)O2, from 1967 until 1982. Enriched from this first experience, our laboratory h«is been involved in the engineering of the core of Ihe L. M. F. B. R. Phenix (563MW). This second reactor has generated electricity distributed by the french electrical network since 1974. Now, the laboratory has been removed from Fontenay to the neighbourhood of the plutonium fuel factory, in Cadarache. Fuels for the futur are under investigation: the fuel of the L.M.F.B.R Super-Phenix, and, the MOX fuel for the P. W. R. It is clear to day, than the progress in the chemistry of the transuranium elements is intimately involved in the developing nuclear technology. But in addition, the discovery of unusual chemical and physical properties with some actinide compounds has also generated important demands for new compounds. That is particularly evident after the discovery of the heavy fermions AnBe^3 and AnPtjj, in 1983 [6, 7] (An is an actinide), and since 1985, the research on the physics of the actinides has been concentrated on the heavy fermion systems. The elaboration of these "exotic" materials for accurate physical measurements on well characterized samples, and the wide size single crystals absolutely necessary for some measurements, involve important studies in the actinide chemistry. During tht last five years, our investigations on the solid state chemistry of plutonium and neptunium can be classified into two categories : - first, synthesis and characterization of the potential heavy fermions (essentially intermetallic compounds), and, studies on the chemical bonding and the localization of the 5f electrons, for the fundamental research side; - secondary, studies on the valence stability of the actinides, studies on non-stoichiometry in the compounds, and, studies on defect structure in solid solutions, for the interpretation of the behaviour of the materials in ihe fuel cycle (reprocessing, nuclear waste, but also fuel properties). IT - 06.2 Today, we focus this presentation on the first category of investigations in chemistry developed in our laboratory. The intermetallic compounds synthetized in our laboratory for the heavy fermion program have been described elsewhere (Table 1, [8]). Our studies on the actinide borides will be discussed now, in respect with the previous classification, i.e. the chemical bounding and the 5f electron localization. STUDIES ON CHEMICAL BONDING IN THE ACTINIDE BORIDES One of the main objective in the study of the so'id state, is the understanding of the electronic energy states and the combinations of them which lead to chemical bonding. The filling of the 5f shell in the actinides with the possible delocalization of these electrons, particularly in neptunium or plutonium compounds,is a subject of most interest to chemists. Our investigations on the actinide borides can be included in this category. The borides are one of the most important types of refractory compounds widely used in various ingineering fields, and rough examination of the metals able to form borides shows that in general their size rather than their electronic configuration is the principal factor governing the boride structure [9]. In the very wide variety of compositions observed for the binary borides MXBV (B/M from 1/16 to "=100, see Table 2), most of the compositions of the transition element borides are with B/M ^2. On the other hand, the compositions of the lanthanide borides are in the range of B/M ^2 (except For the actinides, which have atomic radii ( =0.15nm) intermediate between the one of the transition elements and the one of the lanthanides, the size of the atom becomes apparently secondary, and they form many families of borides where the chemical bonding is an important parameter (Table 2). These actinide borides all contain either two or three dimensional networks of covalent B~B linkages of from 0.172 to 0.183nm (approximately two times the boron radii 0.088nm), and, the metal atoms are contained between the layers of boron or within the networks because the covalent framework of boron atoms seems to be more rigid. Secondary, all boron sublattices are electrons deficient and require electron transfert from the actinides (sp^ bonding). Therefore, the study of the actinide borides brinsra very useful informations in the understanding of the electronic structure of the 5f elements. The transuranium borides synthetized in our laboratory are described in Table 3. The high purity level necessary for the accurate crystal structure measurements on the compounds, has been systematically quantified, and, it had been obtained from the elaboration process previously described [10-13]. IT - 06.3 In the hexagonal diboride structure (AIB2 type, space group P6/mmm nl91) each boron atom is bounded to three others (sp^ type with B-B distance=a/v'3) to form triangular nets which alternate with similar layers of actinide atoms, and each actinide atom has 12 B neighbours (distance An-An=a). The synthesis and the study of the physical properties of the mixed diborides (Npi_yFUy)B2, with y from 0 to 1, had been done in our laboratory, in collaboration with R CHIPAUX [ll-12].The variations of the crystal cell parameters and the volume of the cell described on Figures 1 and 2, show the small effect of the atom radii. The increase in the An- An distance from U to Pu is against the actinide radii (cations or metal). However, this actinide bonding remains in good agreement with the structure of the lanthanide or transition metal diborides. The distances between the layers of actinide atoms (c) are not correlated with the other diborides by the atomic radii (comparison limited to the transition metals and the lanthanides) : they are systematically higher (~ 15 %). This distance decreasing from U to Pu seems to be independent of the actinide radii, whatever the oxidation state considered. Nevertheless, the variation of the other crystal cell parameter "a" appears to be dominant for the variation of the volume of the cell, and the electron transfert possibility between actinide and boron must be considered for the strenght of the chemical bonding. In the actinide diboride serie, the behaviour seems to be the one of the lanthanide borides from the uranium. The structure of the tetraborides (tetragonal "ThB^ type, space group P4/mbm nl27) contains boron basic units sp2 type and Bg type. Each actinide atom has 18 B atoms around it. The figures 3 and 4 present the variations of the crystal cell parameters and the volume cell. These results are fully compatible with the structure of the lanthanide tetraborides (correlated with the metal radii). The distance between the B6 layers (c) increases from U to Pu, with a jump between Np and Pu. The An-An distance also increases, but lightly. Therefore, the boron distance sp^ must significantly increase from NpB4 to PUB4. Including the thorium tetraboride crystallographic data, the behaviour of the actinide tetraborides appears to be close to the variation of the actinide radii (metalllic state). However, like with the diborides, the modification of the 5f electrons behaviour occur from the uranium. We have recorded small deviations for the cell parameters of "UB4", depending of the conditions for the synthesis (Table 4 [10]). That can be attributed to the non-stoichiometry possible in the borides (Uj_xB4 proposed), but, more investigations are necessary before to conclude on the non-stoichiometry in transuranium borides. IT - 06.4 Only one monoboride, PuB, has been proposed from our sy ithesis (cubic "NaCl" type, space group Fm3m). The crystallographic similitude between this monoboride and the mononitride had incited some authors to reject the existence of this phase, but we are confident in our result on cubic boride, and the composition could be PuBj_xNx. This phase, apparently the monoboride, have not been observed with neptunium. The An-An distance and the B-B distance would be the most important comparatively with the other borides studied (Pu- Pu=0.490nm). However, the actinide monoboride has never been a single phase in the material after the synthesis, and it is necessary to chock than that is not due to the impurities present in the materials, during the elaboration. Actinide to boron bond lengths in the binary compounds vary considerably from one structure type to another, even for the same metal (Np or Pu). The environment of the actinide during the elaboration, usually at high temperature, may be change the structure of the 5f shell, and the equilibrium at low temperature remains difficult to reach when the chemical bondings are very strong. The oxidation states of the actinides in these borides aie the result of a competition between the electrochemical environment and the transfert of electrons necessary to stabilize the boron chains in the structure. The last eerie of borides (ternary compounds of Np or Pu) has been studied in collaboration with P. ROGL [13]. The crystal cell parameters are reported in Table 3, and the variation of the volume of the cell of AnBC and A11B2C is presented on the Figure 5. The structure of NpBC is orthorhombic and isotypic with UBC (space group Cmcm). The structure of NpB2C is "ThB2C"type, for the high temperature form (space group R3m). The complete resolution of these structures would be very useful for the bonding chemistry because the introduction of carbon in the boride can bring a new behaviour for the compound (see EuBjCy). Unfortunatly, the determination of the interatomic distances remains difficult In these ternary compounds where the non-stoichiometry can be dominant, and we expect to propose soon results correlated with the atom radii. CONCLUSION The chemical bonding, and particularly the An-B distance, seems to be more important than the radii of the actinide for the crystal structure of the actinide diborides. The localization of the 5f electrons appears to be dominant from the uranium. IT - 06.5 The variation of the An-An distance from U to Pu in the actinide tetraborides confirms the small influence of the radii of the actinide. However, the metallic radii of the actinides roust be considered for this serie of boridee The existence of c ubic actinide boride phase, binary (or mixed) monoboride type, has been confirmed exclusively with the plutonium. This exciting field for the fundamental research on the actinides with the studies on the borides, must be developed to progress in the knowledge of the 5f elements. These investigations developed in parallel with the chemical studies on the behaviour of the nuclear fuels, participate in the increase of the knowledge necessary for the fuel cycle management. More, the results of this work bring some ideas for applications, for intance in nuclear waste management. REFERENCES [ 1] Me MILLAN E. M., ABELSON P., Phys. Rev., 57, (1940), 1185. [ 2] SEABORG G. T.,Mc MILLAN E. M., KENNEDY J.W., WAHL A. C, Phys. Rev. 69, (1940), 366. t 3] ANSEUN F. .FAUGERAS P., GRISON E., Compt. Rend., 242, (1956), 1996 [ 4] FASCARD R, Plutonium 1960, ed. Grison E., Lord W., Fowler R, Cleaver Hume Press, London,(1960), 16. [ 5] BUSSY P., Plutonium ldGO.ed. Grison E., Lord W., Fowler R, Cleaver Hume Press, London, (1960),589. [ 6] STEWART G. R, FISK Z., WILLIS J. O..SMITH J.L.,Phys. Rev. Lett., 52, (1983), 679. [ 7] OTT H. R., RUDIGIER H. ,FISK Z., SMITH J.L., Phys. Rev. Lett.,50, (1983),1585. [ 8] LARROQUE J., BEAUVY M., ACTINIDES 89, Tachkent, USSR, (Sept. 1989), 146. [ 9] ETOURNEAU J., J. of the Less Common Metals, 110, (1985),267. [10] BEAUVY M. LARROQUE J., NABON C, 14 Journees Actinides, Davos, Suisse, (1984). [11] CHIPAUX R .These doctorat, Universite Aix-Marseille 2, (1987). [12] CHIPAUX R, LARROQUE J..BEAUVY M.,J. of the Less Common Metals, 153, (1989), 1. [13] ROGL P., BEAUVY M., LARROQUE J.,19 Journees Actinides, Trento, Italy, (1989), 95. IT - 06.6 Cell parameter a(nm) Secondary stability aspect AnM, phases in air at room temperature this work published data USti-j 0,4606 * 2.10 I 0,4609 cSARI. F VEHN«ZA.WMUU.EPIIM3I Sn quickly oxidized ^UO2 + Sn +... 4 UPbo 0,4784 ± 3.10" j 0,4791 Pb ductile pyrophoric (UO2 + Pb + ... .-4 i none Nplr,3 0.4608 - 1.10 In + Npln2 quickly oxidized iNpO, + In +...) 0,4627 AW MfTCHEu. D.J LAM (1974' NpSn3 0.4628 - 3.10^ Sn quickly oxidized (NpO2 + Sn +...) 0.4622 J GAL 2 HADARI (1B73) grey i 4 none H NPTI3 0,4699 r l.io" : Tl .NPTIX extremely pyrophoric (NpO2 + Tl 4 J NpPb3 0,4813 ^ 2.10" none l b porous • pyrophoric (NpO2 + Pb +..) n Atno i tn-4 ! 0,4607 BOCHVARKONOSEEVSKV.KKUBAfTSEU V 0.4608 3; 1.10 I l\ AHlt\ MENSHIKDVA. CMEBOTAREV IIB58|l ) In quickly oxidized (PuO2 + In +...) U,*»O1W F. H. ELJNQEB. C. C. LAND MB965: nodulus 0,4808 i 1.10 | 0,4807 WOO0.CRAMER.W*u>0E.RAMSEYIiSe9) Pb pyrophoric (PuO2 + Pb +...) TABLE I - Intermetallic compounds (heavy fermions program) TABLE II - Binary borides UB4 BejB B12C3 (Li86) Be«B BUC2 Be2B Be86 (BjSi) HgB4 (AIB4) B12SI3 AIB10 H9B12 A1BJ2 t, (C»B2 ) ScB2 (TiB2) V?8 Cr4B Mn»B Fe2B C03B N13B (CUB22 ) /" Gm Ge C»B4 ScB4 TIB V3B2 Cf2B Mn2B FeB Co2B Mi28 CiB6 ScB« TIJB4 VB Cr5Bj HnB CoB NljBj ScBi; T1B2 V3B4 CrjB2 80384 MI4B3 V2B3 CrB Mf>3B2 NIB VB2 CrjBi MnB2 (Ni2B3) CrB; MnB4 NIB; CrB4 NIB12 RbBjs) StBf, YB2 ZrB (Hb2B) Mo2B Ic38 RU7B3 Rh7B} Pd3B AqB7 Cd In Sn YB4 2rB2 NbjB2 M03B2 TC7B3 RunBg (Rh2B) Pd2B5 Y B6 2rB6 NfcB HoB TcB RUBI.I RhBi.i (Pd2B) VBl2 ZiB12 Nb3B« H0B2 TC3B4 Ru2B3 (C.B6) B«B6 L.2B HfB Tr»2B W28 (R04B) OsB irBi.i Pt3B (AoB2) Hg XI Pb [LaB2l HfB2 Ta3B2 we Re3B OaB2 (IrB2) Pt2B LaB4 TaB Re7B3 Os2B5 Pt382 1 Ll8£ Ta3B4 W2Bj (Re3B2) PtB [•-•Bl2) TaB2 HB4 RsB WB12 (Re2B3i1 ReB2 f Rd'tAc ) ReB3 r.B4 PrB4 Nd2B» Pm 5-B2 IEUB4) GdB2 TbB2 OyB2 H0B2 6rB2 Tma2 YbB4 LuB2 CeB6 PrB6 NdB4 S«285 EuB« Gd2B3 Tb&4 H0B4 ErB4 T.B4 (YbBfi11 LuB« NdB6 S«B4 CdB4 Tb66 DyB6 [HoB6] lErB6) l'»B6] YbBi2 [LuB6 S«B6 GdB6 TbBl2 OyBi 7 H0B12 ErBi2 I»B12 YbB&6 LUB12 S«B66 tzii TbB66 DyB*6 HcB66 ErB66 r.B66 LuB66 ThB4 Pa UB2 NpB2 (PuB) AnB4 Cm Bk Cf Is Fm Md Ho Lr ThBs UB4 NPB4 Pufl2 AmB6 IhB] 7 UB12 NpB6 PuB4 ThB18 NPB12 PuB& PuB 12 IT - U6.8 "1 a, nm b, nm c, nm cell voli B-B, ntn An-An,nm 10-30, 1 "PuB" cubic 0.4905 (4) - 118.0 0.4905 0.490 NpB2 hexagonal 0.31623 (3) 0.39719 206.4 0.5477 0.318 PuB2 hexagonal 0.31862 (2) 0.39491 208 3 0.5518 0318 UB4+X tetragonal 0.70773 - 0.39816 198.9 0.3538 NpB4 tetragonal 0.70884 (2) 0.3996 (3) 200.5 0.3544 PuB4 tetragonal 0.712 (5) 0.410(6) 207.8 03560 UBC orthorhombic 0.3591 1.195 0.3372 (143) NpBC orthorhombic 0.35856 (8) 1.20007 (8) 0.33831 (6) (145) NpB2C "ThB2 C" type 0.65362 (30) - 07522(64) 357,3-(K Table III : Crystallographio results for Actinide Horides Synthesis Process a, nm c, nm of UB4 UO2 + B4C 0.7075 0.3974 U02 ^ 2B4C 0.7075 03974 UO2 * B4C > C 0.7075 0.3974 UO2 * B4C *- B 0.7085 0.3981 Tafcjg IV : Non stoichiometry in tetraboride Uj_xB4: crystailographic results IT - 06.9 10,-1 nm nm 1 • * i .400 C A V70 AnB2 \ = .398 > \ - .396 \ • H 1.50 nm • .318 a // .316 • .314 Uft / d 1 1 t i , 1 Th Pa U Np Pu (Pa) U Np FIGURE 1 - Actinide radii and crystallographic data for hexagonal diborides 10 4> 2 •H T3 IS § m >> u o 4> o o i •o •o •H c o < I CM o - 06.11 nm .710 - Pu Th ?Q U Np PU FIGURE 3 - Actinide radii and crystallographic data for tetragonal tetraborides 1Qrao_3 10*nm J nm i.4O 4 UO M 1-20 H 1.50 ! o 205 200 h 195 h Th Th PQ U Np FTCURE 4 - Actinide radii and volume of the crystal eelltetragonal tetraborido* (0 •p (0 >> o XI O s in _ c; (0 •p a. o og E o o 'o ' i 36. 1H ACHIEVEMENTS IN CHEMISTRY OF ACTINIDE ELEMENTS IN SOVIET UNION Myasaedov B.F. IT - 07 RECENT ADVANCES IN THE THERMODYNAMICS OF TRIVALENT ACTINIDES F. DAVID Instilut dc Physique Nuclcairc Bat. 100, 91406 Orsay, France Abstract: The structure of the trivalent lanthanidc and actinidc aqua ions has been investigated through radiochemical methods. It is observed that the size of the aqua ions are larger for aclinidcs than for lanthanides with the same crystallographic radii Rc. Hydration numbers N and H corresponding to the primary and second sphere have been deduced. In both series N is varying discontinuously with Rc, and the discontinuity corresponds to a greater radius for actinides than for lanthanidcs. These two differences have been related to shorter cation-water distance and covalcncy in ihc actinidc series. These informations have been used to improve our previous hydration model. Gibbs energies arc calculated for spheric monovalent, divalent and trivalent ions. Deviations between experimental and calculated data are less than 0.6%. The model allows evaluation of the covalency effect in the actinide series. We also propose a model to evaluate entropies of Irivalcnl lanthanidc and actinide aquo ions. INTRODUCTION Though trivalent ions are the most frequently occuring actinidc species, their ihcrmodynamic pro- perties are mostly unknown if one considers the whole scries. Formation enthalpies have been delerrmncu for trivalcnt U - Cf ions (1) through calorimetric measurements but entropies of aquo ions and hyJralion enthalpies are uncertain. Plutonium entropy (2) and hydration enthalpy (3) have only been published, based on partially experimental data. Different models have been proposed to estimate entropies (4, 5) and hydration enthalpies (6, 7, 8, 9, 10). Bui to get satisfactory data it would be necessary to take into account accurate knowledge of the structure of the aquo ion. This is governed by the coordination number N and the dynamic hyJ.ration number H corresponding to the primary and outer hydration sphere. Very few studies have been reported on the structure of the actinidc aqua ions. That insufficiency is mainly related to the difficulty to manipulate, solutions of wcighablc amounts of radioactive aclinides and the instability of trivalent uranium, neptunium and plutonium in aerated acid solutions. Up to now no direct measurement of the coordination number N has been achieved. Recently, analysis of the absorption spectra of americium solutions lead Carnall (II) to the conclusion that Am3+ aquo ion is coordinated with 9 water molecule in the primary sphere. The situation with regard to the coordination number N in the lanthanidc series has shown in the past conflicting results of different authors. They have been analyzed recently (12). However, it seems now established that coordination number N is changing from 9 to 8 from the beginning to the end of the lanthanide series. N should vary with the crystaltographic radius Rc as a S shape curve with a discontinuity around europium ion. IT - 08.1 This is indicated by X rays diffraction (13), EXAFS (14), absorption spectroscopy (15, 16, 17) but recent neuiror. diffraction studies (18, 19,20, 12) are the most convincing proof. Tnese discontinuous variations of N vs. Rc in the 4f series are consistent with variations with Re of different chemical properties like equivalent conductivity of chloride solutions (21), molar partial volume (22) or entropies (23,24). STRUCTURE OF THE TRIVAt.ENT ACT1NIDK AOUO IONS Since physical methods are difficult or impossible to apply to actinide elements, we have chosen to study transport properties essentially by radiochemical methods. These properties are related to the size of the ion in solution, therefore they are function of the hydration number N and H. Moreover, they are measurable with elements at extremly low concentration. Tfc limiting diffusion coefficient D° (for ionic strength I = 0) (25) and ionic mobility u° (I = 0) (26) could be measured radiochemicaly with a high accuracy relatively to an element used as reference. Then, it is possible to determine absolute D° and u° values by measuring limiting equivalent conductivities X" of lanthanides. We have the well known relations : D° kT —j = ;— ( Nernsl-Einstein equation ) and u° = X°/F. These quantities are related to the size of the ion in solution. We can define a Stokes radius Rs by the Stokes equation : D _ 0.820 z K_ — X°\x° i. being the charge of the cation and n' viscosity of the medium. However, this Stokes radius has not to be taken as the radius of the aquo ion R since we have to take into account electrostriction phenomena due (o the intense electric field existing in the vicinity of the cation. In accordance with Nightingale data (27) we propose a polynomial relation relating R and Rs when R < 5 A : R = 2.613 + 0.636 Rs - 0.808 R^ + 0.00107 R^ Diffusion coefficient have been measured for cerium, <.. ropium, gadolinium, terbium, thulium, ytterbium and americium, curium, californium and einsteinium with LiCl - HO medium at 25°C (25). Recently accurate relative mobilities have been measured radiochemicaly with cerium as reference for neodynium, europium, gadolinium, terbium, ytterbium (26) and americium and californium (28) in similar experimental conditions. Finally, limiting equivalent conductivities of La CI3, Eu CI3, Gd CI3, Lu CI3 (29) and Dy CI3 (30) were also determined recently, at pH = 5, to prevent hydrolysis of the cation (31). Based on these sets of data, we can calculate an average mobility for cliiiicni M relatively to cerium. The variations, in % of K = u°(M)/u°(Cc) are reported on fig. I, with estimated error bars, versus the crystallographic radius Rc with coordination number N = 8 (32). More rigorously we should retain the value of Rc f«r (lie roal coordination number of the cation (since Rc is a function of N). However, since N is not yet denned for the two f scries, we will use Rc (N = 8) as a parameter allowing a comparison between 4f and 5f elemenis. IT - 08.2 Nd Ce La 92 1.1 1.2 Rc (N=8) A Fig. 1 : Variations of K in % (ratio of the mobility :f element M vs cerium used as reference) vs Re the crystallographic radius will) coordination number N = 8 - The observed variations in fig. 1 calls for several remarks. • First, transport properties ratio arc only varying from 100 to 94 but the accurate determinations indicate clearly the S shape curve previously reported in the case of different physico-chemical lanthanidc investigations. Foi highest ionic mobilities (smallest aquo radius) it corresponds the highest value N - 9. On i he contrary the lowest mobilities correspond to N = 8. This could be explained by increasing charge density Dl at the primary sphere (radius Rl) when the crystallographic radius and primary hydration sphere have the smallest radii. This charge density induces more water molecules interacting electrostatically in ihc second hydration shell. IT - 08.3 The facl that the aqua ions mobility variations arc strictly correlated with those of the hydration number N point out Ihe following conclusion. Considering i!ic excellent accuracy of the K radiochemical determination (0,5% for lanthanidcs), better than the N physical determinations, we have chosen to relate the two observed mobility plateaux to coordination numbers 9 and 8. Therefore, ihe intermediate values of lhe mobilities (for elements samarium to terbium) are supposed to correspond to intermediate values of N. The data on four actinides viz. Am. Cm, Cf and Es arc now obtained with higher accuracy than in the past (33, 34). They show thai the actinide curve K vs R_. should also vary as a S shape curve, similarly as the lanthanide curve, with an inflexion point situated in the 5f series between Bk and Cf at a radius Rc (N=8) = 1.067 A. The corresponding fl electrons is q = 6.8. For lanthanidcs the inflexion point is characterized by Rc (N = 8) = 1.067 A and q = 8.8. The existence of plateaux in the lanthanide curve indicates M for significant variations of Rc we have no variations of N. Therefore, the associations of N = 9 or 8 v.ater molecules are statistically more probable. Since an equilibrium has been stated (35): 3+ 3+ [Cc(H20)j = [Cc(H2O)gj + HjO It means that the plateaux arc corresponding to ihc left and right side of tlic reaction. We deduce that for actinides, similar equilibrium and plateaux should also occur. The strong simiarity of K values for lanthanide and actinide has led us (7) to conclude that we have 9 water molecules in ihc primary sphere at the beginning of the 5f series and 8 at the end. A recent spectroscopic study on americium (11) has given Ihe coordination number 9 for that element, confirming our hypothesis. Therefore, we have again determined N values for the aclinidc scries by the same way we have s'atcd those of lanthanidcs. Bui die aclinidc experimental data show two interesting differences For a given Rc value the actinides considered are less mobile than lanihanides. At the plateaux the diminution is about 1.5% of K. This is the case for insiancc for amcricium and neodynium. It means that the size of the aquo ions is 1.5% greater for actinides than for the corresponding lamhanides. A second observation is: the inflexion point is shifted for actinides, as compared to lanthanidcs, at greater radius (AR = 0.010 A). These two experimental observations could be understood if one considers that the cation-water distance in the primary sphere is shorter for actinides than for lanihanidcs. The reduction of Qic distance means smaller radius R1 of the primary sphere : Rl (AN) Therefore, the charge density on thai sphere is higher and the number of water molecules interacting in the second hydration shell is increasing in comparison to lanthanidcs. On the other hand ihc smaller the crysiallographic radius the grcatct is Ihc radius of the aqua ions for both laruhanidc and aclinidc scries. However, for the same crystallographic radius the radius of the aqua ions arc greater for aciinidcs than for lantinides. Moreover, when we are describing the K = f (Rc) curve, starting from uranium to heavier clement, the nine water molecules of the primary sphere arc closer from (he cation than those of lanihanides. Their centers are situated on a sphere wilh surface S,. (AN) < Se (l.N). When RL is decreasing, Sc is also decreasing. For stcric reasons the radius Rc corrcspondmy lo the point where one water molecule starts to be eliminate is greater for actinides than for lanihanidcs. finally, shorter cation-water distance is also consistent with ihe shift of the inflexion point of fig. 1, from I.067A (actinides) to 1.057 A (lanthanides). Shorter cation-ligand distance has been often related to covalcncy in the aclinidc scries (36) due lo delocali/ation of 5f orbitals (cxcmple is given fig.4 of rcf. 37). The interesting fact deduced from iranspori IT - 08.4 property studies is that shorter cation-water distance (covalency) is observed even for iransplutonium elements, when 5f orbitals are considered 10 be localized similarly as 4f orbiials. The reported data (fig. 1) allow a precise comparison of the relative mobilities (as well as equivalent conductivities or diffusion coefficient) in the irivalent lanthanide and actinide series. Therefore, it is sufficient to determine the absolute value for any one element to have absolute values for the totality of the elements considered. In order to minimize effects of ionic strength, equivalent conductivities are considered. Experimental data reported by Spedding et al. (21) for most of lanthanides have been corrected (31). This is the case because early measurements were carried out at a pH value 6.4. Today we know that hydrolysis of the trivalcnt cation starts at this pH value (38). Recently new equivalent conductivity measures have been realized on dilute trichlorides of some lanthanidcs at pH 5 (29,30). They give X° = 67 ± 0.5 Q1 cm1 for the beginning of the series (29). Taken into account the K data it should correspond to X° = 63.1 ± 0.5 £}•' cm'1 for llv second plateau (39). This estimation has been confirmed by a recent dctertv. nation of X° (Dy3+) = 62.9± ).7 ii1 cm1 (30). Based on these sets of experimental data we can calculate X°, volume V and radius R of the aqua ions for all elements considered. One question remains essentially to analyze. It is the case of light actinides : U3+, Np3+ and Pu3+. Strong dclocalization of 5f orbitals could induce larger covalency for the ions and, therefore, smaller mobilities and larger ionic radius of the aqua ions More work should be done in this connexion. To obtain a complete picture of the aqua ions we have now to determine the number H of water molecules situated in the sr^ond hydration shell. The available data : N and V are not sufficient since it is necessary to take into account the fact that the volume v of water molecule is a function of the distance x to the cation. This is related to the discontinuous properties of water in the vicinity of ions. High electrical field induces elcctrostriction phenomena which are responsible for compression of the watc molecule volume. We obtain the values v = f (Rc) for the v-' hydration spheres by computation (39): the electrostriction theory developped by Conway (40) is applied. Therefore, we have computed the volume VI of the inner sphere and by subtraction : V2 = V - VI, we gel the volume V2 of the external shell. Finally, the knowledge of die volume v2 of one water molecule in the second sphere give an average number H of water molecules. We have verified that the H water molecules are interacting electrostatically with the charged primary sphere (hydrogen bonds). Thus, if one consider trivalent lanlhanidcs aswell as divalent alkali- earth ions (an analog treatment has been achieved for these divalent ions taking N = 6), we verify that H is essentially varying linearly with Dl, the charge density existing al the surface SI of the primary sphere : H = mDl+n [1] The discussion of trivalent actinidc ions is more complex since the observed H values derived from transport properties arc greater than those predicted by equation {!]. We have to take into account shorter cation-water distance, and therefore, greater charge density Dl, due to covalcncy in the 5f scries. From the equation f 1] we can evaluate the reduction in the cation-water distance and VI. At the same time reduction of the inner sphere leads to a larger shell volume V2 = V - V1, since V is determined experimentally. Therefore, the hydraiion number H is increased. Finally, hydralion number N and H are evaluated for actinides and variations vs Rc are compared with lanthanidcs. IT - 08.5 THERMODYNAMIC CONSEQUENCES The knowledge of the structure of the aqua ions aswell as evaluation of volume vi and V2, radius of water in the primary sphere Rwl, could be used to interpret or evaluate some thermodynamic functions. We will now discuss two applications particularly interesting for radiochemists since practically no experimental informations arc available in She cascof5f series. ). GIBBS HYDRAT1ON ENERGY OFTRIVALENT ACTIN1UE IONS Since the three first ionizalion potentials arc not known for any aciinidc, it is not possible to deduce hydration enthalpy AH (hyd) Mn+, or Gibbs energy AG (hyd) Mn+, through a Born Habcr cycle. Several methods have been proposed to compute hydration enthalpy or free energy of h;dralions. They often use a very simplified model based on a Bom c^aation : 2 AG (hyd) = az /(Rc + 0) /. being the charge of the cation, a and P parameters which have to be adjusted for each scries of ions considered (6, 41, 8). Of course, since thai expression is purely empirical without any theoretical basis, we have lo assume a and [5 constant in a given scries. Moreover, (he question of determination of a and p for covalcm actinides appears to be an hard task. An other approach, proposed by Goldman and Mors? (42) seem more jusiiiicd theoretically , but again a parameter N has to be evaluated. Both methods arc not relevant for divalcm and tclravalent ions. Some years ago (7) we have proposed a general expression for the hydration enthalpy of spheric ions, relevant to charge - 1 to + 4. This semi-empirical expression takes into account Born term, dipolar, quadrupolar and induced dipole interactions and an empiric term was introduced to take into account a second hydration sphere and lhc cavity formation. In spite of further simplifying assumptions in thi.s method it wa>; remarkable that the ^ 'ilculated data for 37 ions of charge - 1 t<- +4, the standard deviations from experimental lo computed values arc smaller than 1% for any scries considered. It is also better than the values obtained from a recent model (43) treating successive shells of hydraiion molecules (deviations are 2.1%). This is why our model was used in the case of divalent, trivalcni and tetravaicnt acfinides. Calculated data deviate from 0.5% from she published AH (hyd) Pu3+ data (3). However, the recent expcrimcnlal data we arc reporting leads us to reconsider the previous treatment. We have lo lake inlo account more realistic data on N and H of the lanlhanidcs and aclinidcs. We must also consider that the radius R\y l of water is not a constant as il was previously slated (44, 7, 43). Il deviates essentially from ihc average radius Rvv = 1.38 A in bulk waicr and is depending on the cation considered and localization. Moreover, in this proposed trealmcnt we will also retain for the cavity formation not a constant as admitted (7,44) bm a funclion of the cavity volume. Thus, we propose Ihc following equation : AGhyd=A+B+C+D+P+K+W with 2 A = aZ /(RN + 2RwD Bom tern-. B = bZN/(Rf,M Rwl)2 dipolar interaction. C = CIZIN/(RN + RwlP quadrupolar interaction. D = d Z2 N/(RN + Rwiy* induce dipole interaction IT - 08.6 P = p(-1 )Z parity term K. = wo*(RN)^ cavity formation W = wj* H second sphere A, B, C, D and P have the same signification as in previous paper (7), K refers to cavity formation and W to the interaction related to the second hydration sphere. This new expression is applied to 26 spheric ions : alkalis, alkali-earths, trivalent lanthanides and yttrium. In these cases, N, H, Re and Rwi are available or computed since accurate conductivity data have been published (45, 29, 30, 39). The experimental Gibbs energies are co-nputed with a Bom Haber cycle using recent published data (46). Now, standard deviations between experimental and calculated data is 0.6% (0.25 for trivalent lanthanides). A second interesting feature of the expression [31 is the possibility to evaluate covalcncy in the actinide series. This is due to the fact that expression [3] take into account realistic value;; of N, H and j+ Rwi- From this expression and fig. 1, it is possible to compute AG (hyd)w.c. M without taking into account covalency if one considers actinides as lanthanide like ions. In this case N, H and Rwi for a given crystallographic radius are those one reads on lanthanide corresponding variations wilh Re. 3+ \n a second calculation, we compute AG (hyd)c M taking into account values N, H and Rwj obtained from experimental data. The difference measure the covalency effect : 3+ 3+ 3+ CE (M ) = AG (hyd)w.c. M - AG (hyd)c M A similar comparison between trivalent yttrium and lanthanides shows that AG (hyd) Y3+ computed with expression [3], using N, H and Rwi deduced from lanthanide variations wilh Rc is identical with the experimental value as deduced from Born Haber cycle using recent (hermodynamic data (47 y. The calculated deviation is only 0.16% which is less than the standard deviation observed in the lanthanide series (0.25%). This means that no covalency is observable in the trivalent lanthanide series. For actinides, the covalency is about - 200 k J moH or is ~ 6% of the total hydralion Gibbs energy, it is interesting to notice that calculated AG (hyd) Pu3+ differs only by less than 1 % from the published value (3). Expression [3], after evaluation of N, H and Rwi make also possible computation of AG (hyd) for divalent and tetravalent actinides but results are more uncertain since covalency effect is difficult to estimate. 2. ENTROPY OFTRiVALENT AQUA ACTINIDES As for the hydration Gibbs energy, the evaluation of entropies were essentially based on estimations. Pu3+ is the only trivalent actinidc ion were partially experimental data are available (2). The models found in the littery lure are empirical and do not lake into account the fact that ihcrmodynamic properties of the equa ions are related to the structure of the species and covalcnl character of actinides. Therefore, without a realistic model, published entropies do not seem as safe data. Entropy of ions S° (Mn+) is an additive properly. It can be written as the sum of three contributions : electronic coni'.guraiion Se, mass term SM and contribution due lo hydralion of the ion Sh: [4] Se and SM are accurately evaluated through classie'expression : IT - 08.7 Se = RLn(2J+ l)and SM = 3/2 R Ln M However, no expressions have been proposed which lakes inio account the structure of aquo ions and covaleni effect. It is clear from the preceeding discussion that Sj, has to be considered with contributions related to primary and second sphere. Thus : Sh = oN + pH [5] o and p being the entropy of one water molecule in the primary and second sphere respectively. Since entropy of trivaleni lanthanidc have been measured for most of the lanthanides we can deduce Sj, fiom equation [4] by using S° (M) data (46). The Sh vs Rc variations are smooth S shape curve (fig. 3 of ref. 7) reflecting the change in the coordination number. The fact that ihe discontinuity in the middle of the f series is more or less marked, depending on the studied property, is related to the weight of paramelers o and p of equations similar as [5). Numerous and accurate data of S° (M3+) (23, 24) and the deduced Sh values give the possibility to compute parameters o and p of expression [5]. We have found o = -13,5 and p = -19,? J mo!"' K"1. With these data we can compute Sj, for trivalcnt lanthanidcs and standard deviations between experimental and calculated data is 1.6%. Therefore, we are able to compute Sh and S° (M3+) for actinide ions through equation [5]. By the way, we take into account covalency effect which is included in the N and H data. In conclusion, we observe lhal the complex physico-chemical properties of actinides is not well established under experimental basis. In ihis work we have shown that raaiochemical and systematic studies of lanthanide and actinide ions allow to precise the structure of ihe trivaleni aqua ions. This knowledge gives the possibility to propose more occurate thermodynamic models and estimates of Gibbs energy of hydration and entropies of the trivalcnt actinide ions. This study is continiung toward better determinations of ionisation potentials in the actinide series. REFERENCES 1. L.R. Morss, in "ihe Chemistry ofActinide Elements" eds J.J. Katz', G.T. Seaborg, L.R. Morss, Vol. 2, p. 1278, 1986. 2. J. Fuger, F.L. Octting, in "The chemical Thermodynamics ofActinide Elements and comj)ounds : Pan. 2. The actinide aqueous ions, I.A.E.A., Vienna, 1976 3. L.R. Morss, J. Phys. Chem. 75, 392 (1971). 4. L.R. Morss, Chem. Rev. 76, 82V (1976). 5. F. David, J Less, Comm Met., 121, 27 (1986). 6. R. Guillaumonl, F. David, Radiochern. Radioanal. Lett., 17, 25(1974). IT - 08.8 7 F. David, B. Fourest and J. Duplessis, J. Nucl. Mater., 130, 273 (1985). 8. S.G. Bratsch, J.J. Lagowski, J. Phys. Chem., 90, 307 (1986). 9. O.L. Keller, Radiochim. Ada, 37,165> (1984). 10. S. Goldman, R.G. Bates, J. Am. Chem Soc. 94, 1476 (1972). 11. W.T. Carnall, J. Less Comm. Metals, 156, 221(1989). 12. C. Cossy. A.C. Barnes, J.E. Enderby, A.E. Mcrbach, J. Chem. Phys. 90, 3254 (1989). 13 A. Habenschuss and F.H Spedding , J. Phys. Chem., 70, 2797 (1979), ibid 70, 3758 (1979), ibid 73 442 (1980). 14. T Yamaguchi, M. Nomura. H. Wakita, H. OlHaki.7. Chem. Phys.. 89, 5153 (1988). 15. K. Rajnak, L. Couture, Chem. Phys. , 55, 331 (1981). 16. L. Couture, K. Rajnak., Chem. Phys. , 85, 315 (1984). 17. L. Couture./. Luminescence. 18/19, 891 (1979). 18. A.H Narten, R.L, Hahn, J. Phys. Chem., 87, 3193 (1983). 19. B.K. Annis, R.L. Hahn, A.H. Narten, J. Phys. Chem., 82, 2086 (1985). 20. A.E. Merbach, First International Conference on f elements, Leuven, sept. 4-7, 1990. 21. F.H. Spedding, P.E. Porter, J.M. Wright, J. Am. Chem. Soc. 74, 2055 (1952). 22. F.H. Spedding, P,F. Cu!icn, A. Habenschuss, J. Phys. Chem.., 78, 1106 (1974). 23 S. L. Bertha, G.R. Choppin, Inorg. Chem., 8, 613 (1969). 24. F.H. Spedding, J.A. Rard, A. Habenschuss, J. Phys. Chem.., 81, 1069 (1977). 25. B. Fourest, J. Duplessis. F. David, J. Less Comm. Met., 92, 17 (1983). 26 B. Fourest, J. Duplessis, F. David, Radiochim. Acta, 46, 131(1989). 27. E.R. Nightingale Jr, J. Phys. Chem., 63, 1381 (1959). 28 B. Fourest, E. Haitier. F. David, Lanthanide and Actintue Research 2, 393 (1988). 29 J. M'Halla, M. Chemla, R. Bury, F. David, J. Phys. Chem., 85, N° 1, 121 (1988). 30. B. Fourest, R. Bury, L.R. Morss, J. M'Halla, V. David, to be published. 31. J. M'Halla and F. David, Bull. Soc. Chim. Fr. , 3-4, 185, 1984. IT - 08.9 32. F. David. J. Less Comm. Met., 121. 27 (1986). 33. B. Fouresl, J. Duplessis, F. David, J. Less Comm. Mel., 92, 17(1983). 34. R. Lundqvist, E.K. Hulct, P.A. Baioden, Ada Chem. Scand., A35, 653 (1981). 35. G. Laurenczy, A.E. Merbach, Helv. Chim. Ada, 71, 1971 (1988). 36 S. Siekierski, /. Radioanal. Nucl. Chem., 122, 279(1988). 37. W.T Carnall, H.M. Crosswhite, Rep. ANL-84-90, (1985). 38. The Hydolysis of Cations, C. F. Baes Jr., R. E. Messmer eds. John Wiley, New York, 1976. 39. F. David, B. Fouresl, //WO- DRE 88-27 (1988). 40. B.E. Conway.in "Ionic Hydration in Chem. and Biophysics". Elsevier, Amsterdam, 1981. 41. W. Briichle, M. Schadel, U.W. Scherer, J.V. Kralz, K.E. Gregorich, D. Lee. M. Nurmia, R.M. Chastelerer, H.L. Hall, R.A. Henderson, D.C. Hoffman, Inorg. Chim. Acta, 146, 267 (1988). 42. S. Goldman, L.R. Morss. Canad. J. Chem., 53, 2695 (1975). 43. K.W. Frese Jr., /. Phys. Chem., 93, 5911 (1989). 44. J. O'M Bockris, A.K.N. Rcddy, Modern Electrochemistry, Vol. 1, Plenum, New-York. 1973. 45. R.A. Robinson and R.H. Stokes, in "Electrolyte Solutions", BuUerworth, London, 1970. 46 A.J. Bard, J. Jordan, R. Parsons (eds). Standard Potentials in Aqueous Solu'.ion, Marcel Dekker, 1985. IT - 08.10 THERMODYNAMICS OF ACTINIDES Scad D.D. IT - 09 NUCLEAR PROPERTIES OF PLUTONIUM AND TEEI* IHPORTAHCE IN NUCLEAR TECHNOLOGY Satya Prakash, Radiocheaistry Division Bhabha Atomic Research Centre,Trombay, Bombay 400 085. 1.INTRODUCTION: Plutonium is the only man made element which is now available in tons: quantity and is formed as by product in kilograms in thermal nuclear reactois fuelled by uranium. This together Kith the fact that plutonium ( major isotope Pu-239) is fissionable with neutrons of all energies endows it an important and unique position in nuclear power proyram. Plutonium has very important role to play in the future energy scenario of the world. The present day nuclear reactors (thermal) derivo their energy from the fission of U-235 which is only a minor fraction (0.72%) of the natural uranium;rest being U-238, not fissionable by thermal neutrons. Thus, through thermal nuclear reactors we can unleash only a small part of energy stored in heavy atoms.On the contrary, the use of. plutonium along with uranium in the £ast breeder reactors could convert most of the U-238 to fissile plutonium. Such a reactor will breed more plutonium than it could consume and thus support a growing nuclear energy program for much longer time. With much larger resources of thorium in our country plutonium provides the essential link to shift from uranium based thermal reactors to Th-232, U-233 based fast breeder reactors. This unique and important role of plutonium has led to i highly sophisticated nuclear industry concerned with the reprocessing of irradiated fuel to recover plutonium ( and also unused uranium), facilili3s for fabrication of plutonium based fuels, design ,construction and opera "'.on of reactors using plutonium based fuels along with associated waste management. It is the nuclear properties of plutonium that gave it this important role but again it is the nuclear properties of plutonium that necessitate special consideration in working with Plutonium due to its associated health and safety hazards. Vie will briefly discuss the nuclear properties of plutonium isotopes and also see iiow the nuclear properties are key to decide most of the important issues related to use of plutonium. 2:PRODUCTION AND NUCLEAR PROPERTIES OF PLUTONIUM: 2.1 Production: Plutonium was discovered in 1941 by Seaborg, Kennedy and Hahl[l]. First isotope of plutonium produced was obtained by deutron bombardment of uranium. The more abundant and fissile isotope Pu-239 was discovered shortly afterwards by neutron bombardment on uranium ,where neutrons were produced by (d,n) reaction in the cyclotron [2]. Today 15 isotopes of plutonium are known covering mass number 232 to 246. The neutron deficient isotopes 232-23/ are prepared by charged particle reactions while more well known and abundant isotopes of plutonium 238 to 242 are produced by neutron capture in nuclear reactors. Fig. 1 outlines the path of formation of these isotopes in the reactor starting with capture of neutrons by U-238 followed by two beta decays leading to Pu-239. Rest of the isotopes are foriaed by successive (n,r) or (n,2n) reactions. The short half-life of Pu-243 prevents further capture of neutrons to form higher isotopes and therefore the chain essentially terminates at Pu-242. For details of production of individual isotopes refer to Hyde [3]. 2.1.1:Isotopic coaposition of Plutonium: The mode of formation of plutonium in the reactor as shown in fig. 1 makes it clear that a plutonium sample would not be monoisotopic. Besides Pu-239, other isotopes of plutonium with mass number 240,241 and 242 are also formed due to successive capture of neutrons. Th' relative abundances of plutonium isotopes can vary appreciably as a given isotope being formed !>y capture of neutrons can also decay or get consumed by capture of further neutrons or undergo fission. Therefore in a general way one IT - 10.1 can say that the final isotopic composition of Plutonium will depend upon the type ot reactor and the average burn up ot fuel. Table 1 shows this dependance. Similarly plutonium isotopes prepared by charged particle reaction will also not de monoisotopicas more than one reaction channels would be open simultaneously. 3.NUCLEAR PROPERTIES OF PLUTONIUM ISOTOPES: j.1: Decay aode s j>|_plutoniu» isotopes: All the isotopes of plutonium are unstable- the instability mainly comes through coulomb effect of high proton number in the nucleus of heavy elements- Alpha decay is therefore primary mode of decay,the other modes being beta decay, electron capture and spontaneous fission.A brief discussion on these decay modes is as follows. For details ot decay scheme, Table ot Isotopes by Leaderer and Swirley[14j and for individual mass numbers Nuclear Data sheets can be consulted. i.1.1: Alpba decay: Table 2 summarises the alpha decay characteristics of Plutonium isotopes. Alpha decay systematics have played an important role and made it possible to predict alpha decay energies using these semiemperical systematic^ J.5J . Alpha decays are mainly to ground states of daughter nucleus, with decay to higher rotational levels are also observed with sharply decreasing probability with increasing angular momentum change. Pu-23<+ decays tr, a rotational band which is based on an isomeric state of U-23b less chan 100 eV higher than the ground state. Alpha decay energies are fruitfully utilised to identify the isotopes but some alpha lines are not resolved such as prominent lines of Pu-23S and Pu-240. On contrary Pu-238 can be conveniently assayed by alpha spectroscopy on 'he purified plutonium samples. Alpha spectroraetry has been used for assay of plutonium and determination of isotopic composition 16,7J. 3.1.2: Beta decay aode: Table 3 summarises the Known characteristics of beta decay of plutonium isotopes, fissile isotope Pu-241 being the most important of them. The very low energy of this ground state transition and lack of any accompanying gamma rays makes it difficult to identify and" assay. Pu-241 has a small alpha branching ratio of 0.0024 %. 3.1.3: Excited states of plutoniiua isotopes: The nuclear energy levels of plutonium isotopes are determined by the study of the decay of parent nuclide populating excited states of Pu isotopes. However the characteristic gamma rays associated with the decay of plutonium isotopes comes from the energy levels of daughter products. Table 4 gives some information about the excited states of some even-even plutonium isotopes. In case of deformed nuclei like those of piutonium isotopes the energy levels are described in the framework of unified model of Bohr and Mottleson [8]. Three types of excitations are encountered, they are intrinsic particle excitations, collective vibrationaland collective rotational states. Bands of rotational excitations are built over the vibrational and nucleonic exited states details of which depend on even-even (e- e) and even-odd nature (e-o) of the nuclei. The low lying level schemes of e-e nuclei is simplified because of pairing of nucleons in the ground states. No nucleonic states are formed below about lMeV which is about the energy required to break the pair.For e-e isotopes the ground state is characterised by zero spin(I) and even (+) parity (fl) though rotational bands are known with odd I and fl. Pure rotational bands are found in e-e nuclei at low excitations and further rotational bands are built over the vibrational bands of the energy of the order of 1 MeV. Rotational bands have an energy sequence proportional to 1(1+1) with even values of I. In case of: odd mass nuclei the excited state sequence is more complicated as the excited state of odd ft nuclei are only of the order of 0.1 MeV apart. The rotational bands of the levels based upon the excited nucleonic states have been observed. Further the energy sequence of the rotational bands are not expressed as simply as in case of e-e nuclei as it depends upon the IT - 10.2 coupling of intrinsic and collective degrees of freedom. Only a limited amount of information is available on excited energy levels of odd A isotopes of Plutonium other than Pu-239. These are summarised in table 5. Table 6 gives prominent gamma rays ( energy and abundances) associated with the decay of Plutonium isotopes[4]. It can be seen that in general the abundance are very low. These gamma rays havs been used fruitful!/ for assaying isotoplc composition of plutonium [9,10]. 3.1.4: Spontaneous fission characteristics of plutonium isotopes: Nuclei with high Z value (>90) show spontaneous fission decay mode whose probability increase* with increasing fissionability parameter X= ( Z*/A *3 7.5) as given by liquid drop model (LDM) which shows that fission bnrri-ars monotonically decrease with X. Log Tl/2 for spontaneous fission shows a decreasing trend with X but dependance is not smooth. More smooth dependance could be obtained by Swiatecki[ll] and improved later[12], when correction? ailsing due to difference between actual (calculated) ground state masses and the sa.oth value of mass given by LDM is taken into account. Even-even isotopes show smooth trend in log Tl/2 (s.f.) vs. X but e-o, o-e and! o~o nuclei deviate appreciably with Tl/2 being higher by a factor of 10J to 10s. Table 7 give3 the spontaneous fission half lives oc important plutonium isotopes. 3.1.5:Neutron emission from plunnium samples. Due to spontaneous fission of plutonium isotopes (e-e) a sample of plutonium will have some associated neutron radiation. Pu-240 is the major e-e isotope which contributes maximum to the s.f. neutrons. Measurement of these spontaneous fission neutrons forms a very efiective way of assay of pure plutonium isotopes- by nondestructive methods (see table 7). (a ,n) reaction in the matrix alio produces neutrons whose yield strongly depends on the chemical nature of lighter component of matrix i.e. carbon, oxygen or fluorine etc. For more detailed mention of nuclear properties see Hyde[3]and Leonard[13j. 3.2: Meutron induced reactions of plutoniim isotopes: The most important nuclear reaction that plutonium isotopes undergo is nuclear fission by neutrons. Odd A isotopes can fission•with neutrons of all the energies. The neutron reaction with plutonium isotopes can be classified by the slow neutron reactions and fast neutron reactions, both playing a vital role in reactor physics calculations. In the low energy region the cross sections are characterised by existence of resonances which are also of great importance for thermal reactors. The dependence of the capture cross sections, scattering cross sections on neutrvjn energies have been studied in details for Pu-239, in lesser details for Pu- 240,241 and to still lesser extent for other isotopes of reactor produced plutonium. An interesting account of these cross sections are given in reference [14]. 3.2.1: Slow neutron cross section of Pu-239: Being the main fissile isotopes of reactor produced plutonium the knowledge Oi reaction cross section as a function of neutron energy is of primary importance. The most complete set of neutron cross section data for Pu-239 in the low and the resonance energy region has been obtained by Bollinger et al[15]. The important feature is the large 0.3 eV resonance and the large component at lower energies which is presumably due to a large part to a resonance at negative energy. There are two important features of the cross section behaviour. The first is markedly non 1/v behaviour of the cross section and second is variation of n(E) such that the value of n in 0.3 eV resonance is only about 82 % of the thermal value. Both these features cause the behaviour of thermal reactors containing plutonium sensitive to details of neutron spectra. These features also cause more stringent requirements to be placed on the accuracy with which the variation of 01 (E), a» (E) and n(E) are known [16].The single level resonance parameters underestimate the fission cross section by a factor of 2. The multilevel interference, fits of Vogt [17] fits the cross sections in these regions very well. 3.2.2: Slow neutron cross section for Pu-241: Pu-241 is fissionable by slow IT - 10.3 neutrons and show resonance charactei1st ics similar to those of Pu-239. Si'/.pahie sample of pure Pu-241 is available since 1960 and accurate measurements on total cross sections of Pu-241 have been made by Simpson and Schuman [18] . Fission data has been obtained by I *V (PROMPT) -f-li MeV C TOTAL) of this to*al 223 MeV, 179 MeV (Ek+Epd) is jpent locally (dissipated), 33 MeV IT - 10.5 {vE+Er+Ev11 t-Fcd) i ^ distributed in larger ipace and 11 MeV is totally lost from the regioij. i.i,l: Neutron r iergy spectra: Calculations carried out by Terrel [29] have shown that the expected distribution in the laboratory frame of reference are represented to a good approximation by a Maxwellian distribution of enrgies as f\l(e!) o( /^ Q~^f"^ In this expression T is the temperature index and not the nuclear temperature. The derivation is based on isotropic emission of neutron from moving fragment, but Terrel has shown that the effect of anisotropic emission is small. Another important correlation is between average energy of the neutron emission spectrum £ with the number of emitted neutrons, .1/ ) 62I C^ O1 In this expression a constant 0. !& MeV is the ejnperically determined average kinetic energy per nucleon of the moving fragment and the second term is derived from the average nuclear temperature of ti>e fragments. This expression has been found to hold for all experimental value of v with in the quoted errors. As the energy of neutron induced fission is increased, this additional excitation energy appears in the excitation energy of the primary fragment resulting ultimately in the increase in the number and energy of neutrons emitted. According to Terrel [231 the dependence of v should be given by where EO=6.7 MeV is assrmed for average energy change per emitted neutron. Similar dependence for E is given by where Eo =6.7 MeV is assumed for average energy change pei emitted neutrons. Similarly dependence for E is given by u / / However this has not been varified experimentally as the expected change is small. Multiplicity of fission neutron is an important parameter and the probability Pv for emission of v neutrons in a fission event is a Gaussian. For further detail see 114,27,28]. 3.3.3: Delayed neutron emission: Majority of the neutrons from the excited fission fragments are emitted in about 10"10 seconds of fission event and are termed as prompt neutrons. A small fraction of them (.2 to 0.5%) are however emitted after considerable delay ( .1 or 50 s) and they are attributed to be emitted by secondary fission products formed following the beta decay i.e. an excited state wiih energy exceeding the binding energy of neutron making neutron emission possible. Several fission product contribute to this delayed neutrons. Scveraldelayed neutrons periods among the bromine and iodine fission products frosi the thermal neutron induced fission of U-235 are given by Perlowand Stehney [30]. These same precursors would be expected to contribute to the delayed neutron emission in fission of any nuclei but the proportions could change. The halflives and probable identification ot isotopes of these precursors are given in [14] along with the abundances and periods of delayed neutrons and alto see ref. [31]. 3.3.4: Sose additional fission characteristics of plutoniu*: Detailed studies on fission characteristics of plutonium isotopes particularly Pu-239 such as mass distribution, charge distribution, kinetic energy distribution and recoil range measurements and angular distribution of fission fragments have been obtained and are available in ref 127,28]. 3.3.4.1: Mass distribution: The mass yield distribution in fission of Pu-239 has been carefully studied by radiochemical and mass spectroroetric techniques by several authors like Katco£f[32], V^n-Gunten[3J], Jain[34]. Mass distribution profile of slow neutron fission of Pu-239 and U-235 (for comparison) given in fig 2 [32]. Noteworthy is that fission product yields in Pu-239 cases are higher by a factor IT - 10.6 of 4 to 10 in the mass region 110-120 compared to yields m 0-23^ fisMon. 6imilar is the case for the heavy side of heavy wing of mass distribution. With higher energy neutrons or charged particles with major difference is that ths valley corresponding to symmetric aass numbers gets gradually filled aa excitation energy increases and the nass distribution becomes single peaked and symmetric at an excitation energy of the order of a few tens of MeV [35].Probability of symmetric mass division varies as a function of slow neutron energies also as demonstrated by Reiger et al[36]. It is expected that if the quantum properties at saddlepoint plays an important role as described in Bohr model - than the symmetric fission yield will vary from resonance to resonance. The other important feature of low energy mass distribution is the existence of fine structure in the mass yield curve which has been attributed to the odd-even effect and to some extent shell effect. As an example of importance is the increased yield of Xe-135 ("7.3 %) relative to fe.4 % for U-235 fission. In addition the fission product fields in rare earth /egion is also higher in Pu-239. This is important as this is the fission product region where products are neutron poisons. Mass distribution for Pu-241 has also been reported by Smith et al [37] using physical method and by Chitaraber et al [38] using mass spe^trometric method. The general characteristics are essentially ths same. Spontaneous fission of Pu-242 was also studied by Smith et al and no difference in the symmetric yield was found between this and Pu-241 (n,f) system. 3.3.4.2: Kinetic energy and recoil ranges of fission products: Distribution of kinetic energy has been obtained ir; fission of Pu-239 and Fu-241 using both the physical methods [39-42] as wall as through recoil range measurement [43-45], A critical review of energy released in fission was published by Unifce and Gindler [46], Binges in air Katcoff et al [43] and aluminium by Satya prakash et al[44] for Pu-239 and for Pu-241 by Ramaswami et al [45] have been measured. The latter two authors also obtained kinetic energy distribution as a futiction of mass split. In these studies the interesting observations were made that symmetric fission products gave ncticably lower kinetic energy or recoil ranges which is termed as kinetic energy deficit and has been related to proximity of symmetric mass division to doubly magic number 132 by Satya Prakash et al [47]. 3.3.4-3: Charge distribution ir. fission of plutcaiua: Charge distribution studies of plutonium isotopes is mainly confined to thermal neutron induced fission of plutonium -239. The width parameters for isobaric Oz and isotopic <7A distributions are 0.55+0.02 and 1.8 + 0.25 respectively. The observed odd-even effects are known to decrease with fissionability parameters and average percentage odd-even effect is about 8 % which is substantially smaller than the values of U-235 fission {25%). Charge polarisation seems to decrease with increasing fissility parameter [48] and value for Pu-239 is 0.48±0.02. Charge distribution shows strong effect due to closed spherical and deformed shells [49]. 3.3.4.4: Spontaneously fissioning isoners of plutoniua: A remarkable case of spontaneous fission was discovered by Polikanov et al [50] in 1962-64. An isomeric form of Am-242 was synthesised which was observed to decay by fission with half-life of 14 ms which is a factor of 10ls faster than the halt life for spontaneous fission from ground state. It was established that the isoraer has a low spin and excitation energy less than 2.5 HeV. Additional example of such isomers were found including those in plutonium isotopes as given in table 9. These isomers have been interpreted as nuclei trapped in the second well in the potential energy diagram of the deformation path leading to fission. Such states mwy decay by a gamma emission that to first well or by fission through the outer barrier with half life much shorter because of thinner barrier encountered, fceviesj articles on isomers are given by Bjornholn and Strutinsky [51] and a double hump barrier for fission is well IT - 10.7 explained in ref. [27] . 4:NUCLSAR PROPERTIES OF PLUTONIUM ISOTOPES OF RELEVANCE TO R£ACTOK The reactors operating on uranium cycle produce plutonium. Plutomua thus produced contributes substantially to the energy production by undergo: i-.g fission and thus prolonging the life of the fuel. However due to various reasons the maximum burn up that can be reached is limited depending on the nature of the fuel, fuel design and the reactor design. Therefore at the end of the fuel lifetime valuable plutonium is recovered as a byproduct. Utilisati• r, of Plutonium can be of two types. 1) Recycle of plutonium in thermal ractors 2) Use of plutonium in the fast breeder reactor. We shall consider the effects on both the possible options. j^.l'lutoniu« in thermal reactors: The four important isotopes of plutoniuia A vons the point of view of reactors are Pu-239,240,241,242. Pu-239 and 241 are fis-:;le isotopes except ti»at Pu-241 has short halflife (14.4 Year). Pu-240 is a valuable fertile isotope even though it is a strong absorber. Pu-242 is « parasite to thermal neutrons although the resulting Am-24JJ does lead to a possibly valuablle isotope Cm-244. A.1.1: Basic cross section data: The thermal cross section or the two fiss-le isotopes of Pu-239 and 241 have non 1/v behaviour and also large resonances in all Pu isotopes near thermal energy range. The precision in the knowledge of cross section for plutonium isotopes is inferior compared to that of uranium isotopes. For comparision the 2200 m/sec cross section data for fissile isotopes of uranium and plutonium are given in table 8. ^j.1.2: Resonance integrals: The resonance data for plutonium are summarised in table 9 along with that of uranium isotopes. A comparision between the resonance integrals of the fertile isotopes of Pu-240 and U-238 shows large difference. Essentially all of the Pu-240 resonnace integral (98 %) is due to large resonance at 1 eV. However due to self shielding effective resonance integral per atom is gradually reduced as plutonium builds up in recycle fuel, the effective value may be as low as 1/10 t : of the infinitely dilute value. The epithermal capture to fission ratio for i>u-239 is seen to be 0.567 compared to a value of 0.5 for U-235. Higher alpha value in general and increasing resonance absorption with energy ( i.e. wnere spectrum shifts to higher energy) results in improved stability and transient characteristics and higher safety [52]. J^.1.3:Delayed neutron fraction: The delayed neutron abundance and group half lifa of precursors from Pu-239 and 241 are given in [14] and [53] . These are equivalent to delayed neutron fractions fJ of 0.0021 and 0.00519 respectively compared to 0.00264 and 0.00647 for U-233 and U-235 respectively. The smaiio delayed neutron fraction of Pu-239 should result in larger power excursion ur> . control rod failures and this should be taken in to account in tranaie^ analysis and control design [52]. ^.2:Plutoniua in fast breeder reactor: The most important nuclear paiac-ii. that is of primary consideration is the controlled propagation of fission ciViir is the number of neutrons produced per fission ,v, given in table 11. Thest neutrons are hard currency of the reactor economy. However, since in the neutrou economy one wants to know what return we get on a neutron absorbed { including those lost to capture) we can correct v using the capture/fission ratio (a) to obtain r\, the number of neutrons liberated per neutron absorbed. Fig. 3 gives a plot of n vs energy of neutron absorbed for fissile materials. For thermal reactors, of these fissile nuclides U-233 liberates most neutron per neutron absorption. Since thorium is relatively easily converted to U-233, this accounts for advantage of thorium cycle over plutonium. For the fast breeder reactor, U-235 is clearly the poor cousin of either U-233 or Pu-239 since it is just barely more than two neutron per neutron absorbed and a greater excess is required considering losses if we are to IT - 10.8 continue the chain reaction (average neutron) produce a new fiasiie nucleus {( a second neutron) and have an excess left over tc treed extra fertile materia}s. [53] J: RADIOLOGICAL AND NUCLEAR PROPERTIES OF PL0TONIUH: The radiological hazards of plutonium are primarily due to high specific alpha activity and high fissility. Table 12 ci'«B radiologicalpropertiea of Pu-239. All the isotope3 of importance of plutoisium viz. Pu-238, Pu-240,Pu-242 are also alpha active. Alpha particles though have very small range but have high ionizing power Therefore most of plutonium hazard originate from two major consideration. First is body intake followed by ingastion and second whether plutonium is in soluble or insoluble form. Pu- 241 with relatively shorter half-life has significantlly Ie3s health and safety criterion being a beta emitter of low energy.It is because of small range of alpha particles and majority of plutonium compounds handled being insoluble the exposure to the surrounding tissue in vicinity of plutonium bearing particles would be much greater than the value given in table-ll for insoluble compounds based on giving to all long tissue on average dose of not more than 0.5 Sv/yr exposure [54]. Another important health and safety consideration which originates from the nuclear properties of plutonium is termed as criticality hazard and its control. The criticality hazard of plutonium originates dueto the fact that even-even isotopes of plutonium undergo spontaneous fission producing on an average two to three neutrons per fission. Hence thera wil: be a critical mass of plutonium which can support fission chain multiplication resulting in criticality accidents. The critical mass depends upon a large variety of parameters such as at, its variation with neutrn energy, resonances in cross section, isotopic composition of plutoniura, chemical form of plutonium, solid, liquid or solution, geometry of sample, material surrounding the sample i.e. hydrogenous reflection, multiplication etc. Table 13 gives a very limited information about minimum critical masses (Xg) for U-235 and Pu-239. Critical mass for plutonium is always significantly less than that of U-235 [54]. Criticality hazards are to be taken in to design of any plant handling plutoninm in large amount such as fuel fabrication facility and reprocessing facilities. Important aspects of criticality hazards are covered in reference [55,56]. d: NUCLEAR PROPERTIES OF PLUTONIUM OF RELEVANCE TO FUEL FABRICATION: Radological properties of plutonium dictate design of fuel fabrication plants handling plutonium. Working with alpha tight boxes is the measure requirement. However, if the pl^tonium used is originating from the fuels of higher burn up, shieldings and provision against gamma and neutron radiations are also required [55] .The design of such facility has to also take in to consideration the maximum amount of plutonium that can be worked with or allowed to accumulate or put to store from the point of view of criticality either in normal conditions or in accidental conditions such as those arising due to earthquake and flooding etc.. The differences in the fission product yield distribution in Pu-239 and U-235 significantly effect the fuel behaviour during the irradiations. Fig 2 shows this difference in thermal neutron induced fission. A very similar difference exists in fission caused by neutrons of high energy. The mass yield distribution of the fission products is important in so far as it determines the amounts oxygen getters relative to more noble fission product like Pd, Rh, Ru which do not work as oxygen getter. In case when Pu-239 is the major fissile nucleus in fuel significant drop occurs in the zirconium isotopes from a total of about 30 % to 20 I with corresponding increase in the noble metal fission products ( 40 % to 60 %) resulting in the significant increase in the available oxygen with increasing ourn up. This would thus cause increased oxidation of the clad [57]. f:MUCLEAR PROPERTIES OF PLUTQNWIUM OF RELEVANCE TO REPROCESSING: IT - 10.9 In reprocessing of irradiated fuel the major Kay in which th« nuclear properties of plutonium enter are through the health and safety aspect3. Plutonium in the reprocessing plant is encountered from a very dilute 3tatc is in diss.jj ;er solution or in waste streams, bot associated with very high radiation fields to very concentrated forms as solution and solids in the conversion laboratory. In the latter situation the major health and sal.ty consideiations arise from those of pluton,ium. In addition tc this here too the much higher yields of nobie fission products in plufoniura fission play some role in reprocessing. Those fission products are left as innsoluble residues in the dissolver cells. These .1 nsoluMe particles are identified as alloys of the above elements with U or i'u such o UPu(Pd,Rh>3. These insolubles float in solution in finely divided form xith p^iticle size in the range of 0.5 to 100 micron which necessitates intronuction of very effective feed classification steps [56]. II: CONCLUSION: Nuclear properties of plutonium isotopes have been discussed briefly. The most important nuclear property of the most important and abundant isotopes of plutonium (Pu-239) is its fissility with all the energy spectrum of neutrons. The additional great importance of plutonium -239 comes from its ability to fce ideal fuel for a fast breeder reactor which will enable us to produce energy from practically all the uranium present instead of from mere 1% possible fr^-m thermal Teactoxs. ^he important iissioti tfiarateristics oi ?i:-2i^ has also bees discussed. A discussion has been made though not exhaustive how nuclear properties of plutonium are of importance to different stades in nuclear technology. Finally on the fiftieth anniversary of its discovery I salute its discoverers and to all those who contributed towards understanding the nuclear, chemical and physiealproperties of plutoniura and contribution to its utilisation for an energy abundant and bright future. ACKNOWLEDGEMENT: The author is grateful to shri P.K.Pujari and Dr S.B.Manohar for their help in preparing this manuscript. REFERENCES l.G.T.Seaborg,E.M.tfcMillan,J.W.Kenedy ?.na n.C.Wahl Phys.ReV 69,366(1946). 2.J.W.Kenedy,G.T.St'aborfl,E.Segre and A.C.Wahl Phys.ReV.70,555(1946) 3.The Nuclear Properties of The Heavy Elements, vol 2, Detailed radioactivity properties; E.K.Hyde, J.Perlman and G.T.Seaborg (L97LK Uovic publications lac,, aew^cjcif.. 4.Table of isotopes, 7th edition- ed. C.Michael Lederer and Virginia S.Stforley. A witey inter science publication 1978 and Table of radioisotopes by Edgardo Browne and Richard B Firestone(1986). A Witey inter science publication. b.l.Perlman, A.Ghinso and G.T.Seaborg, Phys. Rev.,77,26(1950). 6.Proceedings of srflinar on "Alpha particle spectrometry- technique and application", Geel,Belgiun, October 14,1981. Int. J.Appl.kailiat. , ib, 239 (1S»«4) . 7.H.C.Jaii. and S.K.Agrawal, artificial radioactivity,ed. K.Narayai: Rao and H.J.ArniXar,1985,pg. 263. Tata McGrawhill publishing company.New York. H.A.bohr and b.R.Mottelson,Dan.Mat.Phys.Medd,27(16):1.174(1953) . y .S.B. Ma tioh-ir, S.K.Agrawal , S . MK. Deshmukh, A.« . Parab, V. p .burte, M.C.Jsin and Satya rraXasu. J.Radioanal. Chem 6J(1),145U-981). lO.k.ounnik•UCRL-7t>418,197 5;UCRL-52220,1977. T.N.Dragner, B.P. !> TABLE- l:Isotopic composition of reactor produced plutonium. A typical example. Plutonium PHWR(CANDU) BWRS isotope Natural Uranium Z\ enric.ied I) Av.burn up b';"uo nwlj/T AV. burn up ilOOUHWD/T 238 U. 3 U.34 239 b7.65 70.10 240 26.06 17.16 241 4.7B b.40 242 i. n t .00 Table-2:Decay characteristics of Plutonium isotopes. Isotope 1/2 Decay EattfeV; (% abundance) 232 36min EC s 98% a^2% 6. ba 233 2omin EC i>9% a w.l* h . JO 234 9.0, hr EC 94% ao.l% 6.1a 235 26 Din EC 99% a.003% 5.8b 236 2.85 Yr a b. 763(69%*,5.71(31%) 5.61u(0.18%) 237 4b.b day EC 99%, aO.OOms 5.65i21%),5.36(79%) 238 86.4 Yr a b. 49(72%) ,5.4b(2a%) 5.352 (0.09%) 239 24360 Yr a b.14(72.b),5.13llb.8%) 5.09(10.7%),5.06(0.037%) 4.99(.013%),4.91(.005%) 240 6580 Yr a 5.16(76%),5.11(24%) b.OJ {0. 1%,' ,4.851.003%) 241 13.2 Yr beta 99%,a.00z4%. 4.8a(75%),4.84(25%) 242 a 4.89(76%),4.85(24%) 244 7.6*io' Yr a 4.55 (calculated) Table-3:Beta decay characteristics of plutonium isotopes. Comments Isotope '1/2 241 13.2±.2 Y 20. 3 Ke\/ 5/2+->5/2- first fobiddeen gs transition 243 4. 98 hr 597 KeV (62%) 84 KeV gamma observed. 490 KeV (38%) 245 10.1 hr yb = 1.2 rteV (cal) 246 10.85 d 330 KeVI127%) Transition to excited 150 KeVI;/3%) state of Am-246 at 47KeV. IT - 10.12 Table-4: Observed excited states of even mass Pu isotopes. + I o r 6+ (i+ others T 236 0 43.4 239 0 44.11 146 .0 303 ,7 514 600 935 986 240 0 42.88 141 .8 292 -C 600 7060,1400 242 0 44.6 Table-5:Some characteristics of odd A plutonium ground state. Pu-237 ground state is 7/2 and a level has been observed at 145KeV with a 1/2+ assignment. Pu-241 The g.s. is 5/2+ and a level has been observed at 172 KeV with 7/2+ assignment. Pu-243 The ground state has been identified as 7/2+. Table-6:Major gamma ray signature of important plutonium isotopes Isotopei Energy(KeV) Intensity(g-s) Pu-238 766.4, 152.77 1.5*1Q5,6.5*1Q5 Pu-239 413.6,129,208 3.4*104,1.4*10:> Pu-240 Pu-241 207.98,164.59,:148.6 2.0*107,1.8*106,7.5*106 Pu-242 Table-7:Spontaneous fission of fissionable isotopes of plutonium. Isotope SF halflife v (SF) SF/g-s Pu-238 4.9*10^ 2.26 l.l*103 Pu-239 2.2 1.0*10 2 Pu-240 1.17*10" 2.17 4.7*10 Z Pu-241 5.0*1015 2.20 1.1*10 2 Pu-242 6.8*101 2.16 8.0*102 Table-8:Important fission data for li-235, Pu-239 and Pu-241. U-235 Pu-239 Pu-241 a. (barn) 679.912 .3 1008+1.'.5 1391+2.2 at (barn) 579.5+2 .0 742.4+3 .5 1009+9 a 0.1734 0.3580 0.379 n 2.071 2.114 2.154 V 2.430 2.871 2.969 IT - 10.1 3 TableM:Resonance -it arals for infinite dilutiondO MeV to 0.5 eV) [53] s.no. Isotope Measured Used 1 U-235 140±8 140 l-.f 280±ll 280 a O.5O±.O2 0.500 2 U-238 l-.c 280±12 269 J Pu-239 I-.c 177 I-.f 31O±2O 312 a 0.567 4 Pu-240 l-.a 837O±38O 8467 5 Pu-241 I-.C 164 I-.f 557*33 517 a 0.317 b Pu-242 l-.a 1280±60 1112 Table-10:Comparision of plutoniuw design parameters and their differences relative to uranium parameters [53J. Parameters rlixed oxide Reason for Consequences core differences 1.Moderator more negative Increased Improved stability temp, coett. resonance and transient absorption and characteristics, spectrum shift. 2.void coetf - do - - do - -do - except for steam break. 3.Doppler coetf. -do- Pu-240 resonances Improved trannsient 4.cold to hot reactivity lncreased Larger moderator Increased control swing temp, coeff. rod requirements. S. Installed reactivity Reduced Reduced depletion none requirements rate reactivity saturates 6.Control rod Increased Larger moderator Increased requirements and Doppler coeff. control rod number. 7.Control rod Reduced Thermal flux reduced -do width 8.boron width -do- -do- none V•Fission lncreased Larger yields Reactivity product and increased reduction. poison. resonance absorption. CONTINUED 11 - 10.U 10.Local power Increased Increased Better fuel peaking water acrth management required 11.Delayed Reduced. Larger power n-fraction. excursion on rod ejection accident. Table-ll:Neutrons liberated per neutron capture in fissile materials Nuclide v/fission n/fisssion n/fission U-233 2.50 2.27 2.60 U-235 2.43 2.06 2.18 Fu-239 2.90 2.10 2.78 Tabl' 12:Radiologicai properties of Pu-239 Max. permissible soluble insoluble Body burden MBq/m If inhaled 1500 600 If ingested 1500 Critical organ If inhaled Bone Lung If jngested Bone GI tract Fraction reducing Critical organs By inhalation 0.18 0.12 By ingestion 0.0001 Maximum permissible concentration in air and water 4th/Wx Exposure, MBq/m (MPC)air 0.075*10-6 1.5*10' (MPC)water - 3.7 3.0 Table-13:Minimum critical masses(Kg) [55] U-235 Pu-239 Solution 0.82 0.51 Metal 22.80 alpha phase(-f =19.6mg/m^l 5.6 delta phase(^»?5.8 mg/m ) 7.6 Oxide sintered (-\ "10 mg/ai ) 35.0 12.7 Green pressed ( -\ "5 mg/m ) >100 32.0 "^is density IT - 10.15 NEUTRON YIELD PER NEUTRON ABSORBED, o o Some AspectB Of Solution Chemistry Of Plutonium S.K.Patil Fuel Chemistry Division Bhabha Atomic Research Centre, Trombay, Bombay-400085, India. I. Introduction Among the transuranium, the chemistry of plutonium is most extensively studied. The studies on the solution chemistry of plutonium are vital as the major methods of separation and purification are based on its chemical behavior in solution. For the analysis of plutonium, often its chemical separation, which also needs the knowledge of solution chemistry, is necessary. The chemical behaviour of plutonium, like uranium or neptunium, as manifested by multiplicity of oxidation states has been extremely useful in its laboratory or industrial scale separations which utilise the redox reactions and its complexing with various anions in different oxidation states, thereby making its separation and purification relatively less difficult. The basic investigations on solution chemistry of plutonium are fascinating and challenging, as it is a unique element which exhibits, in acidic aqueous solutions a range of oxidation states from + 3 to 46 all of which can coexist in significant concentration in the same solution. Under the experimental conditions beginning with a single known oxidation state of plutonium, the change of oxidation state can occur due to preferential complexing of Pu in som« oxidation state or due to disproportionation reaction. Hydrolysis usually leading to polymer formation and the influence of its intense alpha activity, especially whii° dealing with Pu can further complicate the investigations. The understanding of solution chemistry of plutonium is vital to its recovery end purification from spent fuels as well as devising.methods for its removal from wastes, Besides,Cleveland's book , in,the proceedings of a recent symposium , and in a recent book excellent papers on the various aspects of solution chemistry of plutonium are published. (n this paper an attempt has been made to review some aspects of solution chemistry of plutonium. II Oxidation States Solvent extraction or ion exchange methods are frequently used in the complex formation studies of plutonium. It is observed that experimental conditions used for some investigations can induce change, in the oxidation state of plutonium. In the extraction of Pu(IV) from perchloric acid, which is commonly used as relatively non-complexing medium, for studying, aqueous.ccmplexing by a liquid cation exchange DNNS ~ ', HTTAC }, BPHA* ' or a cation ion exchange resin1 ', Pu(IV) is partially reduce** to Pu(III). Though in a number of such studies, nitrough acid is used as the holding oxidant for Pu(IV), it is shown by absorption spectral studies that Pu(IV) in fact is partially reduced to Pu(III). The investigations on the use of various redox reagents K°° s'-iwn that Vanadium(V) is X T - 1 1 . 1 the suitable holding oxidant for Pu(lV), the UBA of which was suggested ' earlier for oxidation state analysis of Pu. In similar studies, on Pu(III), ascorbic acid, iodide or. preferably quinhydrone as holding7reductant has been used . Several wnrkerr have reported difficulty in maintaining oxidation state of Pu(VI) in solvent extraction or ion exchange studies and have used different holding oxidants such as dichromate, bromate or permanganate in some cases without success in preventing reduction of Pu(VI). In general it may be said that in studying the solution chemistry of Pu in various oxidation state, attention must paid to ascertain that the oxidation state remains unchanged and the use a suitable redox reagent or protection from light may help in stabilizing the desired oxidation state. Ill Complex Formation The study of complex formation is important to know the species of Pu as well as devising the separation and analytical procedures. A large effort has been devoted to determine the stability constants of aqueous plutonium complexes . The values of stability constants often show wide variations and one of the factor responsible for this probably is inadequate attention paid to maintaining the oxidation state of Pu. For example the reported values of stability constant (Pj) for sulphate complexing of Pu(IV) obtained at ionic strength ranging from 0.5 to 2.33 show variation from 9.5 to 1100. Recent data obtained using either nitrous aoid or V(V) as ihe holding oxidant for Pu(IV), though has narrowed the variation, the values obtained using the latter holding oxidant are considered more reliable for reasons explained earlier. The,stabi1ity constants of fluoride complexes are critically surveyed and the available values for Pu ions are categorised either as tentative or doubtful. For Pu(111) no stability constants are reported. The fluoride ion selective electrode has proved useful in obtaining reliable data. Recently such studies have been reported. In studying complexing of Pu, especially of Pu(IV) the use of relatively high acidity medium bi comes necessary to minimize hydrolysis and as consequence ot using high acidity liquid junction potential (Bj) changes du flioride ion is added progressively. If free hydrogen ion concentration is measured simultaneously using quinhydrone electrode Ej values can be estimated . However both Pu(VI) and Pu(IV) are reduced by quinhydrone, thereby preventing its use to measure-hydrogen ion concentration. An iterative method was developed to calculate free H , and hence Ej, from measured potential values using fluoride ion selective electrode. The stability constant values reported on fluoride complexing of plutonium are summarised in Table 1. IV Solvent Extraction Extraction with HDEHP Di-2-ethylhexylphoaphoric acid (HDEHP-HY) is well known as a liquid cation exchanger and in the extraction of U(VI) or trivalent actinides from low concentration IT - 11 .2 acid_media behaves as^gucb. However in the extraction of Th(lV) or Pu(IV) ' from nitric or hydrochloric acid by HDEHP nitrate or chloride containing species are extracted. Systematic work now in progress reveals that while Pu(IV) species extracted from H2SO4, is PuY2(HY2>2» those from aqueous HC1 and HNO3 are PuClY(HY2)2, PuCl2Y(HY2)2' Pu(N03)Y(HY2)2 and Pu(NO3)2(HY2)2- Synergic Extraction of Pu(IV) : Synergic extraction of metal ion though is exteusiyely studied until recently only a few data have been reported on the extraction of_Pu. Recently, however, extensive studies have been carried out on the synergic extraction Pu(IV) by HTTA-neutral donor combination. th the conventionally used slope-analysis method and Jobs msii' ..=> have been used, to know the Pu(IV) species extracted. Besides the organic phase adduct formation studies are carried out using a absorption spectral method. When nitric acid is used as tiie aqueous medium of extraction, several mixed adduct species, with nitrate ion participation, are extracted ' . V PI utonium Bee.; In the fuel fabrication of the Pu-based fuels, scrap is generated due to rejection of some fuel material due to its failure to meet physical or chemical specifications. Valuable plutonium in this material has to be recovered for its recycling to fuel fabrication. Though some of this material can be recycled by dry route, frequently wet chemical procedure involving dissolution, purification and conversion of plutonium to PuG2 becomes necessary. The mixed carbide fuel scrap (about 60%Pu) was oxidized under controlled conditions to get mixed oxide which could be readily dissolve in nitric acid with small amount of HF added to aid the dissolution. Direct oxalate precipitation from the resulting solutions containing about 25% U was explored to obtain PuOg product with desirable chemical and physical criteria. The precipitation gave Pu(IV) oxalate, which on calcination, yielded PuC>2 with a high specific surface area tor fuel fabrication. Besides good separation from uranium, decontamination from essentially all undesirable impurities could be achieved. The procedure was used to recover plutonium from large quantities of mixed carbide scrap. Plutonium from solution obtained by dissolving MOX fuel material jcrap (2-4% Pu) could not, however, be processed in a similar way by direct Pu(IV) oxalate precipitation due to the presence of large quantities of Uranium. Ion exchange separation and purification of Pu was studied using macroporous anion exchange resins. Nitric acid solutions obtained by direct dissolution of MOX fuel scrap and containing uranium up to about 300 g/l.were processed directly by using Amberlite XE-270 resin column^ . The separation of plutonium from uraniun was excellent and the elution of Pu was sharp (Table 2) resulting in concentrated plutonium product solutions. IT - 11.3 The effluents resulting from Pu(IV) oxaiate precipitation contained high enough concentration of plutonium to necessitate its recovery. Both ion exchange; and solvent extraction procedures have been reported for the same. A process based on direct loading of nitric acid—oxalic acid effluent, after adjusting .the nitric acid concentration, on Amberlite .'.^-270 was developed and successfully used for the recovery of plutonium form large volumes of oxaiate effluents. References 1. J.M. Cleveland, The Chemistry of Plutonium, Gordon and Breach Science Publishers, New York 1970. 2. W.T. Carnal 1 and G.R. Choppin (Eds) Plutonium Chemistry, ACS Sj'mp. Series 216 (1983). 3. J.J. Katz, G.T. Seaborg and L.R.Morss The Chemistry of Actlnide Elements 2nd Ed. 2vols. Chapman and Hall (1986) 4. S.K. Patil and V.V. Ramakrishna, J. Inorg. Nucl. Chem., 35, 3333 (1973). 5. S.K. PL. * and V.V. Ramakrishna, Radiochim. Acta, 1JJ, 27 (1973). 6. 0. Cristallini, C.A. Dupetit and J. Flagenheimer, Ibid, 5_, 181 (1966). 7. J.J. Fardy and J.M. Pearson, J. Inorg. Nucl. Chem. 36, 671 (1974). 8. K.L. Nash and J.M. Cleveland, Radiochim. Acta, 3JJ, 105 (1983). 9. P.R. Vasudev Rao, S.V. Bagawde, V.V. Ramakrishna and S.K. Patll, Radiochem. Radioanal. Letters, 2£, 95 (1976). 10. P.R. Vasudev Rao, S.V. Bagawde, V.V. Ramakrishna and S.K. Patil, J. Inorg. Nucl. Chem. 38, 1339 (1976). 11. D.J. Savage and J.L. Drummond, Proc. TAEA Symp. Analytical Methods in Nuclear Fuel Cycle, Vienna, Nov.-Dec. 1971, STI/PUB/291 p. 409 (1972). 12. M. Ward and G.A. Welch, J. Inorg. Nucl. Chem. 2, 395, (1956). 13. G.tf. Nair, C.L. Rao and G.A. Welch, Radiochim Acta 1_, 11 (1967). 14. P.K. Khopkar and J.N. Mathur, J. Inorg. Nucl, Chem., 36_, 3819 (1974). IT - 11.4 15. J.J. Fardy and Buchanan, Ibid, 28, 579 (1976). 16. P.R. Vasudeva Rao, S.V. Bagwde, V.V. Ramakrishna and S.K. Patil, Ibid, 40, 123 (1978). 17. A.S. Ghosh Mazumdar and C.K. Sivaramakrishnan, Ibid, 27, 2423 (1965). 18. S.K. Patil and V.V. Ramakrishna, Ibid, 3J3, 1075 (1976). 19. W.E. Keder, J.C. Sheppard and A.S. Wilson, Ibid, 12, 327 (1960). 20. J.M. Cleveland, Coordin. Chem. Rev. !5 101 (1970). 20a A.M.Bond and G.T.Hefter "Critical Survey of Stability constants and related Thermodynamic data of fluoride complexes in aqueous solutions" Pergamon Press, Oxford 1980. 21. S. Ahrland and L. Kullberg, Acta Chem. Scand, 25, 3457 (1971). 22. R.M. Sawant, N.K. Chaudhuri, G.H. Rizvi and S.K. Patil J. Radioanal. Nucl. Chem. Articles ^_1, 41 (1985). 23. D.F.Peppard, G.W.Mason and s.McCarty J.Inorg. Nuol. chem. 1J3 138(1969). 24. D.F.Peppard, M.N.Namboodiri and G.W.Mason Ibid 24_ 979 (3962) 25. G.W.Mason, H.E.Griffin and s.McCarty Lewey Ibid 4J3 391 (1981) 26. J.J.Fardy and J.M.Chilton Ibid 31 3247 (1969) 27. K.H.Lieser, B.Stojanik and H.D.Greiling Rediochim. Acta 21 41 (1984) 28. H.Irving and D.N.Edgington J.Inorg. Nucl. Chem. 2jO 321 (1961). 29. V.V.Ramakrishna and S.K.Patil Structure and Bonding 56_ 35 (1984) (and references theirin) 30 D.D.Sood Proc. IAEA symp. "Back end of the nuclear fuel cycle -.Strategies and option Vienna May 1987 STJ/PUB/758, p.561 (1987) 31. V.V.Ramakrishna and S.K.Patil Proc. DAE "Radiochemistry and Radiation Chemistry Symposium" Bombay Feb. 1988 p.IT-4.1 IT - 11.5 Table 1 Stability constants of the fluoride complexes of Plutonium Anion Method Ionic strength Logpj Logp2 Logp2 Ref Pu(IV) F-ISE 1(H , NaC104 ) 3.8 6.31 8.00 a F-ISE l(do) 4.22 b CIX 2(HC104) 4.97 c DIS 2M(HC104) 4.09 d + Pu(IV) 'F-ISE 1(H NaC104) 7.61 14.77 20.11 26.07 e CIX 2(HCin^) 7.40 f 1(HCIO4) 7. 15 f 1(HNO3) 6.99 f 2(HNO3) 6.73 f DIS 2fHC10/. ) 7.59 13.51 g Pu(III) F-ISE 1(H , NaClO^) 3.14 7.76 12.38 References a R.M.Sawant, N.K.chaudhuri, G.H.Rizvi and S.K.Patil, J.Radioanal Nucl. chem. 9_1 (1985) 41. b G.R.Choppin and L.F.Rao, Radiochimica Acta. 37(1984)143. c V.N.Krylov, E.V.Komarov and M.F.Pushlenkov, Radiokhimiya, 10(1968) 717,719,723. d S K-Patil and V.V.Ramakrishna, J.Inorg. Nucl. chem. 38(1976) 1075. e R.M.Sawant, N.K,Chaudhuri and S.K.Patil. J.Radioanal. Nucl. chem. l_43(1990) 295. f V.N.Krylov and E.V.Komarov Radiokhimiya 11(1969) 101, 103. g S.V.Bagawade, V.V.ramakrishna and S.K.Patil. J.Inorg Nucl. chem. 8 (1976) 2085. h R.M.Fiwant and N.K.chaudhuri Unpublished work. IT - 11.6 Table 2 Ion exchange data for Pu recovery from MOX scrap with Amberlite XE-270 Resin Feed Composition Loading Pu Pu 7M HNO3 wash 0.35M HNO3 elution Vol. tr Breakthrough Loss Vol Pu loss tr Pu in one 1 U Pu HNO3 min % 1 % min bed vol. g/1 1.7 285 12.7 4 50 10 1. 6 7.5 5.7 200 98.4 1.7 321 6.6 5 50 1.2 0.2 7.5 1.0 200 98.4 2.0 239 9.8 6 60 13 1.5 6.0 4.5 130 98.6 2.0 266 8. 1 6 60 3 0.4 10.0 1.0 180 98.3 2.0 232 10.0 7 60 0.01 NIL 1.0 NIL 180 99.0 0.75 100 0.4 6 ~o NIL NIL 3.5 NIL 90 98.0* 0.75 232 10.0 7 60 0.06 NIL 2.2 NIL 90 98, 8* 0.75 218 9.3 7 60 0.5 0.2 1.6 0.1 60 98.8* 0.75 198 4.5 7 30 0.2 NIL 1.6 NJL 75 93.8 0.75 192 4.4 7 20 NIL 0.1 4.4 0.1 200 89.1 STABILITY OF PENTAVALENT PLUTONIUM H. Capdcvila, P.Vitorge CEA DSD/SCS/SGC 92265 Fontenay aux roses Cedex FRANCE SUMMARY Pu(V) stability in Na, H (0.1 or 1 M) CIO4 (0.1 to 3 M), and in environmental conditions is discussed. It is shown that PuC>2+ can be stable from pH=l to 8. Redox potentials and activity coefficients are reviewed, and it is concluded that there was too few experimental determinations of E(PuO2+/Pu4+). The disproportionalion constants of PuC>2+ and Pu4+ are then measured spectrophotometrically and extrapolated to 0 ionic strength according to the SIT: lg(Ky) values at different I [ KV in (Mol/1)-* ] 0.1M 0.5M 1M 2M 3M 3.42 4.36 4.85 4.88 5.13 lg(KIV)at I=1M 4 [K1V (Moiyi) ] -2.06 reviewed E(Pu^+/I>u4+) and E(PuO22+/PuO2+) are used to propose I E(V1/IV) E(VI/III) E(v/rv) E(V/fII) CM) (mV) (mV) (mV) (mV) 0 3005 1008 1052 1031 0.5 1048 1001 1163 1058 1 1071 1015 1200 1080 2 1069 1029 1188 1088 3 1083.5 1050 1200 1108.5 Key Words : Pu, redox potentials, environment. IT - 12.1 I BIBLIOGRAPHY I.I Historical It has been shown in 1949 [ I ] that the pentavalent oxidation state of plutonium exists in aqueous solution . The first observation on which is based this discovery is the bleaching of a Pu(VI) solution treated in order to be reduced. Every oxidation state then identified was highly colored: blue, green and orange-pink for plutoniurn in the +3, +4 and +6 oxidation state in acidic medium. However the autnors have noticed that immediately after adding sulfur dioxide ( reducing agent) to a Pu(VI) solution, it turned nearly colourless and after several minutes yellow green indicating the reduction to the tetravalent wxidaiion state The same phenomenon was observed when hydroxylamine hydrochloride was used as reducing agent. This fact eliminated the possibility of explaining the phenomenon as being due to complex formation with sulfite or sulfate ions. This colorless solution had an absorption spectrum quite different from others oxidation states ones that were well known, which proved the existence of plutonium (V). T'ie main characteristic of this spectrum was a relatively strong, narrow absorption band centered at 569 nm. The same authors [1] have produced Pu(V) by electrolytic reduction. Pu(VI) has been reduced in a 0.5M HC1 solution during 86 min applying a potential of 1.2 V. The number of moles of Pu(VI) reduced equaled the number of equivalents of electricity passed through the cell. Then, the conclusion was that a one electron change occurred. 1.2 Redox potentials 4+ 3+ 2+ + Rcdox equilibria are established rapidly for Pu /Pu and PuO2 /PuO2 couples since metal- oxygen bonds are neither formed nor broken. So, it was possible to measure Ihese reversible redox potentials using standard electrochemical techniques [2]. All these techniques give consistent results and have been tested before by the same authors [3], with uranium for which the redox potentials are well known. They determined redox potentials at several ionic st1 ;ngths, between 0.5 and 3M, in perchlorate media. Then, the authors extrapolate at 1=0 in order to obtain the standard potentials and to compare with the v,thers published results. These values are being examined within the NEA-Thermodynamic Data Bank that will publish a compiled data work on the plutonium. We compile these values in the table 1: we shall use these values these values later. I". - 12.2 in^ie 1 ttVl/V) Ae(VI/V) (Mol/1) (mV) (mV) (Kg/Mol) (Kg/Mol) 0 958 1010.5 0.29 0.54 0.1 940.7 969 0.5 933 954 1 941 959 988 j _ 96/ 101/ The experimental values are here extrapolated to 1=0 using the S.I.T, which introduce 1 fitted parameter, Ae (sec below §111.3.): The results arc the same as in [3) for Ae, but slightly different forE II i> iiiwiu difficult to measure rcdox potciiiiuls of irreversible systems. So, there are very few publication concerning tliis sulijecl. In ordei to determine lhcnrn .±\•na;nic .mil kinetic constants Kasha [4] has studied the equilibria (1 to 3) in different media (| U • j = 0.05 and 2M |t.."iO4-J = lor2M). 3+ 2 1 1 2 Pu i- PuO2 '• + 4 ii* ^ 3 Pir " ' + 2 MiO 3+ 2+ 1 + Pu + 2 PuO2 + 2 H2O ^ 3 PuO, -f 4 K (2) 3H 2+ u Pu " + PuO2 f* Pu' (3) All the concentrations were measured S|iea:ijpliotomeincally. I'he autlior has plotted the spparent constant of the studied equilibrium veisus tmio am! stopped the experiment when it was quite stable. The main problem in these experiments is the hydrolyse of Pu41" species when ihe acidity is less than + 0.05 M. So, we consider that in this work, the more precise value is the PuO2 dismutation equilibrium 2H one. There is also a great uncertainly on the molar absorption coefficients and especially on the PUO2 one because of the speeirop!Kilometer performances. Rabideau |5] has studied the equilibrium: 3+ 2+ 4 2 Pu + PuO2 + 4 H' ^ 3 Pti '• + 2 H?O (1) 2f at different acidities (from O.'.'M to 1M oi llCIf)^) and at a fixed ionic strength (1=1M). A PuO2 ' solution at known concentration vvas added to a solution containing known concentrations of Pui+ and I I .2.3 I'll4"1". The author measured versus time, the mean redox number and ihe redox potential of the solution, then deduced each concentration. So, this experimental method did not measure directly the concentration of each species present in the solulioa To interpret his results the author considered that the apparent constant has reached its equilibrium value when the Pu4+ decreasing speed equals radiolyse one. He had to take into account the Pu4+ hydrolysis, that occurs when the acidity is less than 0.05 M. He determined in this work the Pu4 + disproportionation constant, at 25°C in acidic perchloric medium of ionic strength 1M. Later, Rabideau [6] has studied the reaction (4) in HC1O41M medium. 3+ 2+ Pu + Pu02 «* Pu^+PuO/ (3) 2+ 3+ Known PuO2 and Pu solutions have been mixed. The author measured the redox potential of the solution, assuming that: 4+ [PuO2+] = [Pu ] (4) 2+ + and computed the redox potential value of: i02 /Pu02 . However more recent works (since 1956) reported measurements of redox potentials of the reversible i 4+ couples, but we have used these experimental values to calculate the PuO2" 7Pu redox potential. If (4) were true, potential of the solution, measured during the experiments, would have been the + 4H PUO2 /PU normal potential [3], but we have not keep this hypothesis (4), since in this medium the pentavalent plutonium is not stable and disproportionate. So, the stoichiometrie of the reaction is quite different. We consider that there might be considerable uncertainty in the new numerical value of + 4+ PuO2 /Pu potential we obtain, because the potential solution is measured 2min only after the mixing of the two solutions. It is really too fast to obtain equilibrium. In a recent publication [7] Toth Bell and Friedman have studied Pu4+ disproportionation into 2+ 3+ PuO2 and Pu , in HNO3 media of different acidities (from 0.075 to 0.42M), ionic strengths, temperatures (from 5°C to 25°C) and total plutonium concentrations. Each oxidation state concentration is measured spectrophotometrically. Many experimental values are compiled in this work, but there are several problems in the experimental method. In low acidity several Pu(IV) plutonium species exist (Pu4+, hydroxide and nitrate complexes) and so the absorption spectrum results are too complicated to be properly interpreted. Preparing a Pu4+ solution by dissolving the mother solution in aqueous solution allows Pu(IV) polymer formation at uncontrolled concentratioa IT - 12.4 There are also some problems with the data treatment: the author interpreted his results in order io determine standard equilibrium constant using the Debye Hilckel theory which is not really valid in these conditions. 1.3 Pentavalent plutonium in the environment Until now, it has been considered that the most stable oxidation state of plutonium was tetravalent one. However, thermodynamic predictions, using the most accurate values show that pentavalent plutonium must be taken into account. Stability domain exists for all four oxidation states in solution in environmental conditions. Pentavalent plutonium would be stable from pH = 1 to pH = 7. On the other hand, several authors [8] [9] have analysed water samples from natural environments and concluded that the dominant soluble form of plutonium was Pu(V). These experiments consist in separating Pu(VI) and Pu(V) using a method based on the preferential adsorption of Pu(VI) onto silicilic acid or TiOj powder. These results are in agreement with thermodynamic predictions, but these techniques of separation could be criticized. The adsorption on solid support could disturb the equilibria between the soluble species of plutonium, especially when plutonium concentration is very low. It is then important to use another method to verify the stability domain of pentavalent plutonium. II NOTATIONS lg rdecimal logarithm In :neperian logarithm zj :charge of the ion i aj :activityoftheioni q :molarity of the ion i (mol/1 = M) mj :molality of the ion i (mole/kg = m) g activity coefficient of the ion i e(i j) specific interaction coefficient between the ions i and j of opposite charge VT D :Debye-Huckel term D= 0.5107 ——p. at 25°C 1+1.5VT I :ionic strength T temperature R :molar gas constant (R =8.314 J/mol.K) F rFaraday constant (F = 96485 Cb/taol) A .£IM10) =59.,6mVatT = 25°C =58.17 mV at T^20°C IT - 12.5 E1 standard potential K? standard constant + Ky .dispropostionation constant of PuO2 _ [Pu][Pu02] V 4 + 3 [H+] [PuO2 ] :disproportionation constant of :KIV= III METHOD III. 1 Introduction Thermodynamic predictions indicate that in HC104 media: + 4 + 1) PuO2+ is stable when [H ] is greater than 0.IM but in this conditions, even if Pu disproportionates, Pu(OH)4 precipitation can interfere with measurement and this become more important when [H+] is smaller. 2) Pu4"""is stable in a very narrow domain around [H+] = 1M, but higher HQO4 concentration should notbe used since CIO4 complexation might occur, and ionic strength corrections are not precise in these conditions. Then, it is possible that at least three oxidation states are in equilibrium in the same aqueous solution. The amounts of each one depend on the experimental conditions. 4+ + We choose to measure Pu and PuO2 disproportionation constants; these constants will be used 2+ + together with the known (Table 1) reversible plutonium redox couples ( PuO2 /PuO2 to finally determine the redox potentials of irreversible plutonium couples. III.2 Experimental section Pu(V) disproportionation equilibrium has been studied : + + 2+ 3+ 3PuO2 + 4H ** 2PuO2 +Pu +2H2O (5) + 2+ PuO2 is prepared by electroreduction of a PuO2 solution in HCIO4 0.1M, on Pt eiectrode. We fixed the acidity at [H+] = 0.1M to lower the disproportionation speed. But, we cannot choose an upper pH value because such condition would facilitate unexpected precipitation of Pu(OH)4, We obtained a + PuO2 solution at 95% of purity. Then, the concentration of each species of pluton'ium is determined spectrophotometrically as a function of time and we calculate the value of the expression : IT - 12.6 3+ 2+ 2 [Pu J[Pu02 j Kv + 3 (6) ~ IH+)4[PuO2 ) When this concentration ratio becomes stable, we suppose that it is the value of the equUibrium constant in this medium (this will be discussed below). We also have studied the Pu4+disproportionation equilibrium : ?: 3+ + 4+ O2 + 2Pu + 4H ** 3Pu + 2H2O 0) 2+ 3+ We mixed a PuO2 solution and a Pu one in the same medium (HC104 1M). The last one was 2+ 4 1 already prepared by clectroreduction cf a PuO-2 solution We use HC1O4 1M solution where Pu " " is quite stable. As previously, we plot versus time the decay or the growth of each species and the value of the expression (8) : With the same hypothesis, after about twenty days, when this concentration ratio is quite stable, we 4+ assume that it is the formal constant value in HQO4 1M. Later the Pu concentration become too low, and the measure too unprecise : so the ratio value diverges and is no longer meaningful. These experiments rely on spectrophotomeiric analysis. Each oxidation state is measured using this technique. The conditions are compiled in the following table : Table 2 Molar Wavelength (nm) Absorption of 830 600 569 470 2+ PuO2 566 11.5 + PuO2 18.5 Pu4+ 15 49.6 Pu3+ 5 37 35 The bold values are the molar absorption for the characteristic narrow absorption bands of each state oxidation. We measured the concentrations at these wave lengths taking into account the possible + 3+ interferences that are compiled in the same table. The most important interference is the PuO2 /Pu IT - 12.7 one, at 569nm that is the wavelength where pentavalent plutonium have the higher molar absorption coefficient ni.3 Treatment of the data To calculate the redox potentials of the irreversible couples, typically E(PuO22+/Pu4+), knowing 4+ 3+ 2 + Kjy (or Ky) equilibrium constant and E(Pu /Pu ) (or E(PuO2 '7PuO2 )) redox potential, we use the two following thermodynamic relauons: 2+ 4+ 2+ + 4+ 3+ 7E (PuO2 /Pu ) = 3E (PuO2 /PuO2 ) - E (Pu /Pu ) +A lgKv (8) 2+ 4+ 4+ 3+ E(PuO2 /Pu ) = E(Pu /Pu )- j lgKjy (9) where the equilibrium constants and [H+] depend of the unit .Because the S.I.T used (Mol/Kg) we have calculated all the formal potential with the same unit. To compare our results with published redox potentials values we have to extrapolate them to the same ionic'strength and specially to 1=0. In acidic medium, the Specific Interaction Theory (S.I.T) can model the activity coefficient of cations with reasonable precision [2] and so, we used this theory to calculate standard potentials with our experimental results. Using this theory the activity coefficient expression is: lg y, = -ZJ2D + Ee(S j) mj (JO) where a, = Y» nij (11) The S.I.T approach assumes that the activity coefficient of a single ion, i, is the sum of two terms: the Debye Huckel term which takes into account the long-range electrostatic interactions, and a second term the short-range, interactions between ions of opposite charges. The concentration of the ions of the ionic medium is much larger than the ones of the reacting species. In our conditions the support electrolyte is NaQ04 and all plutonium ions in acidic medium are cations. The previous expression can be written as : (12) IT - 12.8 Reporting this equation into equilibrium constant expression, one obtains: v Cy d = lgK° + lg C D (13) A YB = lgK° + Az2D-AeI (14) where K(I) is the apparent constant in medium of ionic strength I and K° the standard constant of the following equilibrium: aA + bB **• cC + dD (15) For pentavalent plutonium dismutation constant, Az2=10 and for the tetravalent one Az2=-22. In S.I.T approximation equation (15) is used for linear regression. The plot of lgK(I)-Az2D (16) versus I is a straight line, whose slope is (- Ae) and intercept at I = 0 is lgK°. We have to pointed out that the equation (14) assumes that the equilibrium constant is written in molal unit. Before extrapolate to 1=0 our experimental results, it is necessary to convert the K(I) value from molar unit (we measure directly concentrations in mol/l by spectrophotometry and the pHmeter is calibrated in concentration) to molal unit. The same conversion must be done for the potentials of the irreversible redox reactions which are not dimensionless: relationships such as (9) and (10) points out the need of such conversion. IV RESULTS AND INTERPRETATION IV. 1 PuC>2+ disproportionation We have performed several experiments to measure Ky value at different ionic strengths in (Na+, + H , C1O4) media. The concentration and the apparent constant values are plotted versus time (fig N° 2). Equilibrium is reached within about ten or twenty days (this time seems 10 increase with the ionic strength); a constant value for Ky(I) is then measured (see table 3) whereas the concentrations are still changing.hat leads finally to a pure solution of Pu3+. We assume that this phenomenon is due to radiolysis. IT - 12.9 Table 3 lg(K y) values at different I [KyinCMol/l)-4] 0.1M O.5M 1M 2M 3M 3.42 4.36 4.85 4.88 5.13 Using these values and the S.I.T we calculated lg(K°y) and the activity coefficient The results are in the following table: Table 4 lgCKv°> Ae (Kg/Mol) 2.507 0.32 I V.2 Pu4+ disproportionation We have performed one experiment to measure Kjy value in HC1O4 1M medium. As previously the concentrations values and the apparent constant are plotted versus time until it becomes quite stable.(see fig N° 3). After about 20 days a constant value for Kjy is measured during about 3 days: Table 5 lg(KIV)atI=lM 4 [KjV(MoW) ] -2.06 After this period, Kpy seems to increase; but, as previously (§IV.l), the oxidized plutonium species are then hardly detected and Kjy calculation is not precise. Using these experimental values and the redox potentials of reversible couples at the same ionic strength 2+ 4+ (1= 1M) we can compute the formal potential of PuO2 /Pu (see § IV.3) IT - 12.10 IV.3 Irreversible redox potential calculations The above results (tables 3 and 4) together with the (known) redox potentials of reversible couples (table 1) are used to calculate (see section III.3) E(VI/IV) (table 6) in 1M HC1O4 solution and in standard conditions. The results of the 2 sets of experiments can then be compared (see V.I section) Table 6 ECVT/IV) (mV/NHE) deduced from: Ky values Kjy value OM 1M 1M 1005 1071 1017 In addition, using the same results and equations as those mentioned in § III.3, we have calculated all ihe irreversible redox potentials at different ionic strengths (Table 7). Table 7 I Ecmv) E(VI/III) E(V/IV) E(V/m) (M) (mV) (mV) (mV) (mV) 0 1005 1008 1052 1031 0.5 1048 1001 1163 1058 1 1071 1015 1200 1080 2 1069 1029 1188 1088 3 1083.5 1050 1200 1108.5 They can also be calculated from standard potentials, activity coefficients and thermodynamic relations; but the later procedure would propagate more important errors. V. DISCUSSION V.I Comparison of the two sets of experiments There are 54 mV (table 6) difference between the two E(VI/IV) determinations from Kjy and Ky. This is a little more than the uncertainty that can be calculated from the error propagation rules; but the main uncertainty is induced by a systematic error (that can hardly be predicted) : there is no experimental evidence to know whether equilibrium or a steady state (between reducing species IT - 12.11 produced through radiolysis and Pu redox reactions) is achieved. The effect of radiolysis should finally be to increase (in comparison with equilibrium conditions) the concentrations of the Pu species that can be produced through a reversible reaction from another Pu species whose concentration is still important. Hence, [Pu3+] might be bigger than in equilibrium conditions during Kjy measurement + which is then overestimated, arid E(VI/IV) is underestimated (in table 6). For the same reason, [PuO2 ] might be bigger than in equilibrium conditions during Ky measurement which is then underestimated, and E(VI/IV) is again underestimated (in table 6 and 7). The E(VI/IV) determination is smaller when calculated from Kjv: the corresponding experimental results (table 6) are then consistent with worth equilibrium achievement than during Ky measurement. The intensity of the radiolysis effect probably depends on [H+] and [GO4"] and this later parameter was varied during Ky measurements Since the activity coefficients used to correlate [C1O4] influence (on these measurements) are consistent with published ones, no systematic error seems to interfere : equilibrium conditions were then probably achieved during Ky measurements. V.2 Comparison with published activity coefficients For the daia (table 3) treatment of Ky( (extrapolation to 1=0 using the S.I.T) Ae = 2 e(VI,CIO4-) + e(ffl.ao4-) +2 lg(aH2o) - 3e(V,aO4-) -4e(H+CIO4-) (17) has been fitted (see results in table 8). On the other hand, we compare this experimental value with the estimated one (table 8) using the following approximations [2,3,10,11J: 2+ e(VI,ClO4) = e(U02 ,a04) = 0.46 ± 0.24 3+ e(III,CIO4) = e(Am ,a04) = 0.47 ± 0.03 = -0.01651 and the following values : e(VI,C104-) - e(V,C104-)= 0.29 ± 0.05 E(H+aO4-)= 0.14 ±0.02 Table 8 Ae (eqN°18)(Kg/Mol) Estimated value This work measurement 0.37 0.32 V.3 Comparison with published redox potentials and equilibrium constants IT - 12.12 Rabideau [5] has determined Kjy value in HC1O4 1M (see§ I.2).He corrected his experimental values to take into account the Pu4+ hydrolysis: Table 9 (Rabideau measurement) lg(KIV)(lM) = -2.06 (present work measurement) The two following values are nearly the same even if the experimental methods are quite different Disproportionation constant of pentavalent plutonium has been measured by Kasha [4]. He determined it at I=1M, [HC104] = 0.052 and 0.1019M. We have calculated the mean of these two values and compare it with our measurement: Table 19 lg(Kv) = 5.52 (Kasha measurement) lg(Kv)=4.85 (present work measurement) The 0.7 unit log difference between these two values induces only 20mV (eq N° 9) difference on 2+ 4+ E(PuO2 /Pu ). Rabideau later measurements [6] (see § 1.2), are now reinterpreted to calculate E(V/IV) in HCIO4 1M. From the mass balance equations: [VI]0 + [III)0 = [VI] + [V] + [IV] + [III] (18) 3[III]O=[V] + 2[IV] + 3[III] (19) 3+ 2+ and Nemst Law, we have calculated the potential of a solution where Pu and PuO2 have been mixed: 2+ 3+ In the above equations [VI]0 and [HI]0 are the initial PuO2 and Pu concentrations and [X] the concentration of each oxidation state during the redox potential measurements. From Rabideau's data, we have fitted the following equation: E(IV/IH)-E E - E(V/IV) E-E(WV) [l+io—^—]+io—X—*fi+io —x—] [IU]O = -1+3 E - E(V/IV) (20) 2+10 ^ +3* 10 where the formal potentials of reversible couples are known. IT - 12.13 We obtained E(V/IV) = 975m V This value is 225 rnV different from our measurement (table 7). As already stated (§.1.2) Rabideau waited only 2 minutes before each redox measurement which now seems much too short in comparison with the time needed to achieve equilibrium. Toth [7] have measured the Pu4+ disproportionation constant in different media and at 25,15 and 5°C. We have calculated the mean of all the values determined at 1=1 and 25°C to compare with the others results (see above). We obtained : lg(Kjy) = -2.7 Toth measurement This value is less than 0.7 Ig unit from ours (Table 5), still Toth did not use proper data treatment and experimental procedures (§1.2) IT - 12.U BIBLIOGRAPHY [1J: Connick R.E, Kasha M, Me Vey W.H, Sheline G.E "The Transuranium Elements" - Natl. Nuclear Energy Ser IV 14B- Mc Graw- Hill Book Company INC. (1949) p.227 [2]: Riglet Ch, Robouch P, Vitorge P Radiochimica Acta 46_, 85-94 (1989) [3J: Riglet Ch, Vitorge P, Grenthe I Inorganica Acta, 122.323-329 (1987) [4]: Kasha M "The Transuranium Elements" - Natl. Nuclear Energy Ser IV 14B- Mc Graw- Hill Book Company INC. (1949) p.295 [5]: Rabideau S.W J.Am.Chem.Soc. 25, 798801 (1953) [6]: Rabideau S.W J.Am.Chem.Soc 71.2705 (1956) [7]: Toth L.M, Bell J.T, Friedman H.A Radiochimica Acta 42,193-199 (1990) [8]: Nelson D.M, Orlandini K.A Argonne Natl. Lab. Annu. Rep., ANL-79-65 part 111:57-59 (1979) [9]: Bondietti E. A, Trabalka J.R Radiochem.Radioanal.Lett.. 4.2,169-176 (1980) [10]: Grenthe I. Lemire R, Muller A, Nguyen-Trung C, Wanner H Chemical Thermodynamics of Uranium TDB-NEA-OCDE to be published (1991) [11]: Riglet Ch " Chimie du Neptunium et autres actinides en milieu carbonate" C.E.A N° 5535 (1990) IT - 12.15 E-FH DIAGRAM OF PLUTONIUM SOLUBLE SPECIES PuO (OH) 2 3 i I I i l ii I I I 1 I I I I I I I 4 6 8 10 12 14 - Rgure NJ° 1 - IT - 12.16 + PaO2 DISPROPORTIONATION Perchloric medium (HC104,NaClQ4) [K+] «O,1M Ionic strength I-0.5M Measurement of eacfe oxidation state concentration by tpectrophotoraetry and versus time I g 20 40 60 Time (days) Evolution versus time of the apparent constant K(V) 3 5E-04 4E-04 n 3E-04 2E-04 a 1E-04 B an rnn^ • D • • OE+00 i i i i 20 40 60 Time (days) -Figure N°2- IT - 12.17 + PDO2 DISPROPORTIONATION Perchloric medium (HClO4,NaClO<) [H+] «O,1M Ionic strength I-0.5M Measnrentent of each oxidation state concentration by spectrophotomctry and versus Ume 1,0 0,9 0,8 0,7 2+ 3 PuO2 0,6 j 0,5L 3 0,4 • c 3 8 0,3 0,2 0,1 0,0 c 3 i : i 20 40 60 Time (days) 20 40 Time (days) -Figure N°2- IT - 12.18 Pu4+ DISPROPORTIONATiON Perchloric medium HC1O4 = 1M Measurement of each oxidation stac concentration by spectrophotomctry and versus time 1,8 2 1,6 S 1,4 ie s a. 1,2 (0 ft- 1.0 o 0.8 o 0.6 0,4 en t u 8 0,2 0.0 60 -l Evolution versus time of the apparent constant 1CIV I 30 -Figure N°3- IT - 12.19 PLUTONIUM FUEL FABRICATION AT BARC Ganguly C. IT - 13 PLUTONIUM FUEL FABRICATION AT BARC C. Ganguly Radiometallurgy Division Bhabha Atomic Research Centre Bombay 400 085, India ABSTRACT Fabrication of plutonium fuel is a challenging task mainly because of high radiotoxicity and criticality hazards associated with plutonium. In addition, plutonium metal, carbides and nitrides are highly susceptible to oxidation and hydrolysis and are pyrophoric in powder form. Plutonium fuel fabrication in India was initiated nearly two decades ago when the core for the PURNIMA research reactor at BARC, consisting of SS 316 clad PuOz pins was prepared at Radiomataliurgy Division (RMD). This was soon followed by fabrication of Al clad Al - Pu plate fuel elements and sub-assemblies, the experience of which was recently utilised for production of Al clad Al-23% Pu plate fuel subassemblies for PURNIMA III. During the last 15 years extensive research and development have been carried out on fabrication of plutonium bearing ceramic fuel pellets at RMD. The fuels developed so far include: (i).(U,Pu)O2 and (Th,Pu)O2 containing upto 6% Pu for pressurised heavy water reactors (PHWR), (ii) mixed uranium plutonium oxide (MOX), monocarbide (MC) and mononitride (MN) containing between 20%and nearly 70% Pu for liquid metal cooled fast breeder reactors (LMFBR). A sophisticated carbide laboratory has been set up, where the fabrication of (Uo.aPuo.7)C fuel pellets and pins for the first core IT - 13.1 of fast breeder test reactor (FBTR) has been successfully completed. Presently, the facility is being utilised for development of MOX, MC and MN fuels for the forthcoming prototype fast breeder reactor (PFBR). Conventional "powder-pellet" and advanced sol-gel microsphere pelleti3ation (SGMP) processes have been developed for fabrication of MOX, MC and MN pellets. The SGMP route is highly suitable for automation and remotisation and is more attractive because it minimises process steps, radiotoxic#dust and fire hazards (for MC & MN) and fabrication cost and leads to sintered fuel pellets of controlled dsnsity, pore structure and excellent nicrohomogeneity. 1. INTRODUCTION Fabrication and handling of plutonium (mainly Pu239 ) bearing nuclear fuels necessitates special considerations because of the high radiotoxicity, biological behaviour, fissile properties and criticality hazards of plutonium, the man-made 50 years young element. In fact, this handling problem makes plutonium alloy and ceramic fuel fabrication a distinctly different and much more formidable task when compared with that of other conventional metals and ceramics including the natural uranium and thorium based ones. Any,plutonium bearing material is handled in a well ventilated laboratory within tho confines of hermitically sealed glove boxes through alpha (particles) tight neoprene or butyl-based gloves. The glove box is maintained under slightly negative pressure under a dynamic flow of air or high purity inert gas (in case of plutonium metals, carbide and nitride which are highly susceptible to oxidation and hydrolysis) in order to have 3-10 box volume changes per hour. The external hazard of plutonium is mainly from gamma and neutron radiations embedded by higher and other isotopes of plutonium (Pu240, Pu24», Pu242 and Pu238) which are present in significant amounts in plutonium generated from high burn-up spent U238 based fuels. In order to keep the radiation exposure of operators below permissible limits specified by the International IT - 13.2 Commission of Radiation protection (ICRP), restrictions are imposed on the quantity of plutonium beoring material that has to be handled at a time and great emphasis is given towards automation and remotisation in fuel fabrication. Further, special care is taken to avoid plutonium bearing powder and dust build up on equipment and glove box surfaces and often external or localised composite shields are used for protection against gamma and neutron radiations. In addition, since the critical mass of plutonium and its compounds are relatively low (a few kilogram) only small quantity of these materials are handled at a time to avoid spontaneous self-sustaining nuclear chain reaction or what is known as a critical accident which could be catastrophic to the personnel and their surroundings. The crucial role of plutonium in the 3-stage nuclear power programme of India, involving the fuel cycles of pressurised heavy water reactors (PHWR) and the liquid metal cooler fast breeder reactors (LMFBR), for unlocking the huge energy reserves of our vast thorium (360,000 tons) but limited uranium (70,000 tons) resources has been recognised' long back. Accordingly, Radiometallurgy programme was initiated at BARC in the early 60s and small laboratory facilities for fabrication of plutonium bearing ceramic and alloy fuels Were set up in the early 70s. Three m^jor jobs were successfully completed in this facility in the initial years. First, the fabrication of high density PuOz pellets from o^alate derived PuO2 powder by cold-pelletisation at around 350 MPa followed by sintering at 1873K in Ar+8%Hz. The PuO2 "pellet-stacks" were encapsulated in 2s 316 cladding tube by TIG welding. Nearly 200 fuel pins were fabricated for the PURNIMA I zero energy research reactor core at BARC [1]. Simultaneously, Pu-Be neutron sources (point and pencil) were fabricated by berylothermic reduction of tableted PuO2 and Be powder mixture [2]. The neutron sources were double encapsulated in aluminium and stainless steel cans. This was soon followed by fabrication of aluminium clad Al-10% Pu and Al-18% Pu plate fuel elements and sub-assemblies by the "melting-casting- rolling-picture framing-roll bonding" route for reactor physics experiments at ZERLINA research reactor at BARC [3]. IT - 13.3 These initial fabrication campaigns on laboratory scale, laid the foundation for taking up the development and production activities of mixed uranium plutoniurn oxide and the more difficult mixed uranium plutonium monocarbide (MC) and mononitride (MM) ceramic fuels on kilogram scale for irradiation-testing experiments and for fabrication of driver fuel for the fast breeder test reactor (FBTR). The expertise of Al-Pu alloy fuels was recently utilised for fabrication of plate fuel elements and sub-assemblies for PURNIMA III criticality facility, commissioned recently at BARC. The conventional and advanced fuels for PHWR and LMFBR are listed in Table 1. The advanced fuels are those which have not been demonstrated commercially but have the potential of introducing significant advantages in established or new reactors or their fuel cycles. The advanced fuels aim at judicious utilisation of natural uranium and thorium resources, ensuring high burn up and high linear and high specific power ratings without failure, high conversion or breeding ratio and in turn short doubling time. In addition, these fuels should be easy to fabricate, should dissolve easily in HNO3 to facilitate reprocessing of spent fuel and should have minimum radioactive waste generation and disposal problem. Presently, in India only ceramic fuels are being pursued for PHWRs and LMFBRs. The advanced fabrication methods of both conventional and advanced fuels focus mainly on reduction in operator dose by remote and automated processing. The other objectives are minimisation of process steps and fabrication costs and improvement in microstructure of fuel for improved burn up. Table 2 summarises the guidelines for advanced methods of ceramic fuel fabrication. The present paper summarises the author's recent fabrication experience of the following plutonium fuels, mainly at Radiometallurgy Division, BARC: Al-clad Al-Pu plate fuels (U,Pu)O2 and (Th,Pu)Oz "pellet-pins" for PHWR (U,Pu)C and (U,Pu)N "pellet-pins" for LMFBR. IT - 13.4 2. Al-Pu PLATE FUEL ELEMENTS FOR PURNIMA III Three sub-assemblies of Al clad Al-23% Pu plate fuel elements were fabricated for the PURNIMA III criticality facility at BARC, following the flowsheet shown in Figure 1. The photographs of the Al-Pu fuel plate and sub-assembly at different stages of fabrication are shown in Figure 2. Each sub-assembly consisted cf 8 fuel plates of overall dimension 260mm x 62mm x 2mm. Each fuel meat contained approximately 9 g plutonium and were 250 mm in length, 55 mm in width and 1 mm in thickness. Small (1-2%) addition of Zr was found to avoid the dendritic platelets of brittle PuAl4, stabilise the relatively softer PuAla phase and improve the microstructure. Figure 3 shows the microstructures and X-ray diffraction patterns of Al-Pu cast alloys with and without Zr addition. The ternary Al-Pu-Zr alloys could be easily rolled from 25mm thick slab to 3.5 mm thin strip without edge cracking. The x-ray radiograph of the roll-bonded fuel plates after trimming (Figure 4) shows the outline of the Al-Pu fuel meat in the Al cladding. The radiographs were subjected to microdensitometric scanning in order to confirm the homogeneous distribution of plutonium in the fuel meat. 3. (U,Pu)O2 AND (Th,Pu)Oi FUELS FOR WATER COOLED REACTORS As part of plutonium recycling programme in pressurised heavy water reactors (PHWR), high density (U,Pu)O« and (Th,Pu)O2 pellets containing upto 6.7% PuOz have been fabricated. The (U,Pu)O* pellets were prepared by both 'powder-pellet' and sol-gel-microsphere- pelletisation (SGMP) processes. The (Th,Pu)Oz pellets have so far been prepared only by the powder-pellet route, though the SGMP process has been successfully demonstrated earlier [4] for preparation of ThO2-4%CeOa pellets, where vCe' was used to simulate Pu' . 3.1 Powder-Pellet Route In the powder-pellet route followed at RMD for fabrication of IT - 13.5 TI1O2 - and UO2 -based mixed oxide fuels [5,6], the starting materials were PUO2 and TI1O2 powders obtained by calcination of their oxalates and UO2 powder obtained by ex ammonium diuranate (ADU) route. The UO2 or ThOa powder is co-milled with PUO2 powder, precompacted at around 140 MPa, granulated and finally compacted at around 350 MPa by admixing zinc behanate or zinc stearate lubricant. The mixed oxide pellets were dewaxed at around 825K and sintered to high density pellets at around 1973K for 2-4 hours in Ar+8% or N2+8% Ha atmospheres. For ThO2 based fuel, small amount of MgO served as a sintering aid. The MgO was doped to the thorium nitrate solution before precipitation of the oxalate. Small amount of TiOa addition was found to increase the grain size of UOE-4%PuOa from 5-15 um to 30-40 um which is desirable for PHWR fuel pins for retention of fission gases. The UO2-4%PuO2 pellets could also be sintered to very high density by low temperature and short duration oxidative sintering (LTS) in CO2 atmosphere at 1473K for 1 hour followed by reduction in Ar+8%H*. Figure 5 shows a few representative microstructures of (U,Pu)O2 and (Th,Pu)O2 fuel pellets fabricated at BARC. Some of these fuel pellets were later centreless ground and loaded in zircaloy cladding tube and subjected to irradiation testing in the pressurised water loop of CIRUS research reactor, simulating the operating conditions of PHWR. Both UO2-4%PuOz and Th2-4%PuO2 fuel pins could be successfully irradiated to burn-up of 20,000 MWD/T without failure. 3.2 SGMP of UOi-4%PuOi High density UO2-4%PuC-2 pellets, meeting the specifications of PHWR fuel, was also prepared by the sol gel microsphere pelletisation (SGMP) route [7]. Fuel Chemistry Division of BARC was responsible for preparing the gel-microspheres of hydrated UOa+PUO2 IT - 13.6 utilising their internal ammonia gelation process and facility [8]. On the basis of earlier experience on SGMP of ThO2 - and UO2-based mixed oxide using external gelation process [9,10], a mixed Plutonium uranyl nitrate solution of lower concentration (£1.2 molar) was used. This was mixed with the ammonia generator hexamethylene tetra amine (HMTA) and urea in the following volume proportion and cooled: (U, Pu)nitrate =» 1.1 molar HMTA =1.65 molar Urea = 1.65 molar Carbon black pore former ( 30 g/molar heavy metals) was added to the solution mixture prior to gelation. The gelation of the fuel microspheres was carried out in silicone oil bath at 263±2 K, when HMTA decomposed and released NH3 that caused gelation. The carbon black was later removed from gel-microspheres by controlled air calcination at around 1000K. The air calcination was followed by Nz+H2 treatment at 1000K and stabilization in CO2. Thus, dust-free and free-flowing porous UO2-4%PuO2 microspheres of specific surface area in the range.of 5-10 ma/g were produced which crushed easily and lost their individual identity during pelletisation at 300 MPa. The pellets could be sintered to high densities (£96% T.D.) at 1923K for 4 hours. Figures 6(a) shows the microstructure of high density (U,Pu)O2 pellet prepared from 'porous' microspheres by the SGMP route. If carbon black pore former is not added, the microspheres were 'non-porous' and retained their individual identity during pelletisation leading to undesirable "black berry" structure (with lot of open porosity) as shown in Figure 6(b). Such a microstructure, though, favourable for use as LMFBR fuel, is not acceptable as PHWR fuel because it facilitates fission gas release which is difficult to accommodate inside a PHWR fuel pin which has no gas plenum and uses collapsible cladding. The microstructure of the mixed oxide pellets' prepared by the SGMP route showed uniform distribution of closed, spherical porosity in the optimum diameter range of 2-5 micron. Such uniform spherical porosity was responsible IT - 13.7 for a somewhat higher thermal conductivity of (U,Pu)02 pellet prepared by the SGMP route compared to the one of equivalent density prepared by the conventional powder-pellet route as shown in Figure 6(c) . UO2-4%PuO2 pellets prepared by SGMP could also be sintered to high density by low temperature oxidative sintering in CO2 at 1473K, followed by treatment in N2+8%Hz for achieving the desirable oxygen to metal ratio [11]. The combined SGMP-LTS route is most advantageous for fabrication of high density pellets for water cooled reactors since it not only minimises radiotoxic dust hazard and facilitates automation and remotisation by using dust-free and free-flowing microspheres but brings down the fuel fabrication cost significantly by reducing the sintering temperature and time and utilising less expensive sintering gas atmosphere. 4. (U,Pu)O», (UrPu)C AND (U,Pu}W FUELS FOR LMFBR Mixed uranium plutonium oxide containing between 20 and 30% Pu is the reference fuel for all operating and forthcoming LMFBRs in the world, including the forthcoming prototype fast breeder reactor (PFBR) 500 MWe in India. All aspects of the MOX fuel cycle, namely, fabrication, irradiation to high burnup (£10 a/o), reprocessing and refabrication, have been established globally on an industrial scale. So far, more than 150 tons of mixed oxide fuels have been fabricated and irradiated in LMFBRs all over the world. The oxide fuel has however the limitation of low thermal conductivity and low heavy atom density which are responsible for its long doubling time. Because of this disadvantage, the choice of mixed oxide fuel in commercial LMFBRs is vulnerable. Mixed uranium plutonium monocarbide (MC) and mononitride (MW) have been recognised as advanced LMFBR fuels on the basis of their high thermal conductivity, high heavy atom density and excellent compatibility with sodium coolant. Though the fabrication experiences of MC and MN fuels all over the world are less than 500 kg and 150 kg IT - 13.8 respectively, both He- and and Na-bonded MC and MN test fuels have demonstrated high burnup, comparable to that of the oxide. The U-Pu- Zr metallic fuel has been developed recently by the Argonne National Laboratory, USA and has demonstrated satisfactory performance upto high burnup. However, the metallic fuel development is restricted in USA only. As a first step to the LMFBR programme in India, a fast breeder test reactor (FBTR) has been commissioned at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam in October 1985 utilising the indigenously developed plutonium rich mixed uranium Plutonium monocarbide fuel of composition (Uo.3Puo.7)C. Presently, only a limited number of sub-assemblies are used in FBTR which can take the reactor to a maximum power of 10 MWt. The full core of FBTR consists of 65 fuel sub-assemblies which can produce 40 MWt. For the full core the fissile material enrichment would be somewhat lower (555% Pu). Table 3 summarises the fuel composition of interest to LMFBR programme in India. The mixed uranium plutonium mononitride belongs to the same family of non-oxide fuels as the monocarbide and has the same advantages of high thermal conductivity, high heavy atom density and short doubling time. In addition, the mononitrides are less susceptibe to oxidation and hydrolysis compared to the monocarbide and dissolve easily in nitric acid, thus minimising the fuel fabrication and reprocessing problems. The mononitride has the disadvantages of radioactive C14 formation by n,p reaction with N»«. But, this could be overcome by either using N1s in place of N1« , which is however expensive, or making adequate provision for trapping C1* as a highly active solid waste by converting it to calcium carbonate via CO2 and burying the same. 4.1 (U,Pu)Oz Pellets UO2 and PUO2 are isostructural, completely solid soluble and have very similar thermophysical and thermodynamic properties. IT - 13.9 Hence, on the basis of the excellent peiformance record of UO2 and (Uo . e Puo . ;• ) O2 fuels all over the world, the initial choice of fuel for FBTR was (Uo . 2 « Puo . 7 e ) O2 because of nonavailability of enriched uraniun. indigenously. Unfortunately initial metallurgical investigations revealed that this plutonium rich mixed oxide is neither compatible with sodium coolant nor it is possible to fabricate this as single phase mixed oxide keeping the oxygen to metal ratio between 1.98 and 1.99 and pellet density between 90 and 92% T.D. Hence, plutonium rich mixed oxide was not pursued for FBTR. For PFBR application, relatively iow density (£85% T.D.) (Uo . BPuo.2)O2 fuel pellets containing predominantly open porosity is recommended. In the powder-pellet route of fabrication of these fuel pellets, polyvinyl alcohol (PVA) or methyl cellulose pore formers are admixed with the UO2 and PUO2 powders and later removed during dewaxing of the green pellets at low temperature (5873K) in vacuum or reducing atmosphere. UaOB is often used as an addition pore former. Uranium rich mixed oxide fuel pellets for PFBR could also be fabricated by the low temperature oxidative sintering route using COs at 1473K for ': hour for sintering followed by Ar-H2 gas mixture at the same temperature for reduction and obtaining the desirable oxygen to metal ratio. Mixed uranium plutonium oxide pellets, using cerium as a simulator for Pu, has been fabricated successfully by the sol gel microsphere pelletisation process, adapting the external gelation of uranium technique for production of gel microspheres. No pore former was added to the broth prior to gelation. The gel microspheres of hydrated mixed oxide were dried in a continuous belt driven oven at 473K for 15 minutes followed by calcination in Ar+8%H2 at around 973K, cold compaction at 350 MPa and sintering in C02 (1473K) or Ar+8%H2 (1923K) for obtaining relatively low density (85% T.D.) mixed oxide pellet with a predominantly blackberry structure and open porosity as shown in the microstructures in Figure 7. The pores were uniformly distributed and were in the ideal size range of 2-5 micron which would prevent in-pile densification. IT - 13.10 4.2 (U,Pu)C and {U,Pu)N Fuels for FBTR and PFBR Stainless steel (type 316) clad, slightly hyperstoichiometric (Uo.3PU0.7)C pellets of density 90±2% T.D. and containing less than 7500 ppm of resid'ial oxygen and nitrogen impurities and between 5- 15% sesquicarbide was fabricated at RMD and was used in FBTR [12,13]. Such an advanced LMFBR driver fuel in FBTR with hitherto unknown plutonium rich composition, for the first time in the world. Fabrication of MC and MN fuels is difficult compared to MOX because of several reasons. First, these non-oxide fuels are highly susceptible to oxidation and hydrolysis and pyrophoric in some cases (MC for example) necessitating high purity inert (N2 or Ar) cover gas in glove boxes. Secondly, the number of process steps are more than double if "powder-pellet" route starting with UO2 and PUO2 powders as feed back materials are used for fabrication of MC and MN pellets as shown in 7igure 8. Thirdly, stringent control of oxygen content and carbon and or nitrogen stoichiometries are needed to avoid metallic second phase and keep the higher carbides or nitrides within acceptable limit. The essential process steps in fabrication of MC and MN pellets are as follows: carbothermic reduction of tableted oxide-carbon powder mixture in vacuum and flowing N2 for monocarbide and mononitride respectively crushing and milling of MC and MN clinkers followed by cold pelletisation and sintering in Ar+8%Hz. The overall chemical reactions involved during carbothermic reduction for synthesis of MC and MN are as follows: UOa + PuOa + 5C = (U,Pu)C + 4CO UO2 + PuO2 + 4C + fcNis = (U,Pu)N + 4CO IT - 13.11 For synthesis of MC, the vacuum and temperature in the furnace during carbothermic reduction are optimised in such a way that: (i) the conversion of oxide to carbide is complete, (ii) the oxygen and sesquicarbide contents in MC are within acceptable limits, and (iii) Plutonium volatilisation loss is minimum (51%). For synthesis of mononitride, nitrogen flow rate and sequential use of N2, N2+8%Hz and Ar+8%H2 is essential to keep the residual oxygen and carbon impurities at minimum levels and avoid the formation of the higher nitrides [14,15]. The nitrogen flow rate should be as high as possible (£1000 litre/hour) because nitrogen not only acts as a reacting gas but is also responsible for purging the reaction product CO from the reaction zone for completion of thfi mononitride synthesis reaction. The time-temperature profiles of carbothermic synthesis of monocarbide and mononitride for batch size of -1 kg are shown in Figure 9. The preliminary irradiation-testing experiments of both high (£94% T.D.) and low (585% T.D.) density MC and MN "pellet-pins" in EBR II reactor, USA have led to the following "pellet-pin" specifications: Helium bonded pins should contain relatively low density pellets - Sodium bonded pins should contain high density pellets and shroud tube. Accordingly, both high and low density MC and MN pellets were fabricated at BARC. For high density, 0.4% freshly reduced *Ni' and - 1% MOa powders were used as sintering aid for "MC and 'MN1 respectively. Figure 10 shows the microstructure of monocarbide and mononitride fuel pellets for FBTR and PFBR prepared by the "powder- pellet" route. The SGMP route is definitely more attractive for fabrication of IT - 13.12 MC and MN pellets because of the following incentives: the number of process steps are less than half compared to powder-pellet route as shown earlier in Figure 8 generation and handling of fine MC and MN powders are avoided, thus minimising pyrophoricity hazard along with radiotoxic dust hazard the dust-free and free-flowing fuel microspheres are ideal materials for remote and automated pellet fuel fabrication the optimum carbothermic reduction temperatures for synthesis of MC and MN are lower because of the high chemical activity and high specific surface area of the gel microspheres which favour the kinetics of the carbothermic reduction excellent microhomogeneity is ensured since the heavy metals are mixed in the form of nitrate solution. The MC and MN pellet* thus prepared show the desirable blackberry structure (Figure 11) with open porosity. The microspheres retained their individual identity during pelletisation arid sintering [14,15]. Figure 12 shows the recommended flowsheet for fabrication of MOX, MC and MN fuels via the SGMP route. 4. CONCLUSION 4.1 For Al-Pu plate fuel fabrication, it is essential to add -1% Zr to stabilise the relatively softer PuAl3 phase and minimise the dendritic platelets of the brittle PuAl< phase, in order to avoid edge cracking during hot or cold rolling. 4.2 For fabrication of high density (U,Pu)02 and (Th,Pu)O2 fuels for water cooled reactors, the SGMP process is an attractive alternative compared to the conventional powder-pellet route. In IT - 13.13 case of (U,Pu)Oz the SGMP route could be combined with low temperature oxidative sintering in order to reduce fuel fabrication cost. 4.3 For fabrication of (U,Pu)O2, (U,Pu)C and (U,Pu)N fuels for LMFBR, the SGMP process is far superior to the conventional powder- pellet route, particularly for MC and MN, because of lesser number of process steps, minimisation of radiotoxic dust and pyrophoricity hazards and excellent microstructure and microhomogeneity of sintered pellets. REFERENCES 1. P.R. Ps.oy and C. Ganguly, "Plutonium metallurgy in India", Bull. Mater. Sci. Vol. 6, No. 5, Sept. (1984), pp 923-958. 2. V.K. Mahajan, C. Ganguly, M.S. Ramakumar, P.R. Roy and V.K. Moorthy, "Fabrication of plutonium beryllium neutron sources", BARC- 629 (1972). 3. G.J. Prasad, C. Ganguly and P.R. Roy, "Fabrication of aluminium clad Al-Pu alloy fuel plates", BARC/I-423 (1976). 4. C. Ganguly, "Sol gel microsphere pelletisation process for fabrication of conventional and advanced ceramic nuclear fuels". Metals, Materials and Processes, Vol. 1, No. 4 (1990), pp 253-274. 5. C. Ganguly and U. Basak, "Fabrication of ThO2-4%UOz and ThO2 - 4%PuO2 fuel pellets for pressurised heavy water reactors", Trans. PMAI, Vol. 13 (1986), pp 105-111. 6. C. Ganguly, H.S. Kamath and P.R. Roy, "Fabrication and characterisation of ceramic nuclear fuels in BARC", Proc. of Indo- German Seminar on Trends and Techniques in Modern Materials Research, IGCAR, Kalpakkam, Jan. 19-22 (1987), pp 251-271. 7. C. Ganguly, U. Basak, V.N. Vaidya, D.D, Scod and P.R. Roy, "Sol-gel-microsphere pelletisation of UO2 and UO2-PUO2 pellets of PHWR fuels specifications using internal gelation process", Proc. of Second Int. Conf. on CANDU Fuel, Oct. 1-5, 1989, Pembroke, Canada, Published by Canadian Nuclear Society (1989), pp 108-124. 8. V.N. Vaidya, S.K. Mukherjee, J.K. Joshi, R.V. Kamat and D.D. Sood, "A study of chemical parameters in the internal gelation based sol-gel process for uranium dioxide", Jl. Nucl. Materials, 148 (1987), pp 324-331. IT - 13.14 9. C Ganguly, "Sol-gel microsphere pelletisation process for fabrication of high density ThCh blankets and ThO2-based PHWR fuels", Trans. IIM, Vol. 41, No. 3, June (1988), pp 219-230. 10. C. Ganguly, U, Linke and E. Kaiser, "Characterisation of (U,Ce)O2 pellets prepared by the sol-gel microsphere pelletisation process". Metallography, Vol. 20 (1987), pp 1-14. 11. C. Ganguly and U. Basak, "Fabrication of high density UOz fuel pellets involving sol-gel microsphere pelletisation and low temperature sintering", Jl. of Nucl. Mater. (1991) (in press). 12. C. Ganguly, G.C. Jain, P.V. Hegde, U. Basak, R.S. Mehrotra, S. Majumdar and P.R. Roy, "Development and fabrication of 70%PuC-30%UC fuel for the fast breeder test reactor in India", Nuclear Technology, Vol. 72, Jan. (1986), pp 59-69. 13. C. Ganguly, G.C. Jain, J.K. Ghosh and P.R. Roy, "The role of process control and inspection steps in the quality assurance of SS 316 clad plutonium-uranium carbide fuel pins for FBTR", Jl. of Nucl. Mater. Vol. 153, (1988), pp 178-188. 14. C. Ganguly, P.V. Hegde and "... Sengupta, "Status of (U,Pu)C and (U,Pu)N fuel development in BARC", Proc. of IAEA Technical Committee Meeting on advanced fuel for fast breeder reactors, Vienna, Nov. 3-5, 1987, IAEA-TECDOC-466 (1988), pp 7-23. 15. C. Ganguly, P.V. Hegde and A.K. Sengupta, "Fabrication, characterisation and out-of-pile property evaluation of (U,Pu)N fuel pellets", Jl. of Nucl. Mater. (1991) (in press). IT - 13.15 Table 1 : Conventional and advanced fuels for PHWR & LMFBR programme in India Reactor Conventional Fuel Advanced Fuel Natural UO2 (i) UO2-PUO2 (£1%) (ii) ThO2-PuO2 (54%) 233 (iii) ThO2-U O2 (52%) - high smear density pins - high smear density pins containing containing high density high density pellets PHWR (£96% T.D) pellets - burn up : -6500 MWD/t - burn up : 12,000 - 18,000 MWD/t - conversion ratio : 0.6 - 0.8 - conversion ratio : 0.9-1.0 for (Th,U)O2 - fabrication : in controlled area - fabrication : in controlled area inside ot-tight glove box with p-jf & neutron shieldings UO2-Pu02 (525%) (U,Pu)C, (U,Pu)N and U-Pu-Zr [Pu 520%] - low smear density He-bonded - low smear density He-or Na-bonded pins: pins, containing low density 0 He-bonding for vi-pack or pellet-pins LMFBR (£85% T.D.) pellets with low density (585% T.D.) pellets (commercial! 0 Na-bonding for "alloy pins" or "pellet-pins" containing high density U96% T.D.) pellets - burn up : £100,000 MWD/t - burn up : £100,000 MWD/t - breeding ratio : 51.1 - breeding ratio : 1.3 - 1.4 Table 2 : Guidelines for advanced methods of fabrication of ceramic nuclear fuels Safety Economics Performance • Avoid fine powder • Minimise process steps • Improved microstructure for higher burn up - for minimising radio- • Reduce fuel synthesis & toxic dust hazard sintering temperatures - large (^ 40 um) grain size (for PHWR) - for minimising fire • Reduce gr.s cost hazard (for carbide) - high density (>>96% T.D. ) OJ - recirculation & & "closed1 pore (for PHWR) • Automation & remotisation purification - low density (4 85% T.D.) & - for minimising personnel - use less expensive gas 'open1 pore for He-bonded exposure to radiation LMFBR pin • Reduce process losses & rejects - high density {>/ 96% T.D.) & 'closed' pore for Na- bonded LMFBR pins - microhomogeneity Table 3 : Fuel composition of interest to LMFBR programme in India Reactor Oxide fuel Monocarbide Mononitride composition fuel composition fuel composition FBTR Not applicable 10 MWt because of (Uo s'PUo . ri )C (Uo 3Puo . 7 )N poor chemical compatibility of- plutonium 40 MWt rich mixed (Uo 4 a Puo oo )C (Uo 4 0 PUO . 8 8 )N oxide with sodium coolant PFBR (Uo . sPuo . a )C (Uo 8 PUO . t)! C (Uo B Puo . a ) N 500 MWe IT - 13.18 ALUMINIUM Al COMPC NENTS —• MAST:' ALLOY —- • CHEMICAL ANALYSIS I f (Pu. Zf t IMPURITY) MELTING t CASTING • RADIOGRAPHY (PIPING I SEGREGATION) HOT I COLD-ROLLING L • CHEMICAL ANALYSIS (Pu. Zr t IMPURITY) 4 • RADIOGRAPHY CORE PREPARATION (Pu SEGREGATION! PICTURE FRAMING 1 ROLL-BONDING I ) BLISTER TESTING, ULTRASONICS & RADIOGRAPHY (BOND QUALITY a CORE 1 LOCATION) TRIMMING t MACHINING RADIOGRAPHY.VISUAL t METROLOGY CPu DISTRIBUTION t DIMENSIONS) ALUMINIUM ROLL-SWAGING SPACERS • VISUAL I METROLOGY FABRICATED (SURFACE CONDITIONS SUB-ASSEMBLY I DIMENSIONS) COMPONENTS SUB-ASSEMBLY J • VISUAL I METROLOGY RFACE CONDITIONS I DIMENSIONS) Figure 1 : Process flowsheet followed at RadiometaJlurgy Division, BARC for fabrication of Al clad Al-2?% Pu plate fuel elements and sub-assemblies IT - 13.19 encapsulated fuel meat coverplates ligtire / : All'u plate fuel it different stages of fnbri ia I ion u*tixars*:i. Mag.9OX 36 20 • Fig.3: Microstructures and X-ray diffraction patterns of cast alloys of Top : M - 23% Pu ; Bottom : Al - ?3l Pu - 1* Zr IT - 13.21 88y61mg>tm* 72.69 ma/cm w*«*ww*fa^^ I> Figure 4 : X-ray radiograph and microdensitometric scan of a typical Al clad Al-23% Pu fuel plate for PURNIMA III IT - 13.22 X^^:&*V^?W Figure S : Microst rue ture of mixed oxide fuel pellets for PHWR containi ng ~ 4¥> Hu Top : '1'iO-, doped large grain (25 ju) (U,Pu)O2 Bottom : MgO d«>ped (Tli,Pu)O? IT - <• SGMP ROUTE o POWDER ROUTE E >- > O Q oZ o CC Id I 973 1073 1173 1273 1373 1473 1573 1673 1773 1873 TEMPERATURE (K|- Figure 6 : Microstructure and thermal conductivity of (U,Pu)0 pellets containing 4% Pu fabricated by SGMP route Top left : using porous microspheres (no boundaries) Top right : using non-porous mirrospheres(showing boundaries) Bottom : thermal conductivity of pellets prepared by SGMP and powder-pellet routes IT - 13.2A Figure 7 : Mft-rciht nu I uri- >>f ( 1.-, ("e )l> ^ [/t-ll(-1.s containing 30% Ce and i ah r I .'.[ I'lll I IJ !'•• Hdlicii: : .'-I'M i>;.- Mire "I IK;- I i n. t uiT'd surface .-.i-r'.uiif 'iiiili.rii << i L.: ri I 'lit .'mi of porosity I " ! I w • i I * l! i: ill ,' "' [ I' ! t.'II POWDER PELLET ROUTE SGMP PROCESS Step U-nitrate Pu-nitrate Steps |_ U-nitrate + Pu-nitrate 1.2 ADU or AUC Oxalate Preparation of Sol/Broth -{Carbon- Black| Calcination CCalcinatioa n Internal / External 3.4 (UP? powder) yPuCfr powder] Ammonia Gelation Washing & Drying Mixing and Grinding I 3,4 r (Ge/-microspheres of UO^+PoO +C) Tableting (UO? + P11O7 + C) Carbothermic Reduction Carbothermic Reduction vac,]873K,4h for MC clinkers - vac,1773K,4h for MC I Micro- spheres for MN clinkers - N?I177?K,3?h for MN Crushing Milling 10 Prerom pact ion 1 Binder"] 11 Granulation U Pelletisation Pelietisation 13 Dewaxing 14 Sintering Sintering 15 I Pellet-pin encapsulation r Pellet-pin encapsulation Figure 8 : Process steps in fabrication of (U,Pu)C and (U,Pu)N fuei "pellet-pins" via "powder-pellet" and "sol-geJ mirrosphere pelletisation" (SGMP) process* IT - 13.26 ._ For synthesis of (U,Pu)N (1825K, gas flow 1.2 m /h) • For synthesis of (U,Pu)C (1750K, vacuum 1 Pa) 2100 1800 - 1500 - 1200- 900 t* Ar+8%H • 2 a, 20 28 36 52 -> TIME (hours) Figure 9 : Time-temperature profile of furnace used for carbothermic reduction of tableted oxide-graphite powder mixture in batches of ~>1 kg for synthesis of (U,Pu)C and (U,Pu)N IT- 13.27. 3O.um L )N with (U Pu )C With ) ? 0.3 0.7 small M0_ phase white M->C_ phase Single phase Singie phase (U0.45Pu0.55)N (U0.8Pu0.2)N Figure 10 : Microstructure of mixed uranium phitonium monocarbide and" mononitride fuel pellets prepared in Radiometallurgy Division, BARC by "powdei—pellet" route IT - 13.28 Kigui-e 1 Mi r-..si r:,. t I;I <• i-i >.• :i; .>>•••;, h.v. dciisily (^.85% T.D.) MH-IKII .: i-i i i.:ii- ,ii;'.' i i ";!'• ii i i r 11 i c iiellcly prepared in 1 I I .V Figure 12 : Flowsheet for fabrication of oxide, carbide and nitride fuel pellets via the SGMP route IT - 13.30 THE IN-PILE BEHAVIOUR OF PLUTONIUM FUELS C.K. HATHEWS Head, Radiochemistry Programme Indira Gandhi Centre for Atomic Research Kalpakkam, Tarailnadu 603 102 1. INTRODUCTION Plutonium is the only man-made element that is produced in top quantities. Its main utilization is as energy source in nuclear reactor ^While plutonium has been used to fuel research reactors and is now proposed t-• be recycled if thermal reactors, its main application is foreseen in fast breeder reactors. Practically all the operating fast breeder reactors employ a mixed oxide of uranium and plutonium with the plutonium concentration ranqinq from 15 to 30%. The only exception is Fast Breeder Test Reactt• (FBTR) at Kalpakkam, India, which is fuelled by a mixed carbide having plutonium concentration of upto 70%. However, the carbide fuel bei;-, developed for power reactors have a plutonium content of less than 30%. 1U- mixed nitride or metallic alloy fuels which are under development also have plutonium concentration of 15-30%. These concentrations are much higher than that in the MOX fuel proposed under the plutonium recycling programme for thermal reactors- Further, the plutonium fuels in fast reactors are required to give much superior performance under more demanding conditions. Thus if we discuss the in-pile behaviour of fast reactor fuels we will be covering the best performance of plutonium-based fuels. This is the strategy adopted in this talk. 2. FUEL PROPERTIES & DESIGN PARAMETERS Table I summarises the properties of different fast reactor fuels. The mixed oxide is the preferred fuel today mainly because of the vast amount of experience available on not only its performance but also in fabrication and reprocessisng. Its strong points are its thermal stability and the lony temperature range in which it exists as a single phase solid. The other fuels are superior in terms of thermal conductivity and density (especially metal atom density), and are thus capable of higher linear rating and breeding gain. The design parameters of fuel pins based on these materials are shown in Table II. The fuel, made in the form of pellets (in the case of ceramic fuels) or slugs (metallic alloy), is filled into cladding tubes which are then sealed at both ends. The gap between the fuel and the clad is filled with helium or sodium to improve thermal conductivity. Thus we have two types of pin design - helium-bonded and sodium-bonded. The latter ensures better gap conductance, thus leading to lower fuel temperatures. The parameters listed in Table II represent the current designs which have been arrived after much development. All these fuels are capable of high burn-up and have reached 15-20 atom% burn-up in test irradiations. However, the number of pins irradiated to tftis ^evei is rather sma^i in tine case of dense fue^s ^carbide, nitride and alloy). IT - H.i The performance of a fuel 's assessed on the basis of the maximum burn-up to which it can be subjected at the rated linear power. Bunt-up is the energy released per unit mass and is expressed as either megawatt days per tonne (MWD/t) or atom percent fission (percentage of heavy element atoms that disappear on account of fission). Presently the target burn-up is about 20 atom percent fission. Several reports are availableon fuel behaviour in the form of books, conference proceedings -M:d ;c/it 'Sj i•1]. BASIC FACTAORS AFFECTING FUEL PERFORMANCE The two main factors Liiat affetl the pui forriiarice 01 a rue yju, a temperature and irradiation. High remp.-ji a1 ure.; inv/euse o; r'f'j-; i . Further a temperature gradient are developps, especially in the ra-ii^l direction, i^ich acts as a driving force for restructuring or the fuel a~1 •""el1"-trib-'t^on of is constituents. Some of these processes are be.ipf if ~-.:j. i ^ni'ie others are detrimental to high performance. Irradiation ieadb i.u uit ; ist ion ni'j ui me uranium oiiii p.^.^. nuclides, each of which gives out two fission fragments. These energetic fragments cause lattice damage which finally &nds up as a dislocation network. The fission products accumulate in the fuel as burn-up proceed^. Ten percent fission results in the injection of 20 atom percent of fission products (fps) comprising about 30 different elements into the fuel matrix. Figure 1 taken from ref. 1 shows the yields of stable arid long-lived fission products in the form of a periodic table. Out of the total 200%, 25.3% are contributed by Xe and Kr, 22.2% by volatiles (Cs, Rb, I, Br), 43.9% by rare earths, 41.7% by platinum group metals and 41.8% by Mo and Zr. Most of these fps are not soluble in the fuel matrix and, therefore, tend to form precipitates. The solid precipitates occupy their rather fixed volume which causes a relatively- small amount of swelling. Xenon and Krypton form gas bubbles whose additional space requirement depends on their size, temperature and pressure. At low fuel temperatures all fission product atoms remain in solution in the fuel matrix where they are born atom by atom. However, as temperature increases their mobilities increase as indeed that of all intrinsic and .extrinsic lattice defects. Blank [5,6] has analysed this process, which eventually leads to fuel spelling, in terms of a reduced temperature sc^le based on the melting point (Tm) of the fuel ( 6 = T/Tm). The diffusion of individual fission gas atoms fall in three temperature regimes (see fig. 2). At low temperatures(6< 0.35) the diffusion coefficients are very low and hence there is no precipitation of fission products. At moderate temperatures (0.35 < 6 <0.5) the in-pile kinetics are slow. Fission gas bubbles are nucleated but grow slowly. Therefore, in this temperature range, the rate of swelling is low and there is negligible gas release. At higher temperatures (8>0.5) the diffusion coefficients are high and the in-pi 1 e kinetics are fast with a strong temperature dependence. Fission gas bubbles are nucleated and grow rapidly within the grains and on the grain boundaries. The grain boundary bubbles coalesce forming channels from which fission gases escape. Thus, in this temperature range we have high swelling of the fuel accompanied by high gas release. This qualitative picture gets modified' as burn-up proceeds. The pattern in fig. 2 is expected to change slowly with burr;-up on account of the increasing concentration of fission products. In the light, of the above discus-sit n, we can compare the in-pile operation temperature-., of (Jiffevent fuels These considerations are useful in understanding the constraints in fuel pin design. The designer aims at high burn-up and high linear power, consistent with other requirements such as specific power. Increase in linear rating increases the central temperature. In the mixed oxide fuel this is limited by central line melting. In the dense fuels swelling and the consequent fuel-clad mechanical interaction (FCMI) limits the rating and the achievable burn-up. The clad itself loses strength due to irradiation swelling and this may put the upper limit for its in-pile life and thus for fuel burn-up. While fuel design and specifications depend on the type of fuel sons;. common features can be compared (see Table II). The dense fuels operate at higher linear ratings (~S00 W/cm) compared to the oxide fuel (~450 W/cm\. Dense fuels use somewhat stronger clad compared to the oxide. The smea< density, or the density of the fuel averaged over the internal diameter o; the pin, is about 75-80% n all cases, but is distributed differently. In t/K sodium-bonded pins, the initial void is almost entirely in the sodium-fi}leu gap, but under irradiation it gets distributed as porosity in the fuel. In the helium-bonded fuel there is an "as-fabricated" porosity which helps in accommodating part of the swelling. In all cases FCMI starts after the fuel makes contact with the clad. Ihough this happens relatively early in the life of the metallic fuel, the fuel is sufficiently soft as to be restrained by the clad. At high burn-ups, therefore, FCMI is more severe in the case of dense ceramaic fuels. THE MIXED OXIDE FUEL The mi>T;d oxide fuel is typically taken to a burn-up of 12 at%, but fuels are being developed that can attain 20 at% burn-up. At the beginning of IT - 14.3 its life the fuel has a high central temperature when operated at the maximum linear rating of about 450 W/cm, but it restructures within the first few hours developing a central void. Even then the maximum fuel temperature is typically over 2000 C. There is almost complete release of fission gases except from a small unrestructured outer zone. At higher burn-ups it is the precipitation of solid fission products that enhances fuel swelling and FCMI. There is a vast body of literature on the in-pile behaviour of the mixed oxide fuel and the chemistry of fission products[2,3,7,8]. The main areas of concern, which have implications on achievable burn-up, are the internal corrosion of the clad at higher bufn-ups, fuel-coolant chemical interaction and the reactions/precipitation of solid fission products. Only these aspects will be touched upon in the limited space available here. Oxygen potential of the fuel dominates any discussion on its behaviour[9, 10]. Oxygen potential determines the diffusion coefficients of oxygen anions, actinide cations and fission product atoms and ions. (Thus it affects the thermal creep behaviour of the fuel) It also determines the chemical state of the fission products. Under the temperature gradient prevailing in the fuel, the 0/M ratio ranges from 1.96-1.97 at the centre to about 2.00 at the outer edge. While the mixed oxide has a wide homogeneity range so that the fuel remains a single phase throughout this temperature and compositional range, the oxygen potential varies widely (from -600 Kj/mole to -200 kJ/mole). 0/M increases with burn-up. On the basis of post-irradiation examination, fission products ii^ the mixed oxide fuel have been classified as follows: Volatile fp elements: Kr, Xe, Rb, Cs, Br, I, Se and Te. Fps forming metal l.ic phases: Mo, Tc, Ru, Rh, Rb, Pd, Ag, Cd, In, Sn ?, Sb. Fps forming oxide phases: Rb, Cs, Sr, Ba, Zr, Nb & Mo. Fp elements which dissolve in the fuel matrix: Sr, Zr, Nb & rare earths. Some elements are found in different phases depending on the prevailing conditions such as temperature and oxygen potential. For example Mo is found in the alloy phase in the inner region of high temperature and low oxygen potential, but is a constituent of the oxide phase in the outer regions. Post-irradiation measurements [11,12] have shown the presence of "white" metallic inclusions containing Mo, Tc, Ru, Rh and Pd and a "grey" ceramic phase containing mainly the oxides of barium, molybdenum, uranium, plutonium and zirconium. This oxide phase which can be described is [Ba^ Ssr Cs ] [U,PuLn,Zr,Mo]0, is a eutectic oxide having a lower melting point "than tne fuel. To understand their formation we require basic data on the binary and ternary oxides. A fair amount, of data on these systems have been generated in our laboratory [13-16]. The observation of intergranular attack on the clad inner surface in highly rates oxide pins, led to an intensive search for the agent responsible IT - 14.4 for it. It is generally believed that fission product Te is responsible for it. One would expect Te to be tied up with Cs as Cs~Te. However, Cs can also participate in other reactions leading to the formation of Cs4(U,Pu)?07, Cs»CrO,, etc., depending on the prevailing conditions such as oxygen potential ana the availability of the reactants. This can raise the Te partial pressure sufficiently to attack the dad[17, 18]. In our laboratory we have carried out a detailed study of the relevant metal-tellurium systems to arrive at the threshold Te-potential needed for such attack[19, 20]. These detailed studies combined with the modelling of the buffering reactions in the fuel-clad interface leads to the conclusion that the equilibrium Te-potential would be too low for clad attack. It must, however, be noted that radiation effects of fission fragments can significantly raise the Te potential. Such effects have been shown to be significant in the case of iodine [21]. Another point of concern is the behaviour of failed fuel pins. In the event of a breach in the cladding, liquid sodium can come into contact with the fuel resulting in the formation of sodium uranoplutonate (NaJIO.). Recent studies have shown that this can form at relatively low oxygen potentials[22]. Thus the fuel itself can be the source of oxygen for the reaction to go forward. Since sodium uranoplutonate has a much lower density than the oxide, the reaction can result in considerable swelling in th^ region of the failure. CARBIDES & NITRIDES (MX type fuel) The major factors that dominate the behaviour of MX-type fuels are the chemical state of the fission products, swelling and gas release and compatibility with the clad. These fuels are denser than the oxide and have cubic close packed structures. Hence they should have greater difficulty in accommodating fission products. Therefore, burn-up effects can be expected to be more significant. Most of the non-volatile fission products form carbides and these include monocarbides, sesquicarbides and dicarbides. Some of these are partially soluble in the fuel matrix while Zr is fully soluble. Table III lists the likely chemical state of fission products in the monocarbide fuel[23, 24]. One notices that more carbon is tied down by fps than is released by the fissioned fuel atoms, and hence the C/M ratio of the fuel decreases with increasing burn-up. Therefore, the carbide fuel is usually prrepared with 5-15% of the sesquicarbide phase in order to avoid metal formation as the irradiation proceeds. The likely chemical states of fission products in the mononitride fuel are given in Table IV. It is seen that fps do not tie down all the nitrogen liberated in fuel burn-up. One would thus expect nitrogen release at high burn-ups. There are several complex carbides known including those that tie down U and Pu, but similar complex nitrides are not known. Restructuring in MX fuels is somewhat slower and less dramatic than in the oxide. For example a central void does not normally develop. This is because of the lower reduced temperatures as well as the absence of the evaporation-condensataion process which is governed by the partial pressures of the vapour species. Restructuring in these fuels is schematically represented in fig 4 and the transition temperatures are indicated in Table V. The coolest zone near the clad, which retains the grain structure of the "as- fabricated" fuel is designated zone IV. In zone III there is equi-axed grain IT - U.5 growth, with the grain boundaries decorated with fission gas bubbles. Zone II is one of low porosity and high density, where the grains and pores ara large Zone I is very porous, the pores being of irregular shapes and equal in size to grains. Central voids can form in nitride fuels by sintering. Zones I, II and III extend towards the clad with increasing burn-up thus causing zone IV to shrink. Swelling, defined as the effective diameter incr ase of the fuel, arises from 4 contributions [5,25]: 1. accommodation of solid fission products. 2. intragranular and intergranular bubbles. 3. coarse fission gas porosity, and 4. effects of pellet fracturing. Solid fps constitute 75% of all fp atoms and ccuse a swelling of about 0.5% per at% bu. This value increases to about 1% if Cs is included. When temperatures are sufficiently high, fission gas atoms precipitate to form bubbles. At low temperatures (<900 C) tiny intragranular bubbles (1-30 nm) are formed and the swelling rate is low. At higher temperatures larger (40- 300 nm) bubbles are formed both within the grains and at the grain boundary. This causes high swelling rates. Large pores develop in the high temperature zones which add to swelling. MX fuels have a strong tendency to crack, but part of the cracks heal especially at higher temperatures. This also contributes to swelling. Fission gas release is a related phenomenon which depends on temperature, porosity and burn-up. Thus it is easy to see why a higher rated fuel has a higher gas release as also its increase with decreasing smear density. Radial profiles indicate that most of the gases from higher temperature regions are released. Cooler sodium-bonded fuel releases less gas than helium-bonded fuel. Gas release also increases with burn-up. At 12 at% burn-=up even sodium-bonded fuels release about 30% of the fission gases. Gas release also depends on the chemical composition. Increasing carbon content decreases the rate of release. In-pile kinetics are slowed down as we go from carbide to carbonitride to nitride and this also affects gas release[1]. In test irradiations MX fuels have been taken to peak burn-ups in excess of 20%. In the light of early experience it was thought that only the sodium-bonded fuel concept would be successful. However, it has been shown more recently in the U.S and European programmes that helium-bonded pins can be designed and taken to high burn-ups [average 15 at%, peak 20 at %][26, 27j. The performance, however, is strongly influenced by design parameters as well as fuel and clad properties. It is necessary to have a strong clad that, retains adequate plasticity until the end of life [established SS 316, PE 16]. Pellet densities of 80-85% theoretical and smear densities of 75-80% are specified. M,,C3 cortent of 5-10% and an oxygen content of <1000 pm are maintained. fBTR is the only reracator which is fuelled by a mixed carbide fuel. The fuel used in this reactor has high plutonium content (Pu/(L)+Pu) =~0.7). Thus this fuel will operle a higher reduced temperatures and n>ince in-pile kinetics are expected to be fast. The restructuring effects are being studied in out-of-pile simulation experiments. In the fabrication of this fuel, oxygen levels could not be kept Inw. Hence detailed thermodynamic IT - 14.6 calculations were carried out to assess its effect on clad carburization[28]. This was followed by experimental measurements on the carbon potentials of the fuel and the clad and of the CO pressure above the fuel[29,30]. These studies provided adequate assurance that carburisation of the clad may not be significant at the modest burn-up aimed for this fuel. METALLIC FUEL It was recognised quite early in the development of fast breeder reactors that metallic fuels offered the maximum potential for breeding Thus all the early breeder reactors [DFR, E8R-I and EBR-II and Enrico Fermi] used metallic fuels. This fuel, however, lost out to the mixed oxide fuel mainly because of its poor burn-up capability. The original EBR-II fuel achieved a burn-up of only 1 at% by 1961. Even in 1975 the burn-up limit for Mark 1A fuel was put at 3%. Fuel element design was then revised to take advantage of the high rate of swelling[31]. The current designs provide a large fuel-clad gap (25% of the inner area of the rlad). Thus, by uie time the fuel swells to make contact with the clad, interconnected coarse porosity develops resulting in high gas release. The strong clad in the new design is able to restrain the fuel and the fuel creep is adequate to accommodate further swelling in the pores. The development of this fuel was based on U-Fs alloy (Fs is fissium, an alloy of Mo, Tc, Ru, Rd and Pci). However, a viable breeder fuel must contain 10 - 26% plutonium[32]. Since the !J-Pu alloy is unacceptable because of its low melting point and poor compatibility, about 10% Ir is added to the alloy. Fuel pins based on this ternary alloy with D9 and HT9 cladding have been successfully irradiated to burn-up values close to 20 at%. The current design of this fuel pin has a fuel slug diameter of 4.3? mm and a cladding o.d. of 5.84 mm (wall thickness = 0.381 mm). The smear density is 72-75%. The temperature gradient is only <150 C initially. After restructuring, this increases to >200 C thus raising the maximum temperature to>800 C. In the higher temperature zones this fuel would consist of the cubic t -phasj. As the temperature is lowered the equilibrium phases are t, and T and at still lower temperatures § + <^ . The radial temperature gradient thus results in the equilibrium phase regions as shown in fig.5 [32,33]. The addition of plutonium to the alloy fuel alters its in-pile performance. Increase in fission-gas diffusivity and plasticity of the fuel leads to increased swelling and coarser porosity. At high plutonium content (19% and above), anisotropy in swelling is less probably because of the reduction inec-phase. Redistribution of U and Ir becomes significant at high Pu concentrations. This radial migration, which seems to happen after the fuel-clad contact is established, results in an intermediate (Hd-radius) zone which is low in Ir [34]. The behaviour of fps fall into three categories[34]. Those fps which dissolve in sodium end up in the sodium bond above the fuel column and sodium- filled pores. Most metallic elements either dissolve in the fuel alloy or form precipitates. Lanthanides migrate to the periphery of the fuel pin where they precipitate as a separate phase in existing pores. Lanthanides also diffuse into the clad, generating a narrow embrittled layer. The lanthani.-ie diffusion produces .^ interaction layer which has been found to extend to 0.1 mm into' the clad at 10 at% b.u. This brittle IT - 14.7 layer containing nearly 20 \nt% lanthanides does not contribute to clad strength and is one of the major concerns in fuel-clad chemical interaction (FCCI). The diffusion of cladding alloying elements (Fe and Ni) into the body of the fuel has also been noticed. This can "lower the fuel melting point. At steady state operation the U-Fe eutectic point of 715 C is not reahed. However, one must consider the possibility of such temperatures being reached under off-normal conditions and the consequent penetration of the clad. Among the advantages claimed for metallic fuel are its safety features, especially its behaviour under transient overpower and superior neutronic properties. In the integral Breeder Reator Concept, the reprocessing is integrated with the reactor. An electrorefining method is being developed for this purpose, but it is yet to De demonstrated[35, 36]. Although the potential of the metallic fuel is established by recent studies at ANL, it is clear that more work is needed to understand its in-pile behaviour. CONCLUSION Among the four types of fuels discussed in this paper, the mixed oxide is the best developed. While this fuel may be adequate for the near future, it cannot fully exploint the advantages of the breeder reactor technology. There is enough information available to design a carbide fuel, but for high burn up application much development work is called for. Reprocessing experience is also lacking. There is less data on the nitride fuel, but it appears to be free of some of the disadvantages of the carbide fuel. For example, it dissolves readily in nitric acid and hence is amenable to equeous reprocessing methods. The nitride fuel is being pursued in Europe. The metallic fuel is actively under development in USA. Its advantages are high breeding gain, passive safety features and short fuel cycle through pyrochemical reprocessing. These features have to be demonstrated in a full fuel cycle before metallic fuels are fully accepted. REFERENCES 1. Hj. Matzke, Science of Advanced LMFBR Fuels, North Holland, Amsterdam, 1986. 2. Advanced LMFBR Fuels, Topical Meeting Proceedings, Tucson, October 1977, American Nuclear Society (1977). 3. International Conference on Fast Breeder Reactor Fuel Performance, Topical Meeting Proceedings, Monterey, Cali- fornia, March 1979, American Nuclear Society (1979). 4. Status of Liquid Metal Cooled Fast Breeder Reactors, Technical Report Series no. 246, IAEA, Vienna, 1985. 5. H. Blank, J. Nucl. Mater. 5*3, 123 (1975). 6. H. Blank, J. Less.Common Metals, 121, 583 (1986). 7. Proc. Conf. on Fast Reactor Fuel Element Technology, New Orleans, USA, 1971. -IT - 14.8 8. Behaviour and Chemical State of irradiated ceramic fuels, IAEA panel report, IAEA, Vienna, 1974. 9. C.E. Johnson, I. Johnson, P.E. Blackburn and C.R. Crouthomel, Reactor Technol. J_5, 303 (1972-73). 10. J.H. Gittius, J.R. Mathews and P.E. Potter, J. Nucl. Mater. 166, 132 (1989). 11. H. Kleykamp, J.O. Paschoal, R. Pejsa and F. Thummler, J. Nucl. Mater. J30, 426 (1985). 12. H. Kleykamp, J. Nucl. Mater. t3J_, 221 (1986). 1.3 K. Nagarajan, R. Saha and C.K. Mathews, Thermochimica Acta, 90, 297 (1985) 14. R. Saha, R. Babu, K. Nagarajan and C.K. Mathews, Thermo- chimica Acta, T20, 29 (1987) 15. R. Saha, R. Babu, K. Nagarajan and C.K. Mathews, Themochimica Acta, J_31_, 183 (1988) 16. R. Saha, R. Babu, K. Nagarajan and C.K. Mathews, J. Nucl. Mater. Jj>7, 271 (1989) 17. M.G. Adamson, E.A. Aitken and T.B. Lindemer, J. Nucl. Mater. 130, 375 (1985) 18. R.G.J. Ball, W.G. Burns, J. Henshaw, M.A. Mignanelli and P.E. Potter, J. Nucl. Mater. J6^, 191 (1989) 19. B. Saha, R. Viswanathan, M. Sai Baba and C.K. Mathews, High Temp - High Press. 20, 47 (1988) 20. R. Viswanathan, M. Sai Baba, D.D. Albert Raj, R. Bala- subramanian, B. Saha and C.K. Mathews, J. Nucl. Mater. 167, 94 (1989). 21. P.E. Potter, Pure and Applied Chem. 60, 323 (1988) 22. M.A. Migna nelli and P.E. Potter, J. Nucl. Mater. J25.> 182 (1984) 23. H. Kleyk:mp, International Conference on Fast Breeder Reactor Fuel Performance, Topical Meeting Proceedings, Monterey, California, March 1979, American Nuclear Society (1979) 24. P.E. Poster, Proc. Ini.ernationa1 Sympsium on Thermochemistry aid Ch'Brnicil Processing, November, 1939, Indira Gandhi Cnetre fo." Atomic Research, KalpakkaM (In Press). 25. M. Colio, M. Coquerelle, IJ..F. Ray, C. Ro-.chi, C.T. Walker and H. Blank, Nucl. Technol. 63^ 44-: (19.'J3) IT - 14.9 26. G.R. Harry, US Report LA-UR-83-1248 (1983) 27. K.R. Kummerer, J. Nucl. Mate,. J24, 147 (1984) 28. M. Sai Baba, S. Vana Varamban and C.K. Mathews, J. Nucl Mater 144, 56 (1987) 29. P.K. Prakashan, K, Ananthasivan, I. Kaliappan, S. Anthonysamy, P.R. Vasudeva Rao and C.K. Mathews, Int. Symposium on Actinides, Tashkent, USSR (1989) 30. D.D. Sood, Private Communication (1988) 31. L.C. Walters, B.R. Seidel and J.H. Kittel, Nucl. Technol. 65, 179 (1984) 32. H.V. l»-vitt, J. Nucl. Mater. 165, 1 (1989) 33. D.L. Porter, C.E. Lam and R.G. Pahl, Met. Trans. 21A, 1871 (1990) 34. G.L. Hofman, R.G. Pahl, C.E. Lahm and D.L. Porter, Met. Trans. 21A, 517 (1990) 35. L. Burn's and L.C. Walters, Trans. Am. Nucl. Soc, 49, 90 (1985) 36. L. Burris, G.E. Miller, E.C. Gay, J.P. Ackerman, Z. Tomczuk, J.E. Herzog and W.J.Kann,Trans.Am.Nucl. Soc. 56, 68 (1988) IT - U.1O Table 1: Properties of FBR Fuels Mixed oxide MC MN U+15Pu+10Zr Density (g/cm3) 11.0 12.95 .13.53 14.13 Melting tempe- rature (K) 2950 2700 3050 1426 Thermal Conductivity 0.022 0.17-0.22 0.17-0.22 0.18-0.25 (W cm"1 K"1) (1100°C) (800-1200°C) (80D-1200°C) (400-300°C) TABLE 2:PIN DESIGN PARAMETERS OF DENSE FUELS Fuel L'Q^-ISPuO^ Na-(U+10Pu+15Zr) Na-MXc He-MXC Outer clad 6.0 5.22 8.70 8.70 diameter D (mm) Clad thickness d(mm) ' 0.4 0.5 0.50 0.50 Fabrication porosity (%) 5-10 0 5 15 Smear density(%) 73-80 75 80 80 Linear^ating 40-45 33 <80 <80 (kW m"1) T (K) - 750 750 Fuel operational Hot Hot cool Hot and concept cool b Ref. 8, numbers in weight per ceri: c Conditions at maximum rating IT - 14.11 Table 3: The likely chemical state of the fission products in an irradiated uranium-plutonium monocarbide fuel Fission product element Chemical state Kr, Xe • Elemental Cs, Rb Cs.|_xRbx, (Cs1_x Rb Te (Cs.j_ x)o Ba, Sr 8a. Sr C, T, Ln Dissolved in monocarbide or sesq'jicarbide lattices Zr, Nb Dissolved in monocarbide lattice Mo, Tc (UPu)Mo1 Te C- Some Mo dissolved in monocarbide lattice Ru, Rh, Pd (UPu)2 [RUi_xRhx]C2 (UPu) (Ru, Rh, Pd), C Alloys with Ru, Rh + Pd < 1 U + Pu Pd Te Se, Te U2TeC2 Br, I, Sn (Br1-yV Ag, Cd9 In, Sb Alloy phases IT - 14.12 Table 4: The likely chemical state of the fission products in an irradiated uranium-plutonium mononitride fuel Fission product element Chemical state Kr, Xe Elemental Cs, Rb Cs1-xRbx' ^Br1-vIv^ (Cs. Rb )o Te Ba, Sr (Ba1-xSrx^3N2 T, Ln Dissolved in mononitride matrix. Zr, Nb Dissolved in mononitride matrix. Mo, Tc Elemental Ru, Rh, Pd (UPu) (Ru, Rh, Pd)3 Se, Te . CSi.xRbx)2 Te Br, I, (Cs1_xRbx) Ag, Cd, In, Sn, Sb Alloy phases Table 5: Temperature ranges of structural zone's in MC1 N fuels [4X burn-up] '"x x Fuel zone IV zone III zone II zone I central hole T(°C) Trange(°C) T(°C) • T(°C) carbide-like ^900 30-80 >1000 >1300 (0^0.3) carbonitrides <1100 200-300 >1-350 >1400 (0.3 IT - n.13 la na IllaflVa fva [via fvilaf VIII [ib [lib nib IV b Vb VIb Vllb 0 H fission product He J region considered Li Be c 8 C N 0 F Ne dataon solubility in carbide 0 available Na Mg data on solubility i n nitride Al Si P S Cl Ar available. • 0 • o • o K Ca Sc Ti V Cr Mn Fe Co Ni Cu Zn Ga Ge As i>e Br Kr 04 0 2 2.5 • o • o • o 0 O e Rb Sr Y Zr Nb Mo Tc Ru *Rh *Pd Cd In Sh Sb Te 1 Xe 1.2 3-9 2.0 19/. G L 22 A 5.5 I'M.9 5.A 130 '1108 0.1 0-B 30 1.6 22.8 • o • 0 • 0 • 0 • « • • Cs 8a ta Ht Ta W Re rV Pf Au Hg 71 Pb 8i Po Af Rn 19.2 6.2 5-6 Fr Ra Ac Lanthanides Ce Pr Nd Pm Sm Eu Gd Tb P/ Ho Ef 1m Yb Lu 12-8 C6 13.8 1.8 32 0.5 o.c 0.1 Aclinides Th Pa U Np Pu Am Cm Bk C f ES Md No Lw 1 !?• Rare Earths PLMetals Zr Rare Gases Volatile Mo (CS,Rb,l.Br) A3.9 22.A 19 .U 2 5.3 22.2 Ficj. 1 Yields of stable or longlived fission products in the fission of Pu-239. IT - 14.14 6 = Tm 0.6 0.5 0-4 0.35 -10 10 .12 10 (A C>J C -14 o 10 en Q .16 10 -18 10 -1.4 2.0 2.5 3.0 _L Im 8 "T Fig.2 Temperature dependence of the fission IT - U.15 0 T(K) u •i (K) T (K) MN 1-1124 + 2080 2440 124-00 T T = 1405 K ! Tm=2700 K K i Tm=2950K t. 0.70 f 983.5 t 1890 [-2135 2100 N a o • 0.80 t 843 a. -1620 1800 0.50 i ?3 -150 0 •Irradjction growth'j j no vacancies for < gas bubbles j available 0.40} 562 1080 -1200 i 0-30 421.5 |_780_ _ 915 Fig _ 3 Comj^G risen of the in. pik. operQljon__temperatures_ ofdiffereni viee text for dc ,:. ,.'> Low temperatures high temperatures edge centre ZONE IV III II STRUCTURE OF GRAIN GROWTH PSEUDO COLUMIN- VERY POROUS AS.FABRICATED GRAIN BOUNDA AR GRAIN ZONE, CENTRAL ZONE FUEL RY BUBBLES ELONGATED GRA- INS AND PORES FIG-4 SCHEMATIC PRESENTATION OF THEJ-OUR STRUCTURAL ZONES IN ADVANCED FUELS FIG-5 PHASE BOUNDARY LOCATIONS EXPE- CTED AT THE AXIAL MID POINT OF A U-Pu-Zr FUEL PIN AFTER APPROXI- MATELY 1 AT % BURNUP IT - U.17 CHEMICAL BEHAVIOR OF FISSION PRODUCT IODINE IN IRRADIATED OXIDE FUEL PINS FOR NUCLEAR REACTORS M|chj_o VAMAWAKI1 and Kenj i K0NASH|2 1.Faculty of Engineering, University of Tokyo, Hongo, Bunkyo-ku, Tokyo 113. Japan 2.Tokai Works, Power Reactor and Nuclear Fuel Development CorP-» Tokai-mura. Ibaraki-ken 319-11, Japan SUMMARY This review summarizes theoretical and experimental studies of Csl radiolysis in an operating fuel pin. From this review, it is concluded that the radiation effect is important for the understanding of chemical behavior of iodine in operating fuel pins. « (Key Words: chemical behavior^ fissioa product,, iodine* cesium iodide, radiolysis) 1. INTRODUCTION The volatile fission product iodine accumulated in nuclear fuel pins is considered as a responsible chemical agent for the fuel cladding chemical interaction (FCCI) in FBR [1] as well as for the stress corrosion cracking (SCO i(\ LUR [2]. The release of io IT - 15.1 In above phenomena, chemical form of iodine is important. According to the analysis based on thermodynamic data £ll, iodine tends to react with fission product cesium of which fission yield is about ten times greater than that of iodine, resulting in the formation of a stable compound, cesium iodide, and the partial pressure of iodine is kept very low in operating fuel pins. The decomposition of the molecule occurs when the molecule is excited above the threshold energy of decomposition. When the system is in the state of thermodynamical equilibrium, the molecules have a liaxwelI-Boltzmann distribution as to their kinetic energies corresponding to the temperature of the system. The average kinetic energy of molecules is about O.leV at 1000K. In the case of Csl, the threshold energy of decomposition is 4.35eV [5], which is far in excess of the thermal energy. Therefore, there is a quite small possibility of the decomposition of Csl and this is the reason why Csl is taken as highly stable in thermodyna/nical analyses. On the other hand, the energy of the radiations in the operating fuel pins is expected to be much larger than the threshold energy. The radiations having high energies can activate Csl to an excited state in excess of the threshold. Consequently, the radiations can contribute to the decomposition of Csl ana io the increase of the partial pressure of iodine. The radiations in operating fuel pins are divided into three groups, i.e. charged particles (fission fragments, a particles, .lectrons), uncharged particles(neutrons) and photons( r -rays). The charged particles play the main role in the radiolysis of gaseous Csl. The radio lysis by r-rays js important for solid Csl. Theoretical studies have been done for the gas phase radiolysis. Experimental studies have been conducted for the gas phase radiolysis and solid IT - 15.2 Csi decomposetion[8,9]. Gas atoms in the fuel pins (He, Xe and Kr) are excited by fission fragment impact into higher excited state (He*, Xe* and Kr*), or ionic state (He+*, Xe+* and Kr+*), and then Csl is decomposed by the inelastic collision with the excited gas atomsI He(Xe, kr) + fission fragments — He*(Xe*,Kr*), (2) He*(Xe*,Kr*) + Csl -* He(Xe, Kr) + Cs + I. (3) The other mode is also an indirect effect of fission fragments on the Csl decompositionClO]. The kinetic energy transfer from fission fragment to gas atom is considered instead of the internal excitation . The fission fragments generate energetic atoms in the gas region in operating fuel pins. In this case, the principal contributor to the flux of energetic gas atoms in excess of the threshold energy of Csl decomposition, 4.35 eV, is not the flux of the Maxwellian distribution but that of the collision cascade atoms initiated by the radiation of fission fragments.The Csl decomposition in operating fuel pins is induced by the collision with the cascade atoms initiated by fission fragments." He(Xe,Kr) + fission fragments -* cascade atoms, Csl + cascade atoms -* Cs + I. (5) Considering above radiolysis processes, the iodine partial pressures in the operating fuel pins have been estimated. In Table 2, the effects of the three modes on iodine partial pressure are compared. For comparison, following values, were taken from table 4 in ref.[9]I temperature of gas phase is 650K. oxygen potential is -4O0kJ/mo!, partial pressures of Csl and Cs are 1.8X10"^atm and 4.5 X10'14 atm, respectively. The temperature difference across fuel-cladding gap was not considered. The fission density of 4X10'^(fission'em^sec) was assumed IT - 15.A in the calculations with the mode 1 as well as with the mode 3. The radiation dose rate of 21.6 W/g was used in the calculation of the mode 2. These radiation conditions correspond to that in the fuel-cladding gap of a typical LWR fuel rod with a linear heat rate of 300W/cm. Calculation of each mode has resulted in considerably similar iodine pressures to each other, which are much larger than that for thermal equilibrium state. Hence, it can be summed up as to the radiation effects that the three modes of radiolyses occur simultaneously in operating fuel pins. 3.EXPERIMENTAL STUDIES Experimental works have been done to prove high iodine pressure in operating fuel pins. In regard to the radiolysis concept related with the SCC of zircaloy, Cubicciotti and Oavies [11] first reported the rTadiolysis experiment of solid Csl. They confirmed the release of iodine from solid Csl irradiated with Y ray of ^^Co (IMeV) at room temperature. However, the release rate was markedly decreased at cladding inner surface temperature 570 K. When Cs metal was added into the sealed capsule containing Csl, the release of iodine was not observed even at room temperature. Shann and Olander [12] made an experimental study by using a proton beam from the Van de Graaff accelerator to examine the propriety of the Cubicciotti et al.'s concept. They assembled a special SCC test apparatus (Fig.2), in which the outer surface of an internally pressurized zircaloy tube was exposed to a Csl molecular beam and at the same time it was bombarded with a proton beam on the same spot. A 10u A, 175keV proton beam was generated by a Van de Graaff accelerator which was connected by the flange shown on the left hand side of IT - 15.5 Fig.2. As a result, no SCC failures attributable to the radiolysis of a deposited Csi crystal fayer were observed to occur on the outer surface of the zircaloy tube. Oavies et al.[13,14] also tried to prove the Cubicciotti et al.'s concept with an in-pi le irradiation test using real fuel pins(Fsg-3)- As therniodynaniic narkers, Cu and Pt foils were placed in the fuel pins and formation of iodides was checked in order to estimate the iodine partial pressure in fuel pins during irradiation. After irradiation, not only iodine but also cesium were identified on both of the markers, hence the evaluation of iodine pressure was impossible. The results have been generally taken as a negative proof to the Cubicciotti et al.'s concept. However, it should be noted that a stainless steel barrier tube was placed between the fuel and the cladding in order to protect the thermodynamic markers from direct exposure to fission fragments. The barrier tube may eliminate the effect of fission fragments on iodine partial pressure. Recently, Cox et al.[15] and Bibilashvili et al.[16] reported that Csl in the r -field caused the SCC of zircaioy cladding. Above experiments were done to study the r-radiolysis of solid Csl. On the other hand, the radiation effect of gas phase on the iodine pressure, which is described in section 2, was examined experimental l'y by Yamawaki et al.[17-19]. The experiments were designed to evaluate the effect of fission fragments on iodine partial pressure in a gaseous Cs-I system. Electrons, which have comparable velocities with those of fission fragments, were used in order to simulate a radiation in an operating fuel pin. The setup of the reaction chamber used in the experiment is shown schematically in Fig.4. Cesium iodide vapor was irradiated by the electron beam. Atomic absorption technique was used to determine iodine partial pressure liberated by radiolysis of Csl. The increase of IT - 15.6 Cs pressure by Csl radiolysis was observed with increase of electron current. The predicted crucial role of the radiation to iodine was experimentally confirmed. The cross-section of Csl radiolysis, 2.8X lO'^cm^, was determined. 4.C0NCLUSI0N From this review, it is concluded that the radiation effect is important for the understanding of chemical behavior of iodine in operating nuclear fuel pins. The discussions ,however, to determine the iodine partial pressure are serai- quantitative because of limitation of our knowledge on the chemistry in intense radiation field. Further experiments under the condition close to the actual radiation conditions are desired. It is necessary to develop more precise calculation method to simulate radiolysis processes in an operating fuel pin. 5.REFERENCES [1] M.Aubert, D.Calais and R.LE Beuze, J. Nucl. Mater. 58(1975)257. [2] B.Cox and J.CWood,"Corrosion Problems in Energy Conversion and Generation", Ed. C.S.Tedmon (Proc. Electrochemical Soc.,1974.) 275. [3] J.Paquette et a!., J. Nucl. Mater. 130(1985)129. [4] O.Goetzmann, J. Nucl. Mater. 84(1979)39. [5] F.P.TulIy, Y.T.Lee and R.S.Berry, Chem. Phys. Letters 9(1971)80. [6] K.Konashi, T.Yato and H.Kaneko, J. Nucl. Mater. 116,1(1983)86. [7] K.Konashi, K.Kamimura and Y. Yokouchi, J. Nucl. Mater. 125,2(1984)244. [8] W.G.Burns, E.H.P.Cordfunke, P.A.V.Johnson, P.E.Potter, G.Pruis and H.H.Rand, Proc. IAEA Tech. Commit. Meet, on "Fuel Rod Internal Chemistry and Fission Products Behavior", Karlsruhe, (1985) 35. [9] R.G.J.Ball, W.G.Burns, J.Henshaw, M.A-MignaneiIi, and P.E,Potter, J. Nucl. Mater. 167(1989)191. [10] K.Konashi, M.Yamawaki and T.Yoneoka, J. Nucl. (later. 160(1988)75. [11] D.Cubicciotti and J.H.Davies, Nucl. Eng. 60 (1976) 314. [12] S.H.Shann and D.R.OIander, J. Nucl. Mater., 113 (1983) 234. IT - 15.7 [13] J.H.Oavies. F.T.Frydenbo and tt.G.Adamson, J. Nucl. Hater. 80(1979)366. [14] J.H.Davies, F.T.Frydenbo and M.G.Adamson, Proc. Enlarged Halden Program Meeting, Norway, 1977. [15] B.Cox, B.A.Surette and J.CWood, J. Nucl. Mater.. 138 (1986)86. [16] Vu.K.Bibilashvili, M.V.Viadinirova, I.S.Golovnin. I.A.Kulikov, V.V.Novikov, A.S.Sotnikov. Proc. IAEA Tech. Commit. Meet, on "Fuel Rod Internal Chemistry and Fission Products Behavior", Karlsruhe, (1985) 173. [17] M.Yaaawaki, M.Hirai, T.Yoneoka, K.Konashi and M.Kanno. J.Nucl. Sci. Technol. 20 (1983) 852. [18] M.Kanno, M.Vainawaki and T.Voneoka, J. Nucl. Mater. 130 (1985) 395. [19] M.Yamawcki, T.Yoneoka, H.Kaneko and K.Konashi, J. Nucl. Hater. 154(1988)47. Table 1 Typical LVR conditions Radius of pel let ; 0.5 cm Oxygen potential of fuel ; -lookcal/mol Linear heat rate ; 300 W/cn Temperature difference across fuel-cladding gap ; 200 V Total gas pressure in fuel rod ; 10 at K Table 2 Comparison of radiation effects mode of Col!ision particles Iodine partial radiolysis with Csl pressure(atm) non-rad. thermalized gas atoms . 4.3X10-18 [9] mode 1 fission fragments 6.1X10-8 mode 2 excited gas atoms 8.5X10-8 [9] mode 3 cascade atoms 6.6X10"7 IT -15.8 TOO 6OO_5OO -5 • - O -10 £ a. CL s S -15 - - -10 - -16 12 14 16 10-VTC1/K) Fig.1 Radiation effect on the partial pressure of iodine as a function of reciprocal temperature of zircaloy cladding (ref.[7]). I20OSER END CAP TUBE SPECIMEN -THERMOCOUPLE CURRENT FEEDTHROUGH FILLING—: VALVE PRESSURE TRANSDUCER I \ TO DIFFUSION PUMP Fig.2 Apparatus for stress corrosion cracking test with proton bombardment (ref.[J2]). A MARKER FOILS LOCATED IN PLENUM - AU.anoo<'»oota~ or ANDO.I~WIOI LAtVRINT JfylCI O BTAINLEU1THU O 4 MAHKEM 'OILS LOCATfOIW":f« AXIAL OKOOVIi O I PERPORATEO fTAINLIX ITIIL O •ARRIIRTUM ZlflCALOV CVAOOINO HAPNIA PELLET (FLUX OIPREWORI ROM « ANO O MOOS A ANO C HOLE! IN lARKlCr) ANNULAR FlUL TO COINCIOE WITH IVlTH OIRiCT THI CIRCUMFERENTIAL GROOVES ON rUBi. "> ACCESS "tOMCCXi WHIN AS5EM«LEO TO PLENUM Fig.3 Schematic diagram of fuel rods for thermodynamic markers tests (ref.[I4]). Pi stem .Grid .Optical / / window Fig.4 Schematic of reaction chamber for ?as phase radiolysis test (ref.[I9]). IT - 15.10 ^ ACTINIDE EXTRACTION CHEMISTRY WITH THE AMIDE TYPE OF EXTRACTANTS C. Musikas. C. Condamines, C. Cuillerdier and L. Nigond CEA-DCC-DPR-SEMP-SECP CEN-FAR - BP n° 6 - 92265 Fontenay-aux-Roses (France) SUMMARY : The extraction chemistry of the actinide ions and the inorganic acid: with amides, N,N-dialkylamides (RRNCOR') and N,N'-tetraalkyl 2-alkyl propane diamides (RR'NCO)2 CHR") is reviewed. Examples of application of those exiractants in the separation of the nuclear industry are given. The monoamides are advantageous alternative to tributylphosphate and the diamides usefull for the alpha TRU wastes decontamination. The most interesting aspect of these extractants are the innocuity of their radiolytic products and the possibility to incinerate the used solvents without large amonts of inorganic residues as for the organophosphorus extractants. Actinides, Monoamide, Diamides, Extraction, Inorganic acids. Introduction The N.N-dialkylamides (RRNCOR') and the N.N'-tetraalkyldiamides (RR'NCOCHR"CONRR') have been investigated since several years in our labora- tory as alternative extractants to the organophosphorus molecules for the separation of the actinides [1 to 3]. In this paper we will give a up-to-date review of the proper- ties of the amide extractants. The possibilities oi industrial applications will be considered. The monofunctionnal N.N-dialkylamides are good extractants for the (IV) and (VI) actinide ions [4 to 6] and have been proposed to replace the tributylphi sphate in the PUREX process. The comparison of processes using TBP or amides is given Table I. Many advantages are expected by replacing TBP, but plant demonstrations are not yet available and still large amounts of data are needed to design those plants. The N.N'-tetraalkyldiamides [1-2-3] extract fairly well the tri, tetra and hexavalent actinides ions from acidic solutions. Their field of application is the decontamination of the transuranium wastes (TRU W), storable in deep ground repositories. After the alpha activity removal to less than 3.7 x 103 Bq per gram the wastes can be stored in sub-surface disposals. The advantages of the tetraalkyldiamides over the polyfunctionnal organophosphoms extractants resemble to those quoted in Table I for the monoamides. n - 16.1 Table I: Advantages and drawbacks of N,N-dialkylamides over TBP for the PUREX process ; according to the litterature until 1980 Point N,N~dialkylamides TBP Availability Easy to prepare and Co purify but hot Commerciai commercial Extraction U(VI) and Pu(IV) quantitatively Slronger extractant extracted than amities Dcsextraction U(VI) more easily desextracted Selectivity Branched C - O substitucnts increase Lower selectivity selectivity : U(Vf) - Pu(IV); U(VI) Zr(IV) Radiolytic products Carboxylic acids, not cumbersome Dibuli'iyi,phosphoric acide cumbersome Wastes from the IAJ*/ amounts Large amounts cxtractant Diluents Aromatic hydrocarbons aliphatic hydrocarbons II. Extraction chemistry II. 1 N N-dialkviamides The distribution ratios of U(VI) and Pu(IV) between two N,N-dialkylamides with a linear carbonyl substituent, DOBA (C,H7CON(CH,CHCJH5C4H9),) or a branched carbonyl substituent DOiBA ([CH3],CHCO[CH,CfiC2H5C4H9]2) and nitric acid solutions are shown in figure 1. The branched substituent causes decreases in the distribution ratios, higher for Pu(IV) than for U(V1). This feature can be used for the U(VI) - Pu(IV) partition. DOBA DO.BA *! Cat) (mol I-1) HNO3 Figure 1 : Distribution ratios of Lf(IV) and Pu(IV) between IM DOBA or DOiBA into TPH (branched dodecane) and aqueous HNOj IT - 16.2 From saturation experiments and spectrophotometry of the organic phases it has been deduced that U(VI) is extracted as UO-,(NO3),(Amide), from slightly acidic and neutral nitrate solutions. Crystallographk structures of" several [7, 8] dinitrato diamido uranyle compounds showed tha* the two amides are in the first coordination sphere of UO,2+ and are linked to U through the C = O oxygens. The N of the amide moiety is father far from the metal. From acidic nitrate solutions U(VI) is extracted as LJO2(NO3)3 H Amide, the UV spectra of the organic phases were token as evidence for the presence of the UO2(NO3)3' species for which the absorption spectrum is well known (9). Branched or linear amides show the same behavior. Pu(IV) is extracted from slightly acidic solutions as Pu(NO3)4(DOBA), or Pu(NO3)4(DOiBA)3. Steric hindrance can explain the differences between the linear or branched substituted amides. As for U(VI), the acidic species Pu(NO3)6(DOiBA)H2.(Amide)x has been detected by spectrophornetry of the organic solutions obtained by extraction of Pu(IV) from highly acidic nitrate solutions. The ratios of the neutral to the acidic species concentrations for U(VI) and Pu(IV) as a function of the aqueous acidity are shown figure 2. x 0) c s o o c o •H c !l mol.l" Figure 2 : % of UO,(NO3)3 C Figure 3 : Schematic representation and Pu(NO3)6 H2 (DOBA)y in TPH of the [UO2(NO3)2(Amide]2.Amide as a function of ihe aqueous acidity outer-sphere complexe For all the amide cxtractants investigated by us the slopes of log DMcta] as a function of log CAmj,je have higher values than those expected from the stochiometry of the organic complexes (see Table II). It can be seen that the higher deviations are observed for the amides which have a rather low steric hindrance close to the amide nitrogen. Such behavior suggests that the high slopes are due to the interaction of the free amide with the bonded amide in the second coordination sphere of the uranyle complexe. Such dipole-dipole interactions are schematized figure 3. The higher positive charge of the nitrogen of the bonded amide is responsible for this preferential interaction compared with the free amide-free 2 amide interaction. The splitting of the \JO2 * highly loaded amides-aliphatic hydrocarbon mixtures can be understood as the consequence of UO^NO^MAmide);, second sphere interaction with the free amide. Table I'l contain the solubility of UO,(NO3):(Amide)2 into alcanes for various amides. The presence of a C2H5 substitue'nt on the B C of the N substituent is efficient tor the solubility of the amide nitrato uranyle complexes into aliphatic hydrocarbons. IT - 16.3 jl: Slopes of the lines log DMetal vs lod CAmidc for various mono and diamides Extractant Diluent Organic complexe Slope (C4Hy2NCOC1IHa tbutylbenzcnc UO2(NO3)2(Amidc)2 2.67 (QH^COC,^ decane 2.7 (C,HS-CH-[CH3J2)2NCOC!1H2, f-butylbenzene 2.15 (QK^CHfCjHjlCH^jNCOCjH!, dodecanc 235 (C1H,CHJNCO)2CHCI2H2S dodecanc La (NO,)3 (DIAM). 2.4 and La (NO,)3.DIAM (C4H,CH3NCO)2CHCl2Ha dodecane UO2(NO3)2DIAM 2.2 (C8H17CH3NCO)2CH2 t-butylbenaene Am(NO3), (DIAM)2 4 Table III: Solubility of UO,(NO3)2 in various N,N-dialkylamides 1M into a branched dodecane (e = 25°C) Amide (mol H) (mol F) (q,H9)2NCOCHH23 0.336 0 tt 0.29 4 (C2H5-CH-[CH3]2)2NCOCnH23 0.269 0 11 0.15 d (C4H9CH[C2H5]CH2)2NCOCUH23 > 0.5 0 11 > 0.5 4 The inorganic acids are extracted by N,N-dialkyIamides, but it is only for HNO3 that the organic adducts resemble those found rot TBP. From HNO3 distribution measurements and IR spectroscopy we found that the species HNO3(Amide)2, HNOyunide and (HNO3),Amide are present into the organic phase. Amide was either C3H7CON(CH2CHC2HJC4H9)2 (DOBA) or (CH3),CHCON(CH2CHC2HJC4H9)2 (DOiBA). The % of those species as a function of the aqueous HNO3 concentration is given figure 4. The amounts of extracted water are weak when compared with TBP. This low water-amide affinity can explain the poor extraction of H,SOJt H3PO4, HCIO4, HCI, all acids which necessitate water to be transferred into the organic phase [10]. IT - 16.4 Table IV : Influence of the substituents R, R\ R" of (R R' NCO), CHR" (DIAMIDES) upon the distribution ratios of Am(IH) between nitric acid solutions and 0.5M diamide into t-butylbenzei t or benzene R" R R' Diluent C HNO3 H QH!3 QH, 0.11 benzene 2 to 3 H QH, QH, 0.18 benzene 2to3 H QH, QH5 0.45 benzene 2 to Z H CH} QH3 0.55 benzene 2 to 3 H CH3 QH, 1.18 benzene 2 to 3 QH13 CH, QH, 1.16 t-butyl. 5 QH4OQH3 CH QH, 3.28 t-butyl. 5 C2H4OC4H13 CH3 QH, 7.55 t-butyl. 5 QH.OC^OC.H,, CH, QH, 9.43 t-butyl. 5 CJH^OCJH.OCHCCJH^CHJCCCH,), CH, QH, 10.4 t-butyl. 5 3+ * DA_ for R" = H series arc those at the maximum of the curve vs . At the maximum 2N < < 3N 1 [HN03] (mot I" ) Figure 4 : Extraction of HNO3 by 1M DOiBA into TPH. Distribution of organic species with respect to the aqueous acidity. Organic phase 1M DOiBA into TPH IT - 16.5 11.2 N.N'tetraalkvl 2-alkvl propane I'-dinmides The substituted diamides (RR'NCO)7CHR" have been selected according to their ability to extract the actinides (III) from HNO3 liquors. The influence of R.R" and R" substituents are shown Table IV It can be seen that a short R or R1 substituent is necessary to obtain high distribution ratios, probably because steric hindrance inhibit the metal-extractant interaction. The presence of a long chain R" substituent enhances the distribution ratios. The slope of the distribution curve for the unsubstituted diamide is typical of a HNO,-metaI competition for the diamide coordination sites [2]. This competition is less important for the substituted diamides and suggests iheir lower basicity. The basicities of diamides have been compared by measuring their half neutralization pH by HC1O4 (pHi/2) in an acetic acid-acetic anhydride medium. The values obtained by taking the pH1/2 of N- butylacetamide (HC4H9NCOCH3) as scale origin are contained Table V. The slopes of log DM as a function of iug CQ}.nu..^ are not integers and are higher than expected by the stochiomeuy •.k-ttrni'Mc-J ov ffic saturation, method checked bv IR and UV spectroscopies. The results are contained Table II. The lanthanicies form two kinds of organic complexes M(NO3)3.(DfAMIDE)2 and M(NO3)3DIAMIDE. The concentration ratios of the 1:2 to 1:1 complex decreases as the ionic radius decreases showing that the 1:2 complexe is a rather sterically hindered species. Table V : pH at half neutralization by HC1O4 into acetic anydride for various rrr lonnmides (pHi/2 for iF;.outy;amide was taken as the orgin of the scale) Manolamide acronyme Formula pHi/2 DBDMMA 3.43 DBDMBMA 3.1 DBMM3ONMA : 1VCH,NCO),CHC2I I. 2.72 Species such as UO2(NO3)3 H DIAMIDE ou Pu(NO3)6H,(DIAMIDE)? were not observed in the conditions were the monoamide acidic species were obtained. This feature shows that the formation of the acidic species does not depend only of the extractant basicity but also of the competition NO3'/extractant for the metal coordi- nation sites. In the case of diamide this last parameter is in favor of diamide. The inorganic acids are extracted by the diamides. HNO3 forms the adducts HNO3(DIAMIDEK HNO3.DIAMIDE, and (HNO3),DIAMIDE. HCIO., forms only HCIO4.DIAMIDE.-The IR investigations of the strong inorganic acid-DIAjMIDE adducts indicate the transfer of proton from the acid to the DIAMIDE [11, 12] because the C = O stretching band is shifted to lower energies by 140 cnr1. For the less basic morioamide a smaller shift of 10 to 40 cm-1 was observed and was attributed to more or less strong hydrogen bonds. Increase in the dielectric constant of the organic phase favours the protons transfer to diamide, HNO, and dimethyl- dioctylmalonamide are hydrogen bonded in toluene (£= 2.3 IT - 16.6 IH. Chemistry related to practical applications of the amide extractants III.1 N'N-dialkvlamides The main field of N,N-dialkylamides uie is the replacement of TBP. Two of the major drawbacks of TBP, the retention in the organic phase and the precipitation of metals due to radiolysis are not observed with N,N-dialkylamides. The degraded solvents show only small changes in the distribution ratio:, of the metallic ions. The possibility of using the monoamides in a first cycie of the nuclear fuels reprocessing has been tested in bench scale mixer-settler batteries. The flow sheet of one of the runs is given figure 5 and the results in Table VI. Table VI: Results of the first cycle PUREX process using N,N-diaIkylamides Flux U% of feed Pu Zr Ru Np Exti action - solvent > 99.99 99.999 0.C1 0.05 95 - aqueous < 0.0005 0.0008 99.99 99.95 5 Partition - solvent 99.89 0.006 not detected • aqueous 0.11 99.994 95 U desextraction - solvent not detected not detected not detected - aqueous 99.89 0.006 not detected FEEO. SOLVENT O(VI) S01VENI 00|BA IM F-.OV] 2^5, 00,8A IM C08A 0.SM CO8A 0.5H 132ml/h SOml/h 3l.2ml/h 1 SOIVEHT > Eilrocllon It JC- Partitio+ n 20 Bock -E SCRUB HNO IN O.OIH lS9mlfh Figure 5 : Flow sheet for a PUREX process first cycle using N,N-dialkylamides IT - 16.7 III.2 N'N-tetraalky! 2-alkyl propane 1.3-diamide The diamides can find applications fot the removal of the alpha emitters (actinides [III], [IV], [VI]) from acidic or neutral salted wastes. An example of such application is given figure 6 and the results for this bench scale test, Table Vil. Those results are compared with typical results obtained by using the TRUEX [13] process of which the solvent is a solution of TBP and ociylphenyl N,N- diisobutylcarbamoylmethylene phosphine oxydc (CMPO) into an aliphatic hydrocarbon. It can be seen that the efficiency of the diamides compare well with the CMPO. Table VII: Decontamination factors obtained in mixer-settler batteries for a waste treatment with DBDM 3-ONPDA compared with the TRUEX process ones* Element DF extraction DF stripping Am 1.26 104 - 5 1CF* 800 Eu 6.7 103 250 atot 1.1 1CH 125 Pu > 105 - > 105* 274 Np 144 6* 400 U(VI) > 1000 - 105* Fe > 1000 EXTRACTION STRIPPING l_ Solvtnf U-NpIV j a/A = i &/A 1 V * !.0«3t 12 12 v r i.06a i Raf/inaU h NHA 0.2 M j feed HNO, 2 H Aauious phase HNO, ff.l Ml HNOj 5 « U ml/h An. Pu_ 1i0 ml/h AcMnides Solicit 0.5 H in t^utylbtmtnc ltd nl/h STRIPPING U STRIPPING Np -1 HNO, !0 M 3> 0.5 H. Fi"ure 6 : Flow sheet for the counter-current separation of actinide using D3DM-ONDPA (QH.CHjNCO^CHCH.OQH,, IT - 16.8 Die distribution latios of the aetinides ions between niiric acid solutions and radiolytically degraded solvent are plotted figure 7 as a function of the gamma dose received by the samples. Unlike the TRUE A solvent there is no drastic change in the solvent extractive properties after radiolyticor hydrolytic degradation. .. ... -. r _-r r r f 20 ta 40 Doi« M rod 7 : f'ffcet of (lie M)f "o dose upo.'i (lie distribution ratios of II, I'n, Am betwee/i IM I-XAMIOi: (<',,! lv< U,N( o),( IK ,1!.,()( ,H4O( ,,H., into l-butylbcnzene irradiated and SN oi o.sN --- aqueous solution T'lic mono and diainide lype of cxlractants could be interesting alternative to the classical organophospliorus molecules for applications in the nuclear industry, mainly because tlicy are completely in The structure of ihe amide extraclanls can be adjusted for the separation )b!ciiis by the (nonet choice of the (.' •- O or N subslitnents which play an important role because of the n character of the amide bond. This IT character is responsible for the steric hindrance of ihe substituents leading to various sclectiviiics. Mlectronic effects due to the possibility of electrons delocali/ation were observed; for example the basicity of the (' O moiety is sensitive lo the subsiitution of ihe central CU, o/ the diamidc , second sphere dipole-dipole interactions ini]nci)tc the slope log l)^| vs log ('/\nillic values. The easy synthesis and purification of the amide extractants make their use more attractive. IT -16.9 REFERENCES 1. C Musikas and H. Hubert, Proceedings of the Int. Symp. on Solv. Ext, ISEC 83. p. 449, Denver (1983) 2. C. Musikas, Inorg. Chim. Acta 140, 197 (1987) 3. C. Musikas, Sep. Science and Technol. 23 (12-13), 1211 (1988) 4. T.H, Siddall (III), M.O. Fulda, G.S. Nichols, DP 541 (1961 5. G.iM. Gasparini and G. Grossi, Sep. Science and Technol. 15,4, 825 (1980) 6. B.N. Laskorin, V.V. Yaskin, E.A. FiUipov, G.M. Chumakova, V.A. Belov and G.G. Arkhipova, Sov. Radiochem. 20,438 (1978) 7. P. Charpin, M. Lance, M. Nierlich, D. Vigner, N. Descouls and C. Musikas, Acta Crystallogr. Sect. C 42, 560 (1986) 8. P. Charpin, M. Lance, M. Nierlich, D. Vigner and C. Musikas, Acta Crystallogr. C 43, 231 (1987) 9. N. Descouls and C. Musikas, Jour, of the Less Comm. Met. 122,265 (1986) 10. N. Condamines and C. Musikas, Solv. Extr. Ion Exch. 6, 1007 (1988) 11. C. Musikas and H. Hubert, Solv. Extr. Ion Exch. 5, 151 (1987) 12. M.C. Charbonnel and C. Musikas, Solv. Extr. ion Exch. 6, 461 (1988) 13. E.P. Horwitz, D.G. Kalina, H. Diamond and L. Kaplan, Actinide-lanthanide separations, p. 43, ed. G.R. Choppin, J.D. Navratil, W.W. Schulz, World Scientific Publishing (1985) IT - 16.10 SPENT FUEL REPROCESSING~A PERSPECTIVE S.V. Kumar Head, Process Engineering and Systems Division Bhabha Atomic Research Centre 1.0. Introduction Nuclear Energy is now emerging as a clean and vialble source to meet the energy demands in the world. The share of Nuclear Power in the electricity production stands between 75%-25%in 13 countries. Worldwide in 1989 about one sixth of the total electricity generation was produced by nuclear power plants. At present 425 reactors are operational in 26 countries, 102 under construction in 30 countries and 75 are planned in 35 countries. The percentage of electricity generated in various countries is shown in table 1. In this context, the back end of the fuel cycle assumes greater importance to facilitate closing of the fuel cycle . The decision regarding the choice of the fuel cycle is influenced by many factors such as the technology, the economics, the environmental impact, the prcliferation risk, the scale of the nuclear programme, the national energy strategy etc. Countries like France, United Kindgom, Germany, USSR, Japan and India have shown preference for the reprocess-recycle option whereas the policy in USA, Canada, Italy, Sweden and Italy is for long term storage keeping wide open the options of reprocessing or disposal at a later date. Some of the technological aspects in this choice are discussed in this paper. 2.0.Spent fuel Management strategy Ideally the nuclear fuel would remain in a reactor until its burn up is complete. In practice the limits of performance of the fuel will be reached due to the following criteria (1) loss of fissile material and accumulation of fission products affecting the neutron economy (2) the onset of mechanical defects in the cladding (3) the onset of fuel swelling The fuel must therefore be removed from the reactor in order to replenish the fissile worth and the mechenical structure. An approach to the ideal situation is achieved by attempting higher burn up values by improving the fuel design. The technical options available for the management of spent fuel from thermal reactors are IT - 17.1 (1) the once-through mode or direct disposal of unreprocessed fuel (2) Delayed reprocessing (3) Prompt reprocessing and subsequent use or storage of products. The relative merits of these options are discussed below. It may be noted that economic aspects are not taken into consideration since it is dependent on various national factors and pricing policies. 2.1. Once-Through mode or direct disposal This option is based on direct disposal of the spent fuel after discharge from the reactor. A discussion on the merits of this options should take into account the fact that reprocessing involves the generation of high active waste(HAW) which will be immobilised in a vitrified matrix and the medium level waste will be immobilised in cement matrix before disposal. The comparison is based mainly on the chemical and radiological characteristics of the spent fuel and the vitrified waste. Considering the specific heat output, it is found that the heat output per GW(e) yr. will be slightly greater for spent fuel during the initial period of 100 years. Subsequently it may be upto three times greater than the vitrified waste. In view cf the higher quantities of Plutonium in the spent fuel, the total radioactivity and consequent toxicity will be about one or two orders of magnitude greater than the vitrified waste over a longer period of one thousand to one million years. Further, the spent fuel contains 1-129 which is absent in the vitrified waste. The above factors have significant influence on the design of suitable packages and develop satisfactory matrices to immobilise the heterogenous waste present in the spent fuel. 2.2 Delayed reprocessing This strategy assumes that the spent fuel is stored for about 50 years and then the reprocessing is carried out. The main points to be considered here are the impact on the radioactive discharges and the operator dose. The important fission products in reprocessing are Zr-95/Nb-95, Ru 106/Rh iO6, Cs-137 and Sr-90. For Zr 95 / Nb-95(half life 64 days) a delay beyond 5 years has a negligible effect. In the case of Rul06/Rh-106(half life 1.01 years) there may be a marginal advantage. For Cs- 137 (half life 30.2 years) and Sr-90 (28.5 years) the delay for 50 years will reduce the activity by a factor of about 3. IT - 17.2 Therefore delayed reprocessing is not very effective in reducing the total activity of the discharges. It would be better to achieve the reduction in discharge activity by appropriate process flow sheet conditions and additional treatment facilities. Regarding the reduction in the operator-radiation dose, it is found that the reduction would be by a factor of 1.5 to 3. This may have a marginal impact on the shielding requirement. 2.3. Prompt Reprocessing and recycling This option has been investigaged extensively. The reprocessing-recycle system aides in improving the utilisation efficiency of uranium by an order of magnitude. It may be considered that the spent fuel is a "mine" and the reprocessing technology is the "technique" used to exploit the ore. It is estimated that this strategy would reauce the uranium consumption by nearly 40%. This would also result in a more optimised utilisation of the spent fuel storage capacity. Another important consideration in favour of reprocessing and recycling is the possibility of using the long lived actnides as reactor fuel which would eventually get converted to short lived fission products. This would significantly reduce the surveillance time of the Vitrified waste to only few hundreds of years. After the separation of useful constituents of the spent fuel, the fission product and the other transplutonium elements represent only about 3% of the initial quantity of the fuel so as not to pose any formidable challenge for concentrating and containing them for ultimate disposal. Thus, reprocessing the spent fuel and recycling the uranium and plutonium is considered as a prudent option for achieving maximum fuel economy as well as has many advantages froi the point of view of safe management of high active wast? generated in the fuel cycle operation. 3.0 Reprocessing in India 3.1. Background There are seven Nuclear Power Stations operating in India generating 1465 MWe. These include the two at Tarapur, two at Rajasthan, two at Kalpakkam and one unit at Narora. Except for the two units at Tarapur which are boiling water reactors(BWR) the rest of them are Pressurised Heavy Water Reactors (PHWR) . Seven more units are under construction at Narora, Kakrapar, Kaiga and Rajasthan. The nuclear power programme in India envisages the achievement of an installed capacity of about 10,000 MWe by the year 2000. The second phase of the nuclear power development programme will be in the area of Fast Breeder Reactors (FBR) which will utilise the plutonium produced in the thermal IT - 17.3 reactors. An experimental 50 MW(Th) fast breeder test reactor(FBTR) is operating at Kalpakkam. Studies are also in progress for the design of a prototype FBR of 500 MW(e) capacity. India has one of the largest thorium resources in the world. The development of nuclear power programme, therefore, has taken this aspect into account and planning for the eventual utilisation of this vast resource in the U-233-'T'h cycle. 3.2 Indian experience Reprocessing in India commenced with the commissioning of the Plutonium Plant at Trombay in January, 1965. The reprocessing of the spent fuel from the 40 MW(Th) research reactor CIRUS, was carried out in this demonstration plant. The metallic natural uranium aluminium clad fuel was chemically dejacketted and the separation of uranium, Plutonium and fission products was carried out using the Purex process. The plant was not only useful in generating trained man power and expertise for future plants but also helped in identifying areas of further Research and Development in areas like process development, behaviour of in-process equipment in the highly corrosive enviornment, the reliability of control systems, process performance with respect to material recovery, purity and accountancy etc. With the feed back from the Trombay Plant, the Power Reactor Fuel Reprocessing plant (PREFRE) was constructed at Tarapur for the treatment of the spent fuel from the Tarapur and Rajasthan Atomic Power Stations. This plant uses the chop-laach technique for the head-end and the purex flow sheet using uranous nitrate stabilised by hydrazine as the reductant. Several innovations such as pneumatic pulsing, air-lift control of solvent extraction columns, thermosyphon evaporators etc., were introduced in this plant. This plant has been operating since 1975 and has had several campaigns under International safeguards. This plant has provided valuable experience in material accounting practices to meet the International standards. This plant also provided experience in the safe in-land transportation of the spent fuel. The heavily shielded casks to transport highly active irradiated fuel were designed and fabricated in strict conformity with International standards. The experience gained in the logistic of fuel transportation from reactor site to the reprocessing site would be very useful in planning and selection of sites for the future reprocessing plants. To cater to the reprocessing of zircaloy clad, natural uranium oxide spent fuel from the Madras Atomic Power Station, a new plant is under construction at Kalpakkam. Some of the new features in this plant are IT - 17.4 - provision of standby set of process cells to extend the uselful life span of the plant to match with the reactor life. introduction of mixed maintenance co^-apt wherein the equipment requiring frequent maintenance like pumps, valves etc. are housed in shielded cubicles equipped with remote handling facility. This would result in considerable reduction of radiation exposure and also improve the availability factor of the plant since elaborate decontamination which is• necessary for contact maintenance is not needed. The plant will also have provision, as a separate line, the facility for reprocessing spent fuel from Fast Breeder Test Reactor (FBTR) Computers have been extensively used in the design for generation of piping layouts, P.I. diagrams and it is proposed to use the computers and micro processors for data acquisition, alarm monitoring, and some interlock systems. Keeping in view the Nuclear Power Programme, there will be a necess; • to construct additional reprocessing plants. In the design of the future plants, the emphasis would be on the standardisation in order to effect a reduction in cost and the lead times. As mentioned earlier the utilisation of thorium for the generation of Power is one of the important objectives of the Indian Nuclear Power Programme. As a prelude to achieving this, a significant number of aluminium-clad thorium metal and thoria fuel rods were irradiated on an experimental basis in the research reactor CIRUS and the irradiated material was reprocessed in a Pilot plant facility at Trombay to separate the U-223. An engineering scale facility to process the thorium fuel elements irradiated in CIRUS and DHRUVA reactors is being set up at Trombay. This facility would help in establishing the design concepts for the production plants to be built subsequently. 4.0 Decontamination/Decommissioning In view of the increasing number of nuclear facilities, the decontamination/decommissioning is assuming greater importance. 'Decommissioning' means the action taken at the end of the useful life of the facility, to retire the facility from service in a manner that provides adequate protection for the health and safety of the worker, the general public and the environment These actions can range from merely closing down the facility and a minimal removal of radioactive material coupled with continuing maintenance and surveillance, to a complete removal of residual IT - 17.5 radioactivity in excess of the levels acceptable for unrestricted use of the facility and its site. The major areas of interest in the decommissioning are as follows: (1) Pretreatment and chemical decontamination process for metal surfaces (2) Specific decontamination processes and equipment for concrete (3) Cutting and demolishing techniques and equipment for concrete (4) Segmeting techniques and equipment for metal components. The objectives of decontamination during decommissioning are to reduce the occupational exposure, to permit reuse of the item being decontaminated or to facilitate waste management. Application for remotely controlled equipment for decommissioning has considerable advantage in reducing the radiation exposure and the contamination occuring during decommissioning operation. A wide range of specialised manipulators and equipment is being developed to perform remote tasks such as inspection, maintenance and repairs. In designing the remotely operated equipment careful consideration is given to some of the important aspects like fail safe operation, reliability, easy maintainability, use of radiation tolerant components etc. 4.1. Indian experience With the commissioning of the 100 MW(Th) research reactor Dhruva at Trombay, an additional demand arose for the' reprocessing in the Plutonium Plant, Trombay. This necessitated replacement of certain equipment in the process cell. This opportunity was utilised to decommission the plant. The entire programme was carefully planned to keep the personnel radiation exposure as low as possible and minimise the waste generated dvring the operation. The decommissioning procedure comprised of several sequential steps. The campaign of internal decontamination of equipment and piping followed multiple decontamination routes. The maximum possible number of equipment was covered in a single route to minimise the quantity of decontaminants used so as to keep the resultant volume of radioactive liquid wast<=> low. After the internal decontamination, the task of decontaminating the interior surfaces of the cells and the exterior surfaces of equipment and piping was undertaken. The entry of personnel into the process cell was possible at this stage, since the radiation field had been brought down IT - 17.6 sufficiently. After dismantling and disposal of equipment and piping, high pressure water jets, steam, chemicals, pneumatic chippers etc., were used as appropriate to remove contamination on the cell interior surfaces. At the end of the decommissioning campaign, extremely low background radiation levels could be achieved, as a result, the subsequent installation work could be carried out with minimum protective gear. The experience gained during this exercise emphasised the importance of making provisions for decommmissioning during the design stage itself. The success achieved and confidence gained would enable undertaking effective decoininissioning tasks in the future as and when the need arises. 5.0. Future trends Some of the major thrust areas in the reprocessing would be as follows: engineering development to reduce the cost and improve reliability. These include improvements in equipment design. introduction of more effective process control systems based on micro processor and on-line instrumentation, etc. reduction in the wajstjs generation. This would involve changes in the process conditions, utilisation of electrochemical techniques thereby reducing the salt content in the waste etc. improved wast^; management §j:herues_ which would necessitate matrix development, separation of ac*-.inides which would bring about a reduction in tLi surveillance period etc. introduction of robot ics and remote handling equi pment for maintenance, which would considerbly bring down the operator radiation dose requirements. IT - 17.7 TABLE-1 Nuclear Power's share of electricity production,1989* France 74.6 Belgium 60.8 Republic of Korea 50.2 Hungary 49.8 Sweden 45.1 Switzerland 41.6 Spain 38.4 Finland 35.4 Germany,Fed.Rep. 34.3 Bulgaria 32.9 Japan 27.8 Czechoslovakia 27.6 United Kingdon 21.7 United States 19.1 Canada 15.6 USSR 12.3 Argentina 11.4 German Dem. Rep. 10.9 South Africa 7.4 Yugoslavia 5.9 Ne'cherland 5.4 India 1.6 Brazil 0.7 Pakistan 0.2 * IAEA Newsbriefs, April 1990 IT - 17.8 PLUTONIUM CHEMISTRY AIU FAST REACTOR FUEL REPROCESSING P.R.VASUDEVA RAO and T.G.SRINIVASAN Radiochemistry programme Indira Gandhi Centre for Atomic Research Kalpakkam 603 102, Tamil nadu, India INTRODUCTION: The reprocessing of irradiated fuel is characterised by its complexity, caused by the interplay of the chemistry of a number of elements in the irradiated fuel. The primary target of fuel reprocessing being the isolation in a pure condition of the strategic fuel materials (uranium and plutonium), the chemical steps naturally emphasise the need for thorough decontamination of the recovered materials from the impurities (fission products), and the need for near quantitative recovery of the materials, ie., minimisation of losses at each step. Reprocessing of fast reactor fuel involves the recovery of large quantities of plutonium from fuel of high plutonium content irradiated to high burnup levels. The factors mentioned in the above para, viz., recovery and decontamination, natural\y assume even more importance in fast reactor fuel reprocessing. Due to the high costs involved, there is also an added incentive in reducing the number of process steps or devising novel solutions to the chemical problems involved. Reactor fuels have conventionally been reprocessed by the "Purex" process, which involves the dissolution of the fuel in nitric acid followed by solvent extraction using Tri-n-butylphosphate for the recovery of uranium and plutonium. The presence of large concentrations of plutonium in the process steps of fast reactor fuel reprocessing introduces a number of new considerations in design of flowsheets, that do not usually apply to thermal reactor fuel reprocessing. This paper tries to highlight some of these aspects, and also presents a brief account of the studies carried out in this area in the Radiochemistry Programme (RCP) at IGCAR. DISSOLUTION: The dissolution step is considered as a vital step in the case of reprocessing of uranium plutonium idixed oxide fuels. Though uranium oxide dissolves easily in nitric acid, the "thermodynamic solubility" of plutonium dioxide in nitric acid is limited, and therefore fuel material containing "free" plutonium dioxide or Pu-rich dioxide, on dissolution in nitric acid, leaves a residue of Pu-rich oxide that is very difficult to dissolve. This PuO~ is dissolved usually by application of drastic dissolution conditions, viz., use of HF and high temperatures, necessitating the use of dissolvers made of special corrosion-resistant materials. The importance of this aspect in reprocessing of mixed oxide fuels used in the fast reactors can be gauged from the fact that the solubility of the mixed oxide in nitric acid has been considered as one of the specifications for the oxide fuel. At this point, it is not irrelevant to mention that the dissolution of PuOp is an important aspect of recovery operations that arise during fuel fabfication and many other steps. It is therefore not surprising that the IT - 18.1 dissolution of PuCL in various media have been studied by a number of workers. A review of th^'s subject by Ryan and Bray [1] has emphasised the fact that the studies have been carried out on a variety of oxide samples with different characteristics employing a variety of experimental conditions, which makes it very difficult to compare the results. For dissolution of PuCL in pure nitric acid, there are very few quantitative data; for nitric acid containing additives such as HF, more studies have been reported, but still, the mechanism of dissolution has not been thoroughly understood. To bring about the dissolution of PuO~ without the use of corrosive agents such as HF, which was the practice till recently, use of electrogenerated Ag(Ii) oxidant has been recently established [2], and this process is likely to be a boon to the Plutonium chemists. Studies in the Radiochemistry Programme at KalpaKkam have focussed on the use of reducing agents instead of oxidising agents. It has been established 3] that reducing agents could be perhaps as efficaceous as oxidising agents aiding the dissolution of PuO? at least in hydrochloric acid medium. Fig.1 tisents some of the results that support the application of reductive methods '•if PuCL dissolution. A novel concept of photochemical dissolution of PuCL in ' ! has also been evolved in our laboratory [4], This method consists of the i">e of U(IV) reductant, with simultaneous light irradiation As seen in the results presented in Table I, the presence of light enhances greatly the rate if dissolution of the oxide in HC1 medium, and in some experiments, complete Hssolution has been acheived at room temperature in reasonable periods of time. This has evidently opened up an area of investigations - the effect of light on the oxide dissolution mechanisms - though the applicability to large scale dissolutions is perhaps far away. SOLVENT EXTRACTION After the dissolution step, the next important step in the Purex process is the solvent extraction step. As mentioned earlier, tne Purex process involves the extraction of Pu as the TBP complex. TBP extracts only the tetravalent and to a lesser extent, the hexavalent plutonium. Trivalent Plutonium is not extracted. Since all these three oxidation states are possible for plutonium in the nitric acid media employed in Purex, there is a possibility of loss of plutonium into the waste streams unless all the Pu is "adjusted" to be in the tetravalent state. Thus, in reprocesssing of Pu-rich fuels, the adjustment of Pu oxidation state before the extraction step assumes added importance. Conventionally, this had been done through the addition of sodium nitrite, which reacts with both Pu(III) and Pj(VI) in nitric acid medium to convert them to Pu(IV). The use of sodium nitrite is not preferred now since it will result in increased high active solid waste. Use of gaseous nitrogen oxides has already been established as an alternative, and is in fact in use in the Indian plants also. Electrochemical methods have the advantage that they will not need addition of any extraneous chemicals. The reprocessing flowsheet for the FBTR fuel incorporates an electrochemical step for adjustment of oxidation state. TBP extraction: The need to extract Pu in large concentrations brings in a few interesting complications in the solvent extraction steps. The first is the necessity to ensure that very little Pu is lost in the raffinate (waste) streams. When the TBP phase is increasingly loaded with Pu and U, the IT -18.2 availability of free TBP is diminished, and therefore, the distribution coefficients for the extraction are also reduced. It is not practical to employ low loadings of TBP, since the availability of more free TBP will also lead to extraction of the unwanted fission product contaminants. Thus, one is forced to choose between Pu loss and poor fission product decontamination! Very recently, however, an elegant solution to this problem has been proposed by German workers at the Institute for Hot Chemistry at Karlsruhe [5]. They have proposed to make use of the increase in extractability of the Pu(IV) solvate, and the simultaneous reduction in the extractability of the U{VI), with increase in the temperature. They have concluded that for operations at temperatures of 50 deg C or higher, Pu losses can be kept very low even while maintaining a high level of TBP loading. Tests in the miniplant MTLLI have given results that confirm this hypothesis, though one would perhaps like to wait for full scale plant experience with such flowsheets. Third Phase Formation: Another important aspect of plutonium chemistry that creates complications in the large scale extraction is the formation of a "THIRD PHASE" during extraction. When TBP is loaded with plutonium beyond a certain limit, the organic phase splits into two, a light phase consisting mainly of the diluent, and a heavy phase (the "third phase") consisting mainly of the plutonium-TBP solvate. Now, the concentration of Pu in the third phase is sometimes as high as 180 g/L, and therefore, segregation of a third phase during plant operations involves the danger of criticality, if sufficient quantity of the third phase is accumulated. Besides this problem, third phase formation also leads to dynamic instabilities in the extraction systems, and the possibility of loss of plutonium. Thus, it is necessary to always keep in mind the limiting organic concentration for third phase formation (LOC) while designing flowsheets for Pu extraction. This evidently limits the maximum loading of TBP that can be realised, and consequently influences the fission product decontamination. Third phase formation is not unique to extraction of Pu or to extractions by TBP. In fact, the amine extraction systems for Pu separation and purification had third phase formation as the main drawback, though amines are among the ideal systems for Pu recovery. In thermal reactor fuel reprocessing, the concentrations of Pu in TBP phase do not reach levels that might result in third phase formation. In fast reactor fuel reprocessing, however, the danger is very real. In the absence of sufficient data and a thorough understanding of the phenomenon, we are forced to employ flowsheets with low Pu loadings that limit the throughput as well as fission product decontamination. What are the factors that influence third phase formation in TBP extraction of Pu ? This has been investigated in some detail in our laboratory. LOC values have been measured as a function of a variety of parameters, that have led to some understanding of the phenomenon. Our investigation [6,7,8] is in fact the only detailed innvestigation on this subject. Among the parameters that can influence third phase formation, the most important are temperature, aqueous phase composition, TBP concentration and the nature of diluent. The most critical, perhaps, is the temperature. The LOC values rise sharply with temperature, and it has been predicted that no third phase formation will occur in the Pu(IV)-HN0-.-30 %TBP system for temperatures IT - 18.3 higher than 50 deg.C [9]. The data on the LOC measured as a function of nitric acid concentration for 30 % TBP and 15 % TBP indicate that the limiting ratio of Pu/TBP for 15 % TBP is much higher than that for 30 % TBP[10]. This means that from the point of view of fission product decontamination, use of lower TBP percentages can have some advantages. Fig.2 shows the effect of aqueous phase ionic strength on the LOC. It is clear that increase in ionic strength will It ad in in<.roa^e in LOC values. This means that the presence of fission products or other ionic species in the aqueous phase will only hava a positive influence, r i o. 3 sh .'•,;<• the effect of the addition of small concentration of polar "modifiers" tv the organic phase on the LOC value. Thp increase in LOC values with s^all additions of polar compounds underlines the importance of the interactions between the various species in the organic phase. A decrease in the carbon chain length of the diluent has been shown to increase the tolerance for the p1utoniurn solvate in the organic phase [9]. In Fig.4, we see the effect of organic ohase U(V1) concentration on the LOC values. The data indicate that the presence of uranium in the organic phase reduces the LOC for third phase formation. All the above data provide us an indication of the interplay of various factors that influence third phase formation. It is clear that one has to take these factors into account while designing flowsheets for Pu recovery by TBP extraction in fast reactor fuel reprocessing. The above data also provide some qualitative insight into the phenomenon itself. It is clear that the plutonium nitrate - TBP solvate is "incompatible" with the "inert" diluents used, and it is the presence of "free" TBP that stabilises the organic phase. A decrease in the "free" TBP concentration below a certain limit forces out the solvate as a third phase. Additives to the organic phase that are polar in nature aid in "stabilisation" by way of interaction with the solvate. An increase in the ionic strength of the aqueous phase will result in decrease in the water content of the organic phase, which implies an increase in the concentration of "free" TBP. One can also explain the increase in LOC with increase in nitric acid concentration as due to the interaction between the nitric acid solvate and the plutonium solvate in the organic phase, but we need more evidence to confirm such an interaction. PARTITIONING: Another aspect of Purex process that is influenced by the high plutonium concentrations is the partitioning of Pu from U. In the case of thermal reactor fuel reprocessing, this is done by exploiting the ease of reduction of Pu(IV) by U(IV) to Pu(III), which is not extracted by TBP. When the Pu concentrations are high, the U(IV) requirements also go up significantly, and this method of partitioning is clearly unsatisfactory. "Electrochemical partitioning" methods are being investigated in many countries including India. The reduction of Pu(IV) to Pu(III) is carried out electrochemically in this method. While this method will have many advantages over the conventional method, its applicability to high Pu concentrations will have to be established. Interestingly, the U.K flowsheets use sulphate complexing to partition Pu from U [11]. One interesting method of partitioning Pu from U could be via the oxalate precipitation step. Conventionally, oxalate precipitation is employed for the "reconversion" steps, that is, to convert the Pu in solution to the solid form. The solubility of uranyl oxalate being significant, it can be expected IT - 18.A that by proper design of flowsheets, the precipitation step could be used to separate Pu from II. The carryover of uranium in the precipitate will depend upon the solubility of uranyl oxalate under the precipitation conditions; however, some contamination of Pu by U might be allowed if the Pu is to be used for (U?Pu) mixed fuel fabrication. The design of such a partitioning step will need accurate data on the solubility of the oxalates of Pu and U. It is interesting to note, in this context, that very few detailed studies have been reported on the solubility of uranyl oxalate and plutonium oxalates in nitric acid - oxalic acid media. Further studies are clearly needed in this regard. It is especially important to carry out the Measurements by two different procedures - by preparing the oxalate and equi1ibrating with the oxalic acid- nitric acid solution, and by carrying out actual precipitation from the acid solution. Careful interpretation would then be needed to reconcile the results obtained by the two different methods. In RCP, recently, the solubility of uranyl oxalate under the Pu oxalate precipitation conditions was measured by the oxalate equilibration route, and the data [12] seem to indicate that a partitioning step based on oxalate precipitation is indeed possible. A similar concept of Th(IV)- U(VI) partitioning has been advocated for use in recovery of U-233 [13]. RECONVERSION: A number of methods have been investigated [14] for the recoversion of Pu. Among these, oxalate precipitation [15] nas been the most widely employed. Direct crystallisation of uranyl and plutonyl nitrates has been recently advocated as an alternative method for reconversion of uranium and plutonium[16]. Detailed information on such a process is not however available, especially in the case of Pu. Plant scale experience on this method is limited even for uranium. This method, in principle, could have a number of advantages. The applicability of this method for Pu system has to be established. The precipitation of Pu as oxalate leaves supernatant solutions with Pu concentrations ranging from a few mg/L to a few hundred mg/L, depending upon the precipitation conditions. The necessity to recover this Pu, has resulted in a search for easy and efficient, methods of destruction of the oxalate in the supernatant, which is necessary before recovery could be attempted. The classical methods of destruction involve the use of oxidising agents-such as permanganate , which add to the solid radioactive wastes. In RCP, a photochemical method for the destruction of oxalate, using the excited urany) ion as the oxidising agent has been explored[17]. The results of these studies (see Fig.5) indicate that the photochemical destruction of oxalate is indeed an attractive alternative. The method is simple and efficient. The uranium added could evidently be recovered along with Pu. Very often, the Pu streams which provide input to the oxalate precipitation step do contain small concentrations of uranium (from the partitioning step). Thus, in many cases, addition of extra uranium is not likely to be necessary. It is, hoever, necessary to point out that the method has been demonstrated only on a laboratory scale, and scaling up for the plants will involve the optimisation of the means of irradiation. APPLICATIONS OF PHOTOCHEMISTRY: Though the photochemistry of uranium has been studied rather extensively, the photochemistry of Pu has not been given sufficient attention. This is IT - 18.5 partly in view of the poor quantum yields of the photchemical reactions involving Pu that have been studied so far. However, a few interesting applications of photochemistry seem to be possible in Pu processing. One of these involves the application of the photochemical reduction of Pu(IV) to Pu(III) in the partitioning step. It has been noted [18] that UV light irradiation significantly improves the rate of reduction of Pu(IV) co Pu(lII) by hydrazine in nitric acid medium. Hydrazine is employed in the reductive partitioning step to kill the nitrite ions which interfere in the reduction. The direct photochemical reduction of Pu(IV) by hydrazine, complemented by the even more efficient photochemical reduction through the photochemical production of U(IV) by reaction of U(V1) with hydrazine, is perhaps an alternative that needs more examination. Though the reduction of Pu(IV) directly or (through U(IV)) by TBP in the organic phase is also possible, this alternative is less attractive because of production of deleterious by- products such as DBP. The use of photochemistry for external production of U(IV) is certain to be a competitive method. However, in fast reactor fuel reprocessing, in view of the high concentrations of Pu involved, in-situ generation of the U(IV) reductant will be preferred. Another area where the photochemistry of Pu will find an application is the monitoring of the process streams. Due to the high inventories involved, there is a clear need to monitor the movement of the stragetic material in a reprocessing plant. Conventional techniques utilising the radioactive properties of Pu are under development in various countries, including India. In the Savannah River Plant in USA, on-line fibre optic spectrophotometry has been developed [19] for the determination of uranium,and similar determination of Pu will certainly be possible. For applications to streams with low concentrations of Pu, such methods will evidently have very limited use. This has opened up a search for the fluorescence of Pu, which, if successful, may have interesting applications. At present, however, this seems to be a challenging and difficult task. REPROCESSING OF URANIUM PLUTONIUM MIXED CARBIDE FUEL: The Fast Breeder Test Reactor at Kalpakkam uses a Pu-rich mixed carbide as the driver fuel. The reprocessing of this fuel will perhaps be discussed ir. detail in another paper in this symposium- It is however necessary to mention here that the dissolution of the mixed carbide in nitric acid leads to tht formation of a number of organic compounds that cause considerable: interference in the solvent extraction steps. Destruction of these compounds is therefore an important aspect of the reprocessing. Apart from this aspect, from the point of view of Pu chemistry, the reprocessing flowsheet for the carbide fuel is expected to be identical to that of oxide fuels. REFERENCES: 1. J.L.Ryan and L.A.Bray, "Actinide Separations", J.D.Navrati1 and W.W.Schulz (Eds), ACS Symposium series 117, American Chemical Society (1980), p.499 2. L.A.Bray and J.L.Ryan, PNL-5657(1985). 3. A.M.Shakila, T.G.Srinivasan and K.N.Sabharwal, Nucl.Technol., 88,290(1989). 4. B.S.Panigrahi, T.G.Srinivasan and P.R.Vasudeva Rao, being published in IT - 18.6 Radiochimica Acta. 5. H.Schmieder and G.Petrich, Radiochimica Acta, 48,181(1989). 6. T.G.Srinivasan, M.K.Ahmed, A.M.Sh3kila, R.Dhamodaran, P.R.Vasudeva Rao and C.K.Mathews, Radiochimica Acta, 40, 151 (1986). 7. T.G.Srinivasan, R.Dhamodaran, P.R.Vasudeva Rao and C.K.Mathews, Sep.Science and Technol., 23, 1401 (1988). 8. T.G.Srinivasan, M.K.Ahmed, R.Dhamodaran and P.R.Vasudeva Rao, Proc. Radiochem, and Radn. Chem. Symp., Tirupati (1986), paper no. CT-42. 9. Z.Kolarik , Proc. Int. Conf. or, Solvent Extraction, 1977,Toronto Vol I, p178(1979), R.H.Lucas, G.M.Ritcey and H.W.Smith (tds) CIM Special Volume 21. 10. T.G.Srinivasan, M.K.Ahmed, A.M.Shakila, R.Dhamodaran, P.R.Vasudeva Rao and C.K.Mathews, Proc. Seminar on Fast Reactor Fuel Cycle, Kalpakkam (1986), Vol I, p.192. 11. A.L.Mills, Proc.Symposium on Fast Reactor Fuel Reprocessing,Dounreay (1979) Society of Chemical Industry (London),1980,p 7. 12. B.S.Panigrahi, T.G.Srinivasan and P.R.Vasudeva Rao, Proc. Radiochem. Radn. Chem. Symp., Kalpakkam (1989), paper no. RA-35. 13. A.Ramanujam,P.J.Dhami,V.Gopalakrishnan,A.Mukherjee,R.K.Dhumwad, BARC-1486(1989). 14. "Plutonium Handbook", Ed. O.J.Wick, Gordon and Breach, NY, (1967),Chapter 1b,p 553. 15. E.W.Mainland, D.A.Orth, E.L.Field and J.H.Radke, Ind. Eng. Chem., 53, 686(1961). 16. K.Ebert, E.Henrich, R.Stahl and U.Bauder, Proc. Int. Conf. on Sep. Science and Technol., Toronto (1989), M.H.I.Baird and S.Vi^ayan (Eds), Canadian Society for Chemical Eng., p.346. 17. T.G.Srinivasan, S.K.Nayak, R.Dhamodaran and P.R.Vasudeva Rao, Proc. Radiochem. and Radn. Chem. Symp., Bombay (1987), paper no. CT-09. 18. L.M.Toth, H.A.Freidman and J.T.Bel 1, CONF-770506-1(1977). 19. D.R.Van Hare, P.E.O'Rourke and W.S.Prather, DP-MS-88-186 (1988) IT - 18.7 TABLE I PHOTOCHEMICAL DISSOLUTION OF PuO2 IN hCl AT ROOM TEMPERATURE [HC1] = 3 M; [U(IV)] = 0.015 M; [U(VI)] = 0.015 M; [N^] - 0.2 M WEIGHT OF PuO2 TAKEN = 100*5 mg, TOTAL VOLUME OF SOLUTION = 60 mL S.No. ADDITIVES % DISSOLUTION WITHOUT LIGHT WITH LIGHT 1 2.2 5.0 2 U(VI) 1.8 6.4 3 N2H4 1.6 13.8 4 U(VI), N^H4 1.4 69.0 5 U(IV), N2H4 6.3 70.3 6 U(IV) - 62.0 IT - 18.8 Fig,1.Effect of Fe(II) and Hydra- zine on the Dissolution of PuO9 in 7 M HC1 at 373 K. Acidities "are initial acidities.a)7 M HC1; b) 7 M HC1 + 0.65 M hydrazine; c) 7 M HC1 + 0.02 M Fe(II); d) 7 M HC1 + 0.02M Fe(II) + 0.27 M hydrazine; e) 7 M HC1 + 0.1 M Fe(II) 50 100 )50 200 2S0 300 TIME(mln) Fig.2.Variation of LOC^of Pu(IV) with Aqueous Phase Ionic Strength at 303 K. a) NaNO.,: Nominal Acidity 2M; b) NaNO^:Nominal Acidity 2.7M; c) NaNO^:Nominal Acidity 3.7M; d) La(Nd^)..:Noniinal Acidity 3.5M. IT - 18.9 I 30V.TBP/n-DD 30V.TBP/(nDD*2 5%n- OCTANOU o 3 0 % TDP/(tiDD*2 5 % BENZENE) 70 30 T - 303 K 65 ~ 85 - I 60 v. tO - o> IK E 55 *75 3 O °-70 50 cig.5.Halftimes for Photochemical Destruction of Oxalate as a function of Uranium Concentration. i. 5 9 10 [u] mg/ml IT - 18.10 REPROCESSING OF HIGH CONTENT Pu FUELS Balasubramanian G.R. IT - 19 DEVELOPMENTS AND AUTOMATION IN POREX PROCESS CONTROL ANALYTICAL MEASUREMENT SYSTEMS A. Ramanujam Fuel Reprocessing Division Bhabha Atomic Research Centre, Trombay, Bombay- 400 085, India SUMMARY The fuel reprocessing facility based on Purex process depends on efficient process control analytical measurement systems for its successful operation. The process control laboratory plays a vital role in catering to these requirements. This paper describes the various efforts put in to improve its performance •^abilities in three major areas of operation, via. sample nandling, analytical and data processing. In developing automation aids and anlytical techniques, apart from the special nphasis put on reduction in personnel exposure to radiation and .ime required for analysis, due consideration has been given to 1 ,erational reliability and safety of the system. KEY WORDS- PUREX PROCESS, PLUTONIUM ANALYSIS, URANIUM ANALYSIS) i. INTRODUCTION All over the world, Purex process is the most widely accepted technique for the reprocessing of irradiated uranium fuel. It involves the separation of uranium and plutonium from the bulk of the fission products by solvent extraction with tributyl phosphate(TBP) and their subsequent purification by solvent extraction and ion exchange technique. The major tasks of these plants can be classified as (a) attainment of a high degree of decontamination of uranium and plutonium from fission products and from each other with extremely low losses of U and Pu (b) discharge of the bulk of the fission products in highly concentrated form and (c) discharge of medium and low level aqueous, organic and gaseous effluents with stringent controls and within permissible levels of radioactivity. Being remotely operated, such facilities depend heavily on process control analytical measurement systems (1,2) for their successful operation as it is the main agency that comes in direct contact with the sample which represents the status of a process stream at any given time. This calls for extensive process and plant control, inventory and quality control measurement requirements which have to be met by the Control Laboratory. For the Control Laboratory to play its role effectively, the sample analytical schedule should take into account the speed, accuracy and frequency with which these results are required, based on the end use of the analytical data, for a predetermined number of samples1 per day. Further, strict adherence to the prescribed quality assurance practices, both internal and external have to be maintained to have a check on the quality of the measurement systems. These aspects and the various control measurement systems used in the laboratory have been dealt in detail in one of the earlier symposia by Mani(2). IT - 20.1 The process oont fc-1 1 aborat'Ty oper^t.t-s in hazardous radiation environments and requires considerable modifications in sample handling and analytical procedures, The routine operations carried out in round the clock shifts by this laboratory can be grouped as: (1) Sample Handling Tasks: Sample receipt and storage, individual sample re+reival from stored location , sample deoapping for analysis and disposal of unused sample.-? at regular intervals. (2) Analytical Tasks: Analysis of thy- samples for various components like acidity, uranium, Plutonium, fission products, gross beta and gamma aotivi t j es , etc. , as per the analyti ••••)] protocols and priori t. i es. (3) Analytical Data Processing: Calculation of assaly'• i cal results, checking for their validity, compilation and report, i n a. Tn a !aboratorv of this type, automation in sample handling and advanced analytical instrumentation can substantially >-;-n him •.•=-.-. tii>- " -lytical out-put and nan bring about considerable reduction In personnel exposures to radial.ion inherent in thesf ta s j-1;.. w i-ale cr. m put eri sed data handling systems' can speed up th data prooessi ng ability. Tne ma j nte-nanee arid operation f tli is laboratory is expensive in i-erms of trained man-power, one way use of lab. equipments and dispensable analytical wares and once through ventilated active lab. area with modern hot cell areas costing more than Rs . F> lakhs/ oq.M. Therefore efforts towards improvements !n analytical methods, enhancements in analytical output and capability are worth pursuing as they amount to better utilisation of the lab. facilities. Keeping this in view, continuous efforts are made at FRD laboratories to bring about af?- much improvement as possible in the above areas of operation and this presentation summarises nu, inly these devlopments and th- experience in their utilisation. II. AUTOMATION IN SAMPLE HANDLING TASKS Whereas sophisticated and automated analytical instrumentation systems are available commercially and can be adapted to analytical tasks, sample handling and aliquoting tasks are of special nature and call for in-h<>use developmental efforts for automation. FKD is addressing this problem in collaboration with Division of Remote Handling and Robotics ( 3).' in these efforts, due consideration should be given to the operational reliability and safety of the system. However, it is essential that the automated systems should be operated under constant monitoring by laboratory personnel but with minimum intervention. £.Li>Xiii££ JTitclIlly: About 100 medium active samples are received per day and about 300 samples are required to be kept in storage at any given time. These samples would have an individual contact dose of 500 mr/hr or less. Sample storage, search and retrieval for analysis and relocation after analysis account for the major fraction of the 1otal dose received in any analysis. In order to facilitate this task, a computer controlled sample storage facility is under fabrication. It will have provisions IT - 20.2 for storing 300 samples at Ccsired locations in a serial arrangement and for selective retrieval of individual sample from among the stored samples, on command. A schematic diagram of the unit is shown in Fig.l. Pecanper Unit : In routine analysis, manual opening and closing of the screw caps of sample vials entail contact dose to the analyst. To avoid this exposure, a mechanical decapper/recapper set up has been in use for the past 6 years in the laboratory for opening and closing of double capped polypropylene sample vials. Though the unit is working very satisfactorily, it is rather cumbersome to operate and is being replaced by an electrically operated decapper . SameLe Disposal Uoli = Sample disposal after analysis is another task which involves considerable exposure to radiation. To alleviate this problem, a semi automated sample disposal system has been built for emptying, rinsing and disposal •f medium active sample vials. Thi system comprises a magazine i'or stacking 10 sample vials, the "pick and place" pneumatic robot for sample vial movement, puncturing station with two needles for piercing the vial cap and a vacuum set up for draining sample by suction. This unit has been in operation for the past three years without any problem and more than 20,000 samples have been disposed off with this unit. Al.tguobj.nw JQsyJLcs = In any analytical operation, the most .important task is the accurate aliquoting of the sample under analysis. The remote pipetting device used for aliquoting should be very accurate, reliable and maintenance free. To meet tliese requirements , a commercially available Metrohm Dosimat Unit has been modified. The Delivery Unit along with piston burette is located outsid* a hot cell and is connected to a SS pipetting head by a capillary bore teflon tube which goes into the cell through a standard ball and tong pipe. A disposable tip attached to the pipetting head is used for pipetting the solution from the deuapp-ed sample. The piston burette and the teflon tube are* filled with water so as to reduce the volume of air pocket above th*-; sample in the tip. After sucking about lml of the sample into Lho tip, the same is delivered as distinct aliquots of smaller sir-::a, rejecting the first aliquot. The delivery volumes have an error within •»/- 1.0 %. The unit has been in continuous use end i,.jre than 15,000 pipettings have been successfully carried out in the past 8 years. L<QJLZiiQXX fifib&tI2£Y-Sli>.PiasflLf c : It was felt that all the sample handling and remote aliqnoting tasks can be sucessfully met by a computer controlled robot in combination with a sf .iple storage .facility, decapper and pipetter device. With this in view, a ^hemical laboratory robot (LABOT-180) ha^ been designed and is in advanced stage of fabrication. The unit will carry out the above tasks in about 30 sequential steps of operation with option to intervene at any step. The robot will have a lifting capacity of about 180g and the actual area of operation would be-about 1sq.m. It is plenned to locate this unit along with the above IT - 20.3 ies in an enclosed fumehood. The unit is almost ready for testing and the necessary softwares are under develc i iv-fit. A schematic diagram together with the sample storage well is given in Fig.2. III. ANALYTICAL DEVELOPMENT TASKS Assays of U, Pu, H+ and fission products both in aqueous n.'id organic process streams at different, concentration levels constitute the major bulk of the analytical work load. The other tasks and analysis carried out include TBP solvent. and onion exchange resin quality, input acids and reagents, redox reagents like NaN02, U(IV), N2H4, Pu(IV) valency and trouble shooting studies on various process problems. Some of the recent developments which are being incorporated in this domain are outlined here. An example of trouble shooting operation is also included. AlEha Spfecixamfiixli! As^ay ol Eii222±2il0 In HAW. Sample ; In Furex Plant operation it is of vital importance to monitor the Plutonium concentration in the out going high active r.'iff inate (HAW) stream of the first (HA) cycle extraction column so as to ensure that the Plutonium losses are within the limits specified in the flow sheet. Abnormal losses, if any, should be detected as quickly as possible 00 that prompt corrective measures can be initiated. To reduce the time involved in the analysis, a direct alpha spectrometric technique has been standardised (4) for the assay of Pu239+240 without its prior separation from impurities. A diluted aliquot of the sample on a pl^nchet disc is dried, fired and analysed by alpha spectrometry for Pu239+240 in presence of Am241, Cm242 and fission products. From the observed spectra, Pu239+240 content la assayed quantitatively by means of computer aided techniques. As the isotop.lo compositions of Plutonium in these samples are usually known, it is possible to calculate the total Plutonium concentration. The results obtained, for several HAW samples from different campaigns are given in Table 1 along with Pu values obtained by HTTA extraction procedure. The results obtained are satisfactory for process control analytical applications when the Pu 239+240 content in the disc is above 1.5x10 ' Aig. Our experience indicates that this technique is useful for the analysis of HAW sample when the Pu concentration is above lmg/1. While the counting is in progress it. is possible to predict the approximats Pu content in the source from the spectra being acquired. AU Plutonium species including polymers are accounted for by this technique. This system is being automated. Injection Analy3la tax thfi EaJklmakion af Uranium : Flow Injection analysis(FIA) is a simple and elegant technique which finds increasing applications as a process control analytical technique in many industries. In general it involves injection of a sample aliquot at a constant rate irto a steady flowing stream of reagent and passing this reagent-sample mixture through IT - 20.4 a suitable detector. It can be adapted for continuous process stream monitoring with good precision. In a laboratory environment, the same system can be modified to carry out various analyses. The FIA system can be divided into three modules: a) delivery units for maintaining steady flovT of the reagent ( and sample, if required) b) injection module for introducing sample and c) detector module. Injection port assemblies form a oruoial part of the FIA system and in the present instance, they had to be compatible with safe radiochemical practices. The different types of sample/reagent injection ports tested for the FJA analysis oi uranium (U(VI) and U(IV)) are shown in Fig.3. Under standardised conditions (5), the performance of the has been tested for the colorimetric response of U(VI) samples 410nm in the range 40 to 350mg/ml and for U(IV) samples at 65x)iur- in the range 15-120mg/mI in nitric acid medium using Metrohm 6 V.'. Fhotometer and a recorder as detector assembly. This technique under stopped flow mode can be used for the analysis of U(V1 in the range 0.5-4mg/aliq. by alcoholic thiocynate proceduiu. In all these cases the precision is found to be better than + 1.5%. With Nal well-type detector in the flow line, the grc^ gamma counting of the solution under flow is found to have precision of +/•- 5% Table 2 shows so-se of the. results obtained for uranium. It al.so includes the results obtained when the samples were injected into the closed injection chamber via permanently connected tubing. Such a mode can be used for on-line monitoring tasks where gradual variation or 3teady state concentration of uranium is to be monitored. Sample introduction using a six way valve is currently i:nder investigation for the same purpose. Feasibility of using this technique for Pu analysis is under investigation. So far, the moat commonly used methods for the analysis of U are thiocynate spectrophotometry and modified Davies and Gray redox titrimetry. With the introduction of automated spectrophotometc-r and titrator systems, the precision of these techniques have improved. Analyser System JUa Eraufias Control Analysis : In Purox process, spectroscopio terchnlques are mainly used for quality control analysis of various input chemicals and uranium and Plutonium products. Now with the recently acquired high resolution spectre analyser system coming on stream, the scope of this technique has expanded enormously. In addition to the regular tasks mentioned above, the high resolution sequential spectrometer with ICP source has vast potential for applications in process control analytical tasks which include analysis of trace level U and Pu in different complex matrices. Analysis of uranium in PPM and PPB levels is possible in various Purex waste and low level effluent streams and in thorium matrix (6) as encountered in Thorex proce33. Methods have been standardised for the analysis of trace level Pu in uranium and in Purex waste IT - 20.5 streams. The feasibility of using this technique for the analysis of Pu in diluted HAW samples is being established. It would greatly simplify this analysis. It is extremely useful for characterisation of intermediate level wastes and for estimating Na and Al in this complex matrix. The cationic load (mainly Na and Al) analysed by this technique agrees well with the total concentration of anions assayed by conventional techniques. It has been used for the analysis of U233 in wastes generated in U233-A1 alloy scrap recovery tasks during the fabrication of Kamini reactor fuels. Qaa Ch.romato<(raPhy lor Solvent Qualitx GojadkroJ, ; Quality control of input TBP and n-dodecane and DBF content of TBP streams are monitored by gas chromatographic techniques. Presently, methods are being standardised for the analysis of DBP in TBP without prior separation of uranium present in the sampl«=(7) As and when required, the solvent/diluent quality of the plant solvent Is assessed by Pu retention test (8). This technique is found to be more reliable than the conventional "Z" test. Efforts are in progress to identify and correlate some of the observed GC peaks for degraded solvents with the Pu retention values obtained. Grnnma Specfrrpmetr1c Techniques : For fission product analysis, a gamma spectrometric set up employing 4K MCA coupled to 60 cc HPGe detector is in use. Using the same system, Pu assay in product samples is possible by measuring the 129 Kev peak of Pu 239 (9). A LEPS detector is used for Pu isotopic assay end the concentration measurement in pure product samples (10). A total neutron counting technique has been standardised for the assay of Pu in ion exchange eluted product solutions under well defined matrix and geometry conditions. Test results obtained using the above techniques are presented in Table 3. N D Techniques for. ;fche. Estimation Q£ £U in Wastes : Estimation of Pu in solid wastes requires quick and efficient non-destructive assay methods. For monitoring Pu in glove box waste bags, a gamma spectromfetric technique with Nal detector is in use which measures Pu241 peak at 208 Kev and Pu239 peaks in the 384 Kev region (11) In presence of high fission product content, this method has some limitations. The total neutron counting technique has also been employed for this purpose(12). The use of this technique is being extended to the monitoring of Pu In plant discards like degraded TBP and anion exchange resins used in Pu purification (13). Some of the test results obtained for the above wastes are given in Table 4. Studies QD, Resin Degradation : Strongly basic quaternary ammonium type anion exchangers employed for the purification of P'.i undergo thermal, radiolytic and chemical degradation leading to impaired process performance. Under extreme conditions, significant amounts of white precipitate containing Pu are observed in eluted streams. The precipitate is insoluble in 8M HNO3 at room temperature but highly solube in dilute acid, water and TBP. After detailed investigations, the cause of the precipitate formation was traced to the presence of delinked IT - 20.6 quaternary ammonium functional groups in the eluted prodr.^t. The precipitate formation behaviour could be reproduced in the laboratory with tetra propyl ammonium hydroxide in 8M HNO3 (TPAN) and with leached extracts of anion resin obtained by its thermal degradation at 8M HNO3 at 80° C. The precipitate obtained from leached resin extract using Th (as substitute for Fu) was separated and an attempt has been made to identify the compound(14). Based on the thermal degradation studies, a method has been standardised for comparing the chemical stability of anion exchange resins. IV ANALYTICAL DATA PROCESSING Monitoring of more than 100 process tanks on a day to day basis entails enormous amount of data processing. A software package has been developed ( 15) which meets the full da-ta handling " requirements of the control laboratory. The software is a totally menu-driven package which can be used by the laboratory analysts with a few hours of training. The various facets covered under the programme are important instructions regarding sample receipt, location and analysis of samples in different areas, details to be checked for each sample on receipt, special information regarding each process tank including normal analysis required and their approximate ranges, anaLy'-ical protocols to be followed, selection of method and aliquot sise to be employed, detailed method menus, feeding analytical data via easily understood question-answer menus for computation of results, processing and comparison of data with stipulated flow sheet values with in-built alarm for deviations, compilation of data, display of current sample status, different output formats to meet the specific requirements of various groups of the process ranging from process control to accounting. The programme is written in BASIC. It is in constant use since May, 87, V. CONCLUSION The modifications incorporated so far have contributed significantly to the enhanced analytical output of the laboratoi/ which is presently handling about 5000 analyses per month during regular process compaigns. With the introduction of automatio-.i aids now under fabrication further improvements are expected. The main areas of thrust in the near future would be the development of more sophisticated remote techniques with high degree of automation for routine analytical operations and on-line analytical iiistrumentati ons to further reduce the laboratory analytical work load as well as the radiation exposure. VI. ACKNOWLEDGEMENTS The author wishes to express his sincere thanks to Dr.R.K. Dhumwad, Head, Laboratory Section, Shri. M.K. Rao, Head, Fuel Reprocessing Division and Shri. A.N. Prasad, Director, Reprocessing and Nuclear Waste Management Group for their keen interest and valuable suggestions during the preparation of this paper. IT - 20.7 VII. REFERENCES 1) F. Baumgartner and D.Ertel, J.of Radioanal.Chem., 58,11(1980) 2) V.V.S. Hard, IT-4, Radiochem. and Radiation Caem. Symp. IIT, Kanpur, 1985 (1985) 3) A. Ramanujam, V. Gopalakrishnan, R.K. Dhumwad, M.K. Rao, A.N. Prasad, R.K.Modi and M.S. Raraakumar, VI.C-4, Symp. on Advanced Remote Handling Systems and Automation in Nuclear Installations (ROSYMP-90), BARC, Bombay (1990) 4) P.V.Achuthan, A.Madhusudan, A. Ramanujam, R.K. Dhumwad, G.K. Gubbi, A. Ramaswami, Satyaprak^sh and P.R. Natarajan. BARC Report-1514 (1990). 5) A.H.Paranjape, 5.S. Pandit, S.S. Shinde, A.Ramanujam and R.K. Dhumwad, Nat. Symp. On Uranium Tech., BARC, Bombay(1989) 6) R.K. Dhumwad, A.B. Patwardhan, M.V. Joshi, V.T. Kulkarni and K. Radhakrishnan, RA-17, Radioohem. and Radiation Chem.Symp., IGCAR, Kalpakkam (1989) 7) S.C.Tripathy, C.S. Kedari, K.J. Parikh, S.K. Misra, A. Ramanujam and R.K. Dhumwad, Int. Conf. on Chromatography, (ICC-90), AMU, Aligarh(1990) 8) S.V.Kumar, M.N. Nadkarni, A. Ramanujam, M.Vengatesan, V. Gopalakrishnan and J.A. Kasi, BARC Report-933 (1977) 9) P. V. Achuthan, R.G. Bhogale, A. Rarnanujam, M.R.Iyer, D.N. Sharma and N. Raman, BARC Report-1437 (1988) 10) M.R. Iyer, S.J. Choithramani, II.G. Golani, U. Jambunathan and T.S. Laxminarayan, BARC-1321, 50 (1982) 11) U.Jambunathan et al, FRD, BARC, Unpublished Work. 12) S.C.Kapoor et al., FRD, BARC, Unpublished Work. 13) P.R. Rakshe, K.M. Michael, R.U. Yadav , V.P. Singh, K. Vljayan, N. Ramamoorthy and S.C. Kapoor, "Quantitative estimation of plutonium in degraded anion exchange resin and tributyl phosphate by neutron counting" ( Paper being presented at this symposium) 14) P.S. Dhami, V.Gopalakrishnan, A. Ramanujam, R.K. Dhumwad and M. Sundaresan, " Studies on the degradation of anion exchanger employed for plutonium purification" ( Paper b^ing presented at this symposium) 15) V.P.Kansra, P.V. Aohuthan, S. Sridhar, A.Ramanujam and R.K. Dhumwad, BARC Report-1499,(1990). IT - 20.8 Mo.l ± MALXSIS QE BM SAMPLES Using 300 mm Detector in Vacuum (IK MCA) No. Pu ug/disc Pu ug/disc Activity % Am Cm Spectro- o TTA Extra- Ratio ' Devia- dpm/ dpm/ metry X10 ction X10 Am/Pu tion disc disc 1. 77.00 77.30 0.29 < -1 3665 2638 2. 6.83 6.05 1.07 < + 13 1085 899 3. 12.18 9.55 1.33 <+28 2630 1603 4. 7.23 6.80 1.62 < +7 1928 1645 5. 2.53 2.95 2.02 <-15 1353 2135 6. 2.38 3.35 2. 19 <-29 858 2944 7. 5.33 4.58 2.21 < + 17 2019 1313 8. 2.60 2.55 3. 64 < +2 1562 2315 9. 1.65 1.53 5.31 < +8 1443 2460 10. 1.67 1.48 6.96 <+13 1913 2665 11. 1.70 3 .56 7.61 < +9 2139 2647 Table flc^ 2-fA i ESTIMATION QE UCVI> AJJB o(iv) M EM Flow rate HNO3 : 6ml/min Sample 1 ml/min Sample aliquot 0.5ml fixed A. DIRECT COLORIMETRY OF U(VI) AT 410nm Sample Closed-cup Open-cup Closed-cup U O.D % RSD 0.D %i\. SD. Gamma counts g/1 100 sec. 350.0 0.619 0.62 0. 589 0.51 3726 175.0 0.310 0.87 0.306 0.18 2046 116.7 0.216 1.58 0.210 0.54 1410 87.5 0.164 0.53 0. 150 0. 63 103? 43.8 0.083 1.07 " 0.083 0.62 558 350.0 * 0.625 0.45 - - 175.0 * 0.338 0.83 - - * Sample line directly connected to closed cup B. DIRECT COLORIMETRY OF O(IV) AT 650nm Closed cup * Open cup U{IV) O.D % R.S.D 0(IVU(IV)) O.D LS.D g/1 g/1 118.0 0.766 0.60 104.0 0.705 0.30 59.0 0.432 1.10 52.0 0.356 0.33 39.5 0.300 1.50 34.7 0.262 0.29 29.5 0.229 0.66 26.0 0.214 0.57 14.8 0.125 0.68 13.0 0.111 0.10 IT - 20.9 Table. HQ+ 3-fA&B> i. ASSAY QE PLUTONIUM BY HD TECHNIQUES A. ISOTOPIC ASSAY OF PLUTONIUM BY GAMMA SPECTROMETRY (Atom percent) Isotope Gamma Spectrometry Mass spectrometry Pu 238 0.0142 +/- 0.0004 0.0165 +/- 0.0002 Pu 239 93.52 +/- 0.06 93.49 +/- 0.01 Pu 240 6.066 +/- 0.007 6.024 +/- 0.006 Pu 241 0.38 +/- 0.04 0.444 + /- 0.002 Fu 242 0.021 +/~ 0.001* 0.0168 +/- 0.0003 * Value computed by isotopic correlation method JEL ANALXSIf QZ £LUJQNJIJM M GAMMA Afcffi TOTAL WEUTEQM CQIMIIMS (Plutonium in e/1) S.No Coulometry Gamma Total Neutron % Error Spectrometry Counting 1 0.43 0. 44 +2.3 2 1.09 1.11 + 1.8 3 1.98 2.01 + 1.5 4 2.92 2.84 -2. 7 5 5.28 5.34 + 1. 1 6 13.23 13.07 _ -1 .2 7 17.54 17.74 - + 1 .1 8 20.22 19.86 - -1 .8 9 8.70 - 8.63 -0.8 10 14.62 - 14 .76 + 1 .0 11 20. 13 — 19 .85 -1 .4 S.No. 1-5 Results with 60cc HPGe detector S.No. 6-8 Results with 0.5cc LEPS HPGe detector S.No. 9-11 Results with He3 detector IT - 20.10 BY TOTAL NEUTRON COUNTING TECHNIQUE (Plutonium in mg) Matrix Pu Taken Pu Estimated % Error PVC 150 145 -3.3 600 580 -3. .3 Cotton 150 135 -10.0 waste 600 610 -1. 7 Neoprene 150 135 -10.0 600 580 -3.3 Dowex 1X4 33.5 32.9 -1.8 216.7 218.8 + 1.0 TBP 101.9 102.0 +0.1 196.0 188.6 -3.8 IT - 20.11 A.Cylindrical Sample Storage Onit B.Gear System Controlled by Stepper Motor C.Sample Bottle Pushers D.Pushed Sample Bottle \J FIG. 1.SAMPLE STORAGE UNIT VITH PUSHER SYSTEM 1.Shielded Sample Bottle Cup 2.Robot 3.Sample Storage Unit 4.Opening for Sample Pusiier 5, Dt;cappt?r 6.Lt3cation for Sample Bottle Cap 7.Fxpetter 8.Location for Fresh Tips 9.Dilution Vial Rack 10.Location for Ejected Tips 11.Location for Reagents 12.Location for Vial Caps 13.Suction Needle for Sample D\sposal FIG.a.GENERAL LAYOUT D.c J.ABQT - 180 REAGENT Fl.OW SAMPLE ' INJECTION DETECTOR TUBE MAGNETIC BAR PINCH 4- DIGITAL DISPLAY OR HECOHOER (a)OPEN CUP SYSTEM REAGENT TLOW DISPOSABLE TIP ' SEPTUM DIGITAL DISPLAY OR RECORDER DETECTOR (b) CLOSED CUP SYSTEM WITH SEPTUM FROM HEAGLNT DOSlMAT PINCH VENT FROM SAMPLE DOSlMAT --» OIGITAL DISPLAY OR RECORDER PINCH DETECTOR (c) CLOSED CUP SYSTEM (WITH OUT SEPTUM) ON LINE FIG-3. SCHEMATIC DIAGRAM FOR FIA IT - 20.14 MANAGEMENT OF RADIOACTIVE WASTKS GENERATED DURING PLUTONIUM SEPARATION AND HANDLING OPERATIONS MS Kumra Head, Waste Management Division Bhabha Atomic Research Centre Bombay 400085. 1.INTRODUCTION The Indian nuclear power programme, envisaged from the beginning, adopted a closed nuclear fuel cycle. This involves separation Plutonium from the ipent fuel and recycling the same for nucles: power production, with a view to ultimately utilise vast reserve of thorium available in the country for nuclear power production. The recovery of plutonium from spent fuel, which is being don successfully for the past many years, along with other associate'* activities of fuel fabrication and relevant R&D programme- generate radioactive wastes which need to be managed in a techry • economically and safe mariner, These wastes characterised by tti>. decay heat generation and long half lives of the radionuclide^ present, account for more than 99% of the radioactivity in the entire nuclear fuel cycle. Though the wastes are produced in all the three physical states, major portion appears in the liquid streams. The strategy adopted in India and status of technology development in the management of these wastes is presented in this paper. 2.ORIGIN QE PLUTONIUM CONTAMINATED WASTES Reprocessing of spent fuel, plutonium fuel fabrication and research and development activities are the main present source;.; of plutonium contaminated waste. In future, decommissioning of reactors and plutonium handling facilities will also generate such wastes. The bulk of the plutonium contaminated wastes are generated during various fuel reprocessing steps given below: Disassembly operations: for removing external parts of th»; fuel assembly generates metallic waste Decladding operations'- for removing the fuel clad material gives rise to liquid waste' in case of chemical decladding and solid waste in the form of hulls during mechanical chop leach method. Fuel dissolution: for dissolving 1;he fuel in nitric acid prior to processing for plutonium recovery generates gaseous effluents containing radioactive noble gases, tritium, iodine etc. IT - 21.1 Solvent extraction: for separation of uranium and plutcniuu from the dissolved fuel solution and recovery of plutonium gives rise to aquous and organic liquid wastes containing fission products, transuranics including unseparated uranium and plutonium. These liquid wastes are of major concern. In addition to the above, solid and liquid radioactive wastes are generated in varying quantities and contamination levels from chemical laboratories, ventilation and off gas cleaning operations, decontamination and plant maintenance activities associated with fuel reprocessing. 3.MANAGEMENT QE WASTES As pointed out earlier fuel reprocessing ib the major source ol waste associated with plutonium. Further, the wastes generated during fuel reprocessing covers almost all types of wastes in gaseous, solid and liquid form. Management of these wastes is aimed at discharging radioactivity at as low as reasonably achievable but well below permissible levels. providing isolation from environment for desired periods of time ensuring adequate safety to the working personnel maintaining proper waste storage/disposal records, and providing desired surveillance. 3.1. Gaseous Wastes. The principal gaseous wastes arise during fuel shearing 31- dissolution operations as well as due to the ventilation an< process air systems operation Particulate tope .".er wit , entrained activity which includes Iodine-131, Iodi.ne-129 ana radioactive noble gases are the main contaminants. Treatment involves removal of particulate and entrained activity with the help of scz'ubers and filters. Thei decontaminated gaseous streams are released through high stack fitted with monitoring equipment to register activity level and flow rates. 3.2.Solid Wastes 3.2.1.Cladding hulls. The principal solid wastes are metallic discards and zircaloi- cladding hulls. Usually these wastes are moderately active Zircaloy hulls contain a very small portion of the undissolved IT - 21.2 spent fuel and constitute a significant radiological hazard. In addition to fission products, it contains long lived alpha emitters also. Utmost care is needed in handling as there is risk of fire hazard due to the pyrophoric nature of the material. The management practice consists mainly of their collection in steel drums by remote handling techniques and embedding in cement matrix. This helps in overcoming fire hazard due to friction and fixing the activity. T?ie waste conditioned in this manner i .s transported to the disposal site and confined by emplacing in underground circular vaults, known as tile holes. Schematic of a tile hole is shown in Fig.l. There is provision to retrieve the waste containers. ."! 2.2. Other Solids. A variety of materials are generated as low active solid waste from reprocessing operations and from research laboratories handling transuranics. These materials may be in the form of contaminated concrete, cellulosic material, glassware, contaminated metalic pieces and spent resins. Their management, involves monitoring and segregation into different typet- depending upon combustibility, compressibility and activity level. Combustible wastes are volume reduced by incineration. The system available for this purpose are free air incineration, controlled air incineration, fluidised-bed incineration, pyrolysis and pyrohydrolysis. High temperature slagging incinerators are also under development in some countries to handle all types of solids contaminated with alpha activity. Incineration units are equipped with sophisticated gas cleaning set up which includes cyclone separators, cloth filters, absolute filters etc. Non-combustible but compressible wastes are volume reduced by compressing in a baling unit, provided with elaborate ventilation system. Depending upon level of contamination, solid wastes are disposed directl*y after incorporation in cement.The disposal facilities include RCC trenches and tile holes located in controlled areas. Schematics of tile holes and RCC trenches are shown in Fig.l and Fig.2 respectively. 3.3. Liquid Wastes There are three major categories of liquid wastes from reprocessing: low level wastes such as fuel storage pond water, condensates from evaporators and many process water streams. Medium active wastes such as those arising from chemical decladding, TBP wash solutions etc. High active wastes containing bulk of the FPs are in the form of raffinates from separation processes. IT - 21.3 3.3.1. Low Active Liquids: These wastes are generated in very large quantities. The principal waste management practice adopted involves concentration and removal of radioactivity from bulk of the liquid and discharging the decontaminated waters into the environment on ensuring that their radioactive content is below the prescribed safe limits. Concentration is done by using various decontamination processes such as filtration, chemical treatment, ion exchange etc. Schematic of one such facility is shown in Fig.3. The waste concentrates from these processes are conditioned by their solidification in cement, bitumen, polymer itc. and finally disposed of in shallow underground engineered containment systems. 3.3.2. Medium Active Liquids: Medium active liquids with significant alpha acxivity are generated in research laboratories dealing with plutonium and other transuranic activities. Volumes are generally low. These waste solutions are treated by neutralisation-precipitation and the concentrates are solidified in cement. In order to avoid air contamination, treatment and conditioning systems are located in hoods or glove boxes with proper ventilation. Medium active liquids from reprocessing facilities are generated in much larger quantities. Their treatment involves evaporation and bituminisation. The waste is mixed with asphalt and bituminised in a wiped film evaporator. The product has high chemical durability with low leach rates. Schematic of the process is shown in Fig.4. J.3.3. High Level Liquid Wastes. High active liquid wastes (HLW) are self heating in nature besides being sources of intense nuclear radiations. These wastes are stored in high integrity stainless steel tanks located in underground concrete vaults with adequate provision for - onstant cooling. Such a facility is an interim arrangement prior to processing and ultimate disposal. Indian nprogramme for the management of high level wastes envisages a three stage strategy viz: (i) Immobilisation in solid matrix , (ii) Interim storage under surveillance for about 20 years, (iii) Disposal, in d«ep underground repository. Immobilisation. Vitrification process for immobilisation of HLW in borosilicate glass matrix has been developed and a plant has been commissioned utilising this process. A simplified vitrification process flowsheet is shown in Fig.5. The waste is volume-reduced by concentration in a thermosyphon evaporator. Glass forming agents IT - 21 .4 and the waste concentrates are fed to a metallic process pot heated in a multizone induction furnace. The waste gets first calcined and then converted into a vitreous mass, which is collected in stainless steel container. The entire process is carried out remotely in a concrete cell in a biologically shielded manner. Interim Storage. The vitrified waste generates significant quantities of heat and is kept under surveillance for about 20 years before final disposal. An underground near surface interim storage facility for vitrified wastes has been set up. This facility consists of an inner vault for housing the vitrified waste containers with an outer vault to take care of structural requirements. Self regulating passive air cooling joncept has been adopted to remove decay heat from the waste containers. Schematic of the facility is shown in Fig.6. Disposal in Deep Underground Repository. The final step in the management of HLW is ultimate disposal. The important aspect in this context is the associated long term hazard due to the presence of actinides in these wastes. A deep underground repository in a granite formation is envisaged in the country for disposal of verified waste containers. As of today such a deep repository has not been established anywhere in the world, though lot of efforts are being put in this direction by many countries. Qualitative acceptance criteria for deep disposal of such wastes have been developed by IAEA. Present work in this regard in India pertains to site selection and thermo-mechanical studies in a deep mine under inactive simulated heat load conditions. IT - 21.5 1500 DIA L 50 > 50 i 6 TH WITH HOLD FAST fRt;~:T R. C.C REMOVABLE COVER WATER PROONNG 5 mm THICK U S PLATE HUME STEEL PIPE 25 mm THK. SPUN CONCRETE 25 m if. MINIMUM CONCRETE LINING APPLIED UNDER PRESSURE 5 mm. THK M. S. PLATE CEMENT CONCRETE 12 0 (ai 150 c/c BOTH WAYS P. C. C. (1 : S 6 I ALL DIMENSIONS IN mm ) I30TH PC C-( i: 3 61 — '--PC C FILLING II MO) 230TH RUBBLE SOLING - « C C ««ft ( NECO NOT B£ PROVIDED IF W£ATH£RE0 ROCK IS MET WATER PROOFING AT-THAT LEVEL) I ALL OlMCNIIONt IN mm ) FIG. Z RCC Tr«nch Schtmailc IT - 21.6 CODE DESCRIPTION I PRE TREATMENT 2 SETTLER 3 VACUUM TRANSFER TANK 4 SLUDGE TANKS 5 SLUDGE METERING TANK 6 CEMENT HOPPER S CONVEYOR 7 SOLIDIFICATION UNIT 8 SOLIDIFIED SLUDGE DRUM 9 ION-EXCHANGE COLUMN I (0 SPENT ION-EXCHANGE DRUM II POST TREATMENT TANKS 12 SAND FILTER P PUMP ™^ 12 S SUMP DS -an 8/W M CHEMICAL MIXER J7 C CHEMICAL |N w- IN —» 0 DECANT V VACUUM DP CA COMPRESSED A«R S W SERVICE WATER TO SOLID WASTE DISPOSAL SlTEzrC^^^^ DS DEWATERED SLUDGE DP DISCHARGE POINT 2760 BW FEET FROM SHORE CF SACK WASH RIN CENTRIFUGE Fl FLOW INDICATOR FIG- 3 FLOWSHEET OF LIQUID WASTE TREATMENT PLANT TO STACK VOG r SYSTEM CONDENSER i V FEED ASPHALT STEAM CONDENSATE WIPED FILM FEED EVAPORATOR I THERMO-SYPHON INJ a EVAPORATOR t EXTERNAL 00 CONDENSER en DISCHARGE COOLER CONCENTRATE PRODUCT DRUM CONDENSATE FEED TANK STORAGE TANK EVAPORAT1ON-BITUMIN1SATION PROCESS SCHEMATIC FI6.4 . I IT - 21.9 o CM I To be given at International Symposium on Radiochemistry and Radiation Chemistry to commemo- rate 50 years of the discovery of plutonium. Speculative subtracer chemistry of plutonium in environmental conditions. Robert Guillaumont Radiochemistry group. Institute of Nuclear Physic, 91406 Orsay Cedex (France) Introduction Since several years, the chemical behaviour of radionuclides at subtracer scale level concentration (c.d when the number of species in which they are involved is less than hu.dred) has been a topic of interest for radiochemists [Gui86, Gui91, Pen90]. Indeed, at least in any radioactive filiation inevitable transient situations exist in which the last surviving parent atoms and the first nascent daughter atoms are present in denumerable quantities. Man-made chemical reactions: run with isotopes of transactinide elements belong to this field as well as some reactions with natural or artificial radionuclides occuring in, or over, the earth. The aim of this paper is to discuss the behaviour of Pu in environmenial-like situations This is an illustration of the concepts of subtracer chemistry because most of the nuclear and chemical properties of Pu isotopes fit well the criterias required for observable reactions. For radiochemists, chemical reactions can be classified according to three possible pathways between the chemical species considered as micro 01 macrocomponei'ts : pathway 1- micro-micro pathway 2- macro-macro pathway 3- micro-macro and two classes, according to the time, t, needed for chemical species to react : class A- t, ranging from hours to one year for man-made reactions or natural processes on earih, class B- t, ranging from several years to millions of years for geological processes. Reactions 1A are not observable in the laboratory because of kinetical hindrance. Rcaciio IB for which the reacting lime is no longer an obstacle to be achieved could be observed in th< environment. Exemples for 1A/1B reactions are Pu self-redox reactions. Exemples foi reaction 3A/3B, always kinetically allowed, are complexation of Pu aquo ions by macroamoums of natural complexing agents with respect lo Pu . We shall present and discuss, in several steps, the predicted behaviour of Pu quantities, down to few atoms,based on the situation known for usual concentrations. First, a short survey of Pu concentrations in environmental situations will be given. Then, we shall select the most probable equilibria of Pu species to be achieved, now or in a far future. Finally we will examine how these equilibria are modified in the case of few Pu atoms. This paper is neither a review of Pu in the environment nor a review of chemical properties of Pu in the environment, which already have provided a considerable amount of literature [Per80, TraSO, BonBO. Com8J, Wat83, AH84, Kim85, Pry86, Nit86, Mas87J. IT -22.1 / u concentrations mci in environment. Since aboul 40 years [Per80],.fallout have spread out over the world 238 to 241 Pu isotopes (238 and 240 abundance ratios with respect to 239 are 3.10 a;;ti O.ISi). and :htir ..<;;. ^'.-.•.c/s, like Am241 growing from the short lived Pu 241. Ground or underground nuclear tests gave rise to local Pu deposits as well as some local radwaste sources which increase the Pu amount into the envi- ronment [Cho88]. Since a long lime, Pu concentration has been measured as Pu239+240 ur/and Pu238 in different earth compartments. For Pu239+240, concentrations are expressed with the following correspondence : ldpm = 0.0166Bq = 450fCi, !fCi = 6.7 106 atoms = 1.11 10"17 mole. Chemical state of Pu in fallouts is the elementary state for ?/3 of total Pu atoms and the high fired oxide P11O2 stands for 1/3. To day, distribution of Pu, as well as its oxidation state, is quite known in most of the geosphere and biosphere systems and in living material [BonSO, Rai80, Mor86J. Despite of the fact that considerable amount of Pu has been relcazcd in environment (about 3.5 t in the atmosphere. 1 t in the ground, 0.1 t/yea: in Ihe sea water), steady concentrations of Pu, as soluble or dispersed panicles, is quite low and [pile i.onsimi i:i aqueous sysu.->:v Per if.-t* .ce the concentration is aboul 130 atoms/1 in the layers of the atmosphere close to the soil, between 0.1 to 0.5 fCi/1 in lakes water (and ^100 fCi/g in. sediments), about lfCi/1 in sea water land <5fCi/g in sediments) in sea water. In the 10 first centimeters of soil, Pu concentration is more than 5 fCi/g and, with a typical dislribution coefficient Kd in the range of 1(H cn^/g, concentration in ground water could be found less than 10-* atoms/cm-' at equilibrium. Then, more than 99% of Pu stay in the soil. Of course local concentrations can be higher, but the ntmibc of Pu atoms remains in the range of some hundreds. As a general (rend, Pu(lV) and Pu(Ill) oxidation stales are associated with solid phases while Pu(V) and Pu(VI) are in natural waters. Concentration in living material is less than 10"^ Bq/kg c.d. less ll'in 10-* aloms/g. Tlie repartition of these atoms is probably helerogenous. Finally, in rocks which contain uranium, production of Pu239 occuis through neutron reaction on U238 since a very loi^ time. The ratio i'u239/U238 is equal to 3 10'l2|A1184] which gives around 7.5 10^ atoms of Pu per gram of rock if the concentration of U is 1 ppm. Lixiviation of the locks could give highly diluted ground or underground aqueous solutions containing few Pu atoms, the volume of which depending on local conditions. Very small systems (10 10 10^ u^ up lo mm•*) with 'rapped Pu atoms could be found in rock fluid inclusions. All these figures lead to the evidence that environmental Pu chemistry is relevant to radiochemistry in most of the large aqueous sytems (lake, sea) where its concentration, as dissolved Pu is in the range of 10"''•*•' M. Bui, in some natural systems of restricted spatial extension, Pu chemistry is relevant lo subliacer scale level chemistry because the number of atoms 'n presence could be very low. Expected Pu reactions in environment According to numerous interdependent parameters on which depends the behaviour of Pu in natural waters or natural aqueous solution, and in addition the lack of both thermodynamic and kinetic relevant data, it seems unrealistic to try to modcii/.e it. All predictions can only be speculative, as it has been already pointed out, [AII80 and 84, Cho85 and 88, Kim85] and as the following ones are. We shall resiricl )\w discussion to whai could happen at a true tracer scale level of Pu in aqueous solutions for which Eli, pH and cornpicxing agent concentrations are comparable to those found in natural waiers (Eh(V) = (0.9 ± 0.1) - 0.06 pH for oxic conditions, Eh(V) = 0.21- 0.06 pH for anoxic conditions |Cho85, Nii87J,and presence of inorganic anions and organic materials, ionic strenght arour.i 10 , sec window in figure 1). These assumptions suppose no solid phase formation whipping 1 •,: by some coprecipitatioii mechanism, neither colloid formation nor sorpiion IT - 22.2 of Pu species on any solid material like colloidal polysillic acid, iron hydroxide or clay particles. Indeed such processes are not predictible by thermodynamics. In addition, we shall also suppose equilibrium fully completed. Of course, all these conditions are very far from those which really exist in the environmental aqueous solutions. In such aqueous solutions which are characterized by the macrocomponem properties, Pu species can "a priori" participate to acid base or redox type reactions with macrocomponents species. They are labelled 3A3B in the previous section and are of the of general form : Pu(N)j <===> Pu(N+X)2 X = 0, (acid-base)l, 2. (redox) (I) where N and N+X stand for oxidation state. For these processes the stoechiomctric coefficients of * i species are necessarily equal to one. When the right conditions are met, Pu can also undergo self-redox processes, labelled IB, such as : Pu(IV) + Pu(V)<===> Pu(III) + Pu(VI) (II) 2Pu(IV) <=====> Pu(iir> + Pu(V (III) 2Pu(V) <====> Pu(IV) + Pu(VI) (IV) or including the overall processes : (II) + (III) or (II) + (IV), the stoechiometric coefficient of which being not necessarily equal to one. These remarks will be of interest in later development. Each process described above is under the dependence of a formal equilibrium constant (see table 1). Which one of these reactions can be expected to occur in environmental like conditk ? and which Pu species are involved in ? are the questions we have now to try to answer. As it is well known and has been used many time Eh-pH diagrams drawn for different conditions give roughly the conditions of existence and coexistence of thermodynamically stable Pu species. One can also compare, for a given reaction (I) to (IV), the K values which are necessary to measure the equilibrium concentrations of the different Pu species takei: as a percentage of the total Pu concentration, with the estimated K values, as a function of pH for different values of inorganic and organic complexing agents concentrations. This route is used in the following developments only for IB reaction because they are of special interest as we shall see later. Needed K values for all the expected Pu reactions are summarized table 1. Complexation with inorganic anions Carbonate anions has been recognized as the main common complexing anions in natural waters for actinides. Its concentration in granitic water, which corresponds to logPcO2, is '" tne range -3.5 to -1 or more. Figure 1 shows Eh-pH diagram for the average value logPcO2 = "2- in *c r; .-. of occuring pH values, drawn with the set of data summarized in table 2. Most of these data ?.••• recent ones. All the values are corrected values to zero ionic strength with the help of ionic s* lie interaction theory. Some Pu data are extrapolated from adjacent actinides in the right o Jation stale. Most of these data are close to those found in literature survey [Kim85 and 90, Ai)80,82] and give probably a good idea of what can happen to Pu(III), Pu(V) and Pu(VI) aquo ions undergoing, at tracer scale concentration, hydrolysis and complexation with CO2 anions up to pH 9, although hydroxo-carbonate complex, if they exist, are not taken into account. It seems more difficult to know what really happen to Pu"*+ aquo ion in the same conditions because of the incertitude on the formation constant of Pu(OH>4 species which is of several order of magnitude. Table 3 summarizes the story of published data. To take into account hydrolysis characteristics of actinide M^ + ions, especially Pu^+» we have chosen -5 < log P4 < -9 at zero ionic strength. The highest value [Cap90c] is close to the log (54 values obtained for different M^+ actinide ions for which experimental data are more unquestionable than that of Pu^+ . Indeed Pu(IV) can be oxidized, or, if not, precipitation of Pu(OH)4 as well as Pu4+ disproportionation occurs when the pH of an acidic Pu^+ solution increase. These phenomenons hide the hydroliiic Pu'*+ reactions. The IT - 22.3 log 04= -5 value give a good fitting of Pu(OH)4 solubility and Pu4+ disproportionation IVit90j. Above pH 9, all the data are less reliable, but natural waters of pH ]0 are rather scaice The main consequence of our lack of knowledge about Pu(IV) hydrolysis is a -as it is known - considerate shift of the various predicted stability domains of Pu oxidation states [Cbo8&}, which, in addition are not so well defined as shown in figure 1, because af the inaccurracy of the others data. Figure 1 is drawn with •, P4 = -5- Figure 2 gives the variation of log K values for equilibria (II) to (IV) as a function of pH. The shadded aeras correspond u> log 04 = -5 or -9 for Pu(IV). As it was pointed out, log K = f(pH) have not been drawn for reaction (f). From comparison of the two figures, the possible existence of Pu(V), Pu(IV) and Pu(IIl) is confirmed for PCO2 = 10"2. as well as the possible coexistence of Pu(V)/Pu(IV) and Pu(IV)/Pu(IlI). They show also that only disproportionation of Pu(V) has some chance to occur above pH 7 (point D on figure 1 and light overlap on figure 2) and that, may be, could be measured. Increase of PS'OT •would lead to the same conclusions, and •>. de^reas^ v:> the .)>;><• :••'.& fines. These conclusions could be modified by the presence of another inorganic complexing anion : SO4^~, PC^-*-, F~, ci" if their concentration is high enough. In many cases, they are so low that no complexation occurs in the presence of carbonate anions in the pH rau^e of interest. Complexation with organic natural material Dissolved organic carbon can be found in significant amount in natural water as soluble monomeric species or polyelectrolyte substances such as humic substances as well as suspended solid matter [Cho85,88, Mor89]. Its influence on Pu behaviour has been recognized from a long time as very complicated and modeiization of Pu behaviour, on the basis of known data, is not yet possible. More data lire needed. Nevertheless, from the excellent analysis of Choppin and Allard [Chc>85], it is followed that only humic acids, which have carboxylic groups, have to be taken into account as complexing agents of Pu. These organic macromolecules could dominate solution chemistry of Pu(III), even in the presence of carbonate. For Pu(IV), hydrolysis is dominant over complexation by carbonate anions at pH < 9. Carbonate complexation of Pu(V) could be more significant than organic one. Competition between carbonate and organic complexation arise for Pu(VI). Finally, high*;; oxidation states VI and V can be reduced to lower ones because of the presence of some reducing functional groups on organic substances. On the basis of these qualitative trends, and using the representative data of table 4, tentative variations of log K values for reaction (II) to (IV) have been set up as a function of pH for the following typical conditions : concentration of. humic acid, HA, (with a theoretical capacity of 5 meq/g in COOH turctional group) equal to 1 mg/1. According to the data of table 4, the Pu species are always PuA+ and PuO2A + in the range of pH of interest if we assume that organic complexation dominates hydrolysis of PuO2^+ without taking into account -eduction of Pu(VI) by humic acids. Published data on redox reaction are relatively scarce [Cho85]. Figure 3 shows these variations. Overlap between estimated K values and the needed ones to observe equilibrium give indications on a possible simple disproportionation for both Pu(IV) and Pu(V) in the presence of humic acids in some pH range. Complexation by inorganic and organic anions If we assume that ihe behaviour of Pu is under the influence of all the factors taken into account up to now, the situation is more complicated. Figure 4 shows the variation of log K values for the same reaction.- as above, versus pH for HA concentration equal to 1 mg/1 and log PCO2 = "2- The conclusions are not very different from !hr previous one about disproportionation of Pu(IV) a"d PU(V)- IT - 22.4 Conclusion Though the reported variations of log K as a function of pH must be considered with caution, it seems chat complexation of Pu with humic acids and carbonate 9.lions could lead in some natural waters or some natural aqueous solution fitting the above considered compositions, to favorable thermodyriamic conditions for the development of self-redox reaction of Pu unless kinetic, of course, do not hinder them. We shall discuss now the kinetic aspects for the completion of these reactions, which are very important aspects. Completion of the expected Pu reactions As pointed out, reactions between microcomponents, which include at least one rate controlling step rate necesseraly of second order with respect of two species Ei and E2 of the microcomponents, need a long time to reach equilibrium. It, is difficult to appreciate this time because of the lack of relevant kinetic data which are incomparably less numerous than thermodynamic ones. Drastic extrapolation have to be made. Let us suppose a controlling step rat'- of second order between E] and E2 with kinelic constant k (Ml"'s"'). Whatever are the initial Ej and E2 concentration in ihe range of meiallic element concentration, C(M), the half-time of the raction T(S) is more or less given by : T = 1/k C If the number, N, of atoms of the microcomponent is contained in a volume V (I) it follows : lud X = 23.7+log V-log N-log k The fastest reactions present k = 101 1 which gives for N = 1()2 and V in the range I 11 ^ to 1 cm-' : 1 min< t Decrease of k value leads to drastic higher values of x . II we suppose t = 10 years, k values range from 1.5 10^ to 1.5 IO" within the limits of the previous V values. They are still high values for rate constant of reactions More is the spatial extension of the system, where the lew atoms uic iruppcd, more the hindrance of the reaction disappears. For Pu most of the data about the kinetic of redox or self-redox, reactions are related to aquo ions |Cle7(), Tol90, AhrS6, Cho83]. Mechanisms of reactions hav? been shown to lie very complicated. The reaction rate depends, at least, on pi), break or formation nl 1'u oxygen bond, com- plexation of aquo ions and initial composition. It increases with |Hf| when Pu-O are broken and increases with the decrease of |H+| when Pu-O bond are formed. Complcxaiion seems to inuease it. The rale of reaction hetwen the four Pu oxidation stales is last. In acidic solution, it seems pH independent. Disproportionate of Pu^+ (2 10"-' M) in MCI 0.5 M lakes 10 days to he completed and only 20 hours in HN()i0.3 M (8 10"^ M) and the rale varies as |H'| '. It could lie ox peeled that this reaction will he fast at pH > 5. Increase of tcntpciatuie lias a very important el feet. Disproportionaiion of PuO'^ + has l>ccn reported 10 be fast at pH < 1.7 ami pi I > 7. but takes tew days at pH 3.5. All these data give indications on the possible completion, ovci geological peiioil ol lime of the previous selected, Pu equilibria especially in closed natural systems wb iii small spatial extension. IT - 22.5 Expected behaviour of few Pu atoms Whe have shown [Pen90, Gui90, Gui91] thai, at the level of very few atoms trapped in closed system, the law of mass action, expressed in terms of the number of reacting species. c.\, is inadequate for describing equilibrium state of the system. One must use the average number of the species, Ngi , calculated by using the canonic probability for the realization of the microstates of the system, P, set up in terms of the exact, but limited, Ng, values : NEi = The summation goes over all the possible discrete values of Ngj. This approach, different from the classical one, has be developed because in such situations Stirling approximation and the concept of chemical potential of classical statistical thermodynamic do not apply. Of course when the N£j values increase, equilibrium constant K computed for the average populations '•: omes equal to K, the classical one, and therefore the limiting value of the ratio : p= K/K is equal to one For al< reactions between rnicro-macrocomponents such as reaction (1) for which ail stoechiometric coefficients of the species of a single microcomponent are equal to one, it has been shown that p = 1, whatever the number of species are. So the behaviour of some Pu atoms with respect to complexation is, in average, the same as at tracer scale level and that is the firs* conclusion we can point out. For reactions between micro-microcomponents, p is a function of the number of reacting species, which must be established for each case. Lei us consider two silnations : first, the increase of the number of Pu atoms in a given oxidation state assuming that they participate tj equilibria (III) or (IV), which will be achieved, may be in a far future, and secondly, the last stepi of the life of Pu atoms before their extinction assuming that, at that time, equilibria (III) and (IV; will have been fully completed. Growing of the number of Pu atoms Figures 5 show the variations of p= K/K for reactions (III) and (IV) seen «a or disproportionate reactions, versus the number of Npu(|V) Npu(V) speciesi for typical selected K values and simple initial conditions with respect of the initial number of species. When K « i and when NpQ(iv) or Npu(V) are less than about one hundred, Pu(!V) or Pu(V) disproporiionatton would not be as important as it should be at tracer scale concentration or for higher concentrations in situation of equilibrium achievement. When K = I or K >> 1, the situation is more complex because the disproportionation of Npu(iv) or Npu(y) < 100 can be more or less important than it would be at higher concentration. It depends on the Pu atom number, which can be even (2 modulo 2) or odd (3 modulo 2). If Npu(]\r) or Npu(V) fluctuates in the range of few atoms to some hundred, the behaviour of the system appears chaotic. Figures 6 show the variations of p = K/K for reactions (III) and (IV) seen as remutation reactions, versus the number of NpU(ju) or Npu(iv) species and for typical K values and different initial conditions with respect to initial numbers of species. If K < 1, remutation of Pu (III) and Pu(V) or Pu(lV) and Pu(VI) would be higher for few atoms of Pu than at tracer scale. It would be (ht reverse for K > 1. indeed in each situation, p reaches quickly the value p = 1 and for the chosen ini tial conditions, no 'onger chaotic behaviour is then expected because of the stocchiomctry of ih rCaCl'°n- IT - 22.6 Let us now to point out how much the behaviour of Pu is sensitive 10 the. initial number of reacting species. Death of Pu atoms If we assume that equilibrium (III) or (IV) has been completed, for a large number of Pu atoms, N, the decay of Pu towards extinction will lead to subtracer scale situations when the number of Pu atoms will decrease to some hundreds. As N decreases slowly, we can suppose that the equilibrium is continuously restored. In this case the initial numbers of reacting species are at each stage those of the previous equilibrium, which are related to the successive K and N val^- For instance for reaction (IV) they are : NPu(V) = r—. Npu(iv) = Npu(vi) = NPU(V)VKT The variation of p as a function of the stoechiometric allowed denumerabie N values eru b-iween those shown in figures 5a,5b for different K values. If the equilibrium corresponds, for instance, to K = 1, K values fall down to about 10"' as N values decrease to the smallest allowed r.es, N being even or odd. So, variations of p are, first, close to those corresponding to K = 1, and •jily, close to those corresponding to K = 10*-' (figure 5b). These considerations apply to •jaction (III) with appropriate notations. The last question we have to ask ourselves is about the possibility of measuring completed ' quilibria in a given system. The relationship between the half-life, T(s), of a radionuclide and the number of atoms, N, accessary to have Idpin, is : T = 85.7 N which gives N = 7.56 109 atoms for a emission of Pu239+240 or N = 3 107 atoms for Pu238. Detection of 100 atoms of fu by a counting is out of the limit of the best solid detectors used in the best conditions. Tentative to detect 1 a per day above the background in the 5.5 MeV energy range would not change the situation. So, the reality of the Pu equilibria in a large closed system has good chances to never be confirmed despite of the fact that chemical methods of separation of Pu oxidation stales are known [Kab88, Geh86, Cho85], But it could be certainly proved by analysing a great number of very small closed systems. Cor il •„. L :ie present developments have no practical importance in the environmental Pu chemistry and iisem to lie on the border of fiction. It is why the title of this paper refers to speculative chemistry of Pu. Nevertheless, such reflexions on Pu chemistry give the opportunity to rise questions about the behaviour of highly diluted matter. If the problem can be easily handled from a tbeort'Val point of view, it is difficult to confirm experimentally some of the theoretical uor.cluiions concerning IB reactions because of the necessity to meet contradictory conditions : the artificial radionuclide must have numerous oxidation states, and a relatively short half-lite allowing counting of few atoms but giving long time enough to achieve the completion of the equilibrium. Unfortunately, Pu isotopes fit well onlv two of the conditions over the three needed. Ac knoweldg merits. The author thanks P. Vitorge, H. Capdevila and V. Moulin for helpfull discussions.and providing him with up dated data. IT - 22.7 References - Actinides in the environment- Session of " Actinides 89 - international Conference", Tashkent, USSR . Under publication in Journal of Radioanalytical and Nuclear Chemistry,!990. - Ahrland, S.- Solution chemistry and kinetic of ionic reactions - in " The chemistry of actinides elements ". Second edition. Ed. J.J. 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Choppin, G.R. and Morse, J.W.- A study of technics for separating Pu in different oxidation states - Radiochimica Acta 43/4 (1988) 211-215. - Kim, J.I.- Chemical behaviour of transuranic elements in natural aquatic systems - in " Handbook on the physics and chemistry of the actinides ". Ed A.J. Freeman and C. Keller, Elsevier Sc., Vol 4, 1985, p.413-455, and references included. • Kim, J.I.- Geochemistry of actinides and fission products in natural aquifer systems - CEC Project Mirage. CEC Luxembourg EUR 12858 EN-1990, p 1-97. temire, R.J. and Tremaiiw, P.R.- U and Pu equilibria in aqueous solutions to 200°C - J. Chem. •*;ig. Data. 25 (1980) 361-370. "" ierse, Ch. - Thesis - Technical University. Munchen, 1985. Mahajan G.R., Rao, V.K. and Natarajan.- Interaction of hutnic acid with Pu (PI)- Radiochem. ' iMasanobu Sakanoue - Transuranic nuclides in environment-Radiochimica Acta 42/3 (1987) 103- 111. Metivier, H. and Guillaumont, R.- Hydrolysis and complexing of tetravalent plutonium - Radiochem. Radioanal. Letters 10/3 (1972) 27-35 and J. Inorg. Nucl.Chem. Supplement, 1976, p. 176. - Morse, J.W. and Choppin, G.R. - Laboratory studies of plutonium in marine sediments - Marine, chemistry, 20 (1986) 73-89. and references included. - Moulin, V. - Personal communication of unpublished results. - Nitsche, H. and Edelstein, N.M. - Solubilities and speciation of selected transuranium ions - Radiochimica Acta 39/23 (1985) 23-33. - Nitsche, H.- Effect of temperature on the solubility and speciation of selected actinides in near natural solutions - Inorganica Chimica Act? 127 (1987) 121-128. - Newton, T.W. and Sullivan, J.C.- Actinides carbonate complexes in solution - in " Handbook on the physics and chemistry of the actinides". Ed A.J. Freeman and C. Keller. Elsevier Sc., Vol 3, 1985, p.387-406. - Peneloux, A. and Guillaumont, R - Solutions de dilution extreme et loi d'action de masse - Comptes-Rendus Acaddmie des Sciences, serie II 310, (1990)1607. - Perkins, R. W. and Thomas, C. W.-Worldwide fallout- in" Transuranic elemenis in the environment ". Ed. W.C. Hanson, published by Technical Information Center U S Department of Energy.1980, p.53. - Pryke, D.C. and Rees, J.H.- Understanding the behaviour of actinides under disposal conditions - Radiochimica Ada 40/1 (1987) 27-32. - Rai, D. Seme, R.J. and Swan son, S.L.- Solution species of Pu in the environment - J. Environ. Qua I. 9 (1980) 417-420. IT - 22.9 a) - Riglet, C. Robouch, P. and Vitoige, P.- Standard potential of MO22+/M(>2+ and M4+/M3f redox system for Np and Pu species - Radiociiimica Acta 46/2 (1989) 85-94. b) - Riglet, C. - Thesis - University Pierre et Marie Curie. Paris, 1989. - Riglet, C.and Vitorge, P.- Standard potential of MO22+/MO2+ of U and others actinides - Inorganics Chimica Acta 133 (1987) 323-329. - Robouch, P.- Thesis, University Louis Pasteur. Strasbourg, 1987. - Smith, R.M.and Martell, A.E.- Critical stability constant - Vol 4 : Inorganic complexes. Plenum press, 1976. - Torres, R.A. and Choppin, G.R.- Eu and Am stability constants with huir.ic acid - Radioch imica Acta 35 (1984) 143-148. - Toth, L.M Bell, J.T. and Friedman, H.A.- The disproportionate of Pu4+ in nitric acid solution - Radiochimica Acta 49/4 (1990) 193199. - Transuranic elements in the environment.- Ed. W.C. Hanson, published by Technical Information Center U S Departement of Energy, 1980 (700 pages). - Vitorge, P.- Personal communication - 1990. - Waiters, R.L., Hakonson.T.E., and Lane.L.J. - The behaviour of actinides in the environment - Radiochimica acta 32 (1983) 89-103. - Weigel, F. Katz, J.J. and Seaborg, G.T.- Plutonium- in " The chemistry of actinides elements " Second edition- Ed. J.J. Katz, G.T.Seaborg and L.R. Morss. Chapman and Hall. 1986, p.788 and references included. Reaction logK * (see text) I - 1.3 (a 1.3 (b II 1.90 (c - 1.90 (d III and IV 1.95 (c -3.16 (d II+III and H+IV 3. (c -4.62 (d for instance K ni = [Pu(III) [Pu(V)]/[Pu(IV)]2. a) value for [Pu(N+X)] equal lo 5 10'2 Cpu, Cpu total Pu concentration 2 b) value for [Pu(N)] equal to 5 lO- Cpu 2 c) values for [Pu(IV)] or [Pu(V)] equal to 5 lO- CPu, d) values for [Pu(lV)] or [Pu(V)] equal to 0.45 Cpu (reaction II) or 0.95 Cpu (other reaction) Table 1 - K values needed to observe possible Pu equilibria IT - i:.-.10 Pu E(N+1/N) Hydrolysis (c Carbonate complexation aquo ion mV log *Pi log *P2 log *P3 log *jl4 log Pt log P2 log p-j Pu3 + -6.6 -14 -25.8 7.7 (d 11.9 (d 13.3 (d 1.015 (a Pu4 + -5(to-9) 0.996 (b PuO^+ -9.6 4.7 (e 6.2 (e 5.3 (e 0.954 (a PuO22+ -5.2 -12 -20 9.3 (f 14.2 (f 17.4 (f N n + + *pn for M + 11H2O <=====> M(OH)n( ' ) + nH 2 N 2 + Pn for M+ nCO3 - <=-=> M(CO3)n( - ) a) Rig87,89a ; b) Cap90a ; c; Cap90b eKcept for the lower value of log *P4 for Pu4 + see table 3 and text : d) Rob37. Value for Am3+ ; e) Rig89b. Value for NpC)2+ ; 0 Rig89b K for CO2g + H2O <===> CO32- + 2 H+ equal to -18.15 4 + log P5 = 40 for Pu Table 2 - Set of thermodynamic constants corrected to zero ionic strength. log 34 References and comments Pu 1 -10.80 Mct72. Recalculated fiom expcriniemal data. Pu 1 -12.7 Kim85. Cited as a result of I.ie85 Pu 0 -9.5±0.2 Bae76. Calculated from daia of Mci72 Pu 0 -9.5±1.2 A1I80. Selected from Bae76 and Smi76 Pu 0 -9±4 LemKO. Calculated taking into account Bac76 value and oilier data Pu 0 -9.5 AIIK2.K4. Selected from Bae76, M<-t72 and RaiXO u 0 -5.13 BruSy. SkB data base tiled in Kig8'.>e as personal communication U 0 -10.3 Bac76 Th 3 -1 1 Rob87. Cited as a personal coinnuinicalion from Grenthe Th 0 -5.5 Ely 90 Th 0 -15.9 or Bac76 -17.4 transformed. if necessary with log Kw - -14. Pu data arc also quolcd in WciH6. Table 3 - Hydro1 vsis formation constanl of M(()ii)4 species for some aclinide ions IT - 22.11 Pu (III) Pu(VI) v pH a (a pKa (b log Pi (c 1 log 02 (c log Pi (d g P2 (d 5 0.6 4.8 10 13.58 7.88 11.2 6 0.8 5.4 12.22 14.64 8.84 12.1 7 0.92 5.9 13.85 15.43 9.56 12.75 e) 8 0.97 6.5 1 4 15.54 9.65 12.86 e) 9 1 6.5 14.4 15.7 9.8 13 a) degree of dissociation of humic acid, b) pKa= pH-log ct/(l-a). Tor84. c) deduced from Am(IIJ). log Pi = 3.5+10.9u, log 02 = 10.4+5.3a. Cho85. d) deduced from U(V1). log Pi = 5.0+4.8a. log P2 = 8 5+-1.5a. Cho8v e) for pH > 7 all the data are cuijdptiiu.cd iiom i-Apciiniciiiai data replied L>y Cho85. f) considering that no reduction by humic acid occurs. Other typical values are: PuO2+. log Pi ~ 2 10 3 (by analogy with NpC>2+) Mou90; Pu4+, log Pi = 9.8 + 9a, log P2 = '5 + 9a. Cho85; Pu3\ log Pi = 3.1!. Mah89. Other typical values for actinides appear in Kim90 Table 4 - Typical formation constant for typical humic acid IT - 22.12 Figure captions Figure 1 - Eh - pH diagram for Pu species, log PCO2 = '2- ^nwn wiln dala oi toble 2- Disproportionation reactions occur around points A, B, C and D. A-3 Pu(IV)<««o2 Pu(DI) + Pu(VI); B-3 lP>KV)<-«>Pu(ni) + 2 Pu(VI); C-2Pu{IV)<=c>Pu(ni) + Pu(V); D-2 Pu(V>e™>Pu(IV) + Pu(VI). The coexistence of the four oxidation states occurs along and around AB segment. The window encloses Eh-pH conditions of most of the natural waters;, shadded areas account for uncertain borders. Pu species are: 1) PuO22+ ; 2) PuO2OH+; 3) PUO2CO3 ; 4) PuO2(CO3)22-; 5) PuO2(CO3)34"; 6) Pu4+ ; + 4 3+ 7) PuO2 ; 8) PUO2CO3-; 9) Pu(OH)4 : 10) Pu(CO3)4 -; 11) Pu ; 12) PuCO3+ ; 13) Pu(CO3)2-; 14) Pu(CO3)33". Figure 2 - Pu compilation with carbonate anions, log PCO2 = "2- lo£ K f°r reactions (II) to (IV) as a function of pH, drawn with data of table 2. Reaction (1!): line 1; (III): line 2 and (IV): line 3. Lines labelled : a. are for log *p*4 = -5 and lines labelled : b, are for log *p\} = -9 (see text). Domains of needed logK values t<> observe reactions are limited by lines 4 and 5 according to table 1. Figure 3 - Pu complexation with 1-mg/l of typical purified,huniic acid. Same legend as figure 2. Figure 4 - Pu complexation with carbonate anions. log PCO2 = -2 ancl ' mffl of typical purified humic acid. Same legend as figure 2. Figure 5.a. 5b - Npu(rri) pu(v) j variation of p = A JT (reaction II) or p = Npu(V) j^- (reaction III) as a function of N, the initial number of Pu(IV) or Pu(V) species for different K values. Initial concentrations are: Npu(||])0 = Npu(v)0 = 0, (reaction II) or Npu(iv)0 * Npu(VI)0 * 0, (reactioa III). N = 2,3,4, p values are given fo. the following sets of N values : N » 2 modulo 2 (upper branch) and N == 3 modulo 2 (lower branch ) unless K < 1. Figure 6.a. 6h - pUfJVr pu(V) Variation of p » _ _- K (reaction 10 or p - — K Np(m) Npu(V) N N (reaction III) as a function of N, the initial number of Pu(III) or Pu(IV) species for different K values and different initial conditions. a) N - Npu(Hi)0 • Npu(V)0. (reaction II) or N = NpuQyjO • Npu(vi)0. (reaction III). N=1.2,3,4 .... b) N - Npu(in)0. Npu(v)o » 50 and 100, (reaction II) or N » Npu(iv)0. Npu(vi)0 -50 and 100. {reaction III). N • 1,2,3,4. IT - 22.13 i 2 3 4 5 6 _ _ _ _ in 20 A D 7 mT~ -^^^ MIDlnTTri rrrm —' 10 - c ^\ 0 i 10 LUUDJUI fan -10 - 11 12 13 14 J l_ 1 I 0 12 3 4 5 6 7 8 9 10 FIG-1 pH PH IT - 22.14 an o -10 2> III — PuA — IV — Pu(0H)+- v —PuOJ — VI — PuO2A*- 6 7 8 9 FIG-3 pH at o 1a -10 -20 III IV PulOHJi V Pu Of H Pu02C0a- VI •PuO2A* u PuO2ICO3)- 6 7 FIG-A IT - 22.15 10 20 30 40 ., 50 80 100 N c 20 -10 60 80 100 '._„_*_! - ..J .1 1 ^_^_.l..—J 50 100 150 200 250 300 FIG-5b N IT - 22.16 10V100 o 40 to go .. too 10V50 10 20 30 40 .. SO 102, Si. J ,M> 4U 60 to too N 10-2/100 10, - 1 , . 1 10 JO 4(1 60 I 10'2/50 .... i ...... I ., 10 20 30 40 N 10-2 . I ...... 20 40 60 HO 100 N IT - 22.17 ENVIRONMENTAL ASPECTS OF ACTINIDES Pillai K.C. IT - 23. THE BIOLOGICAL ASPECTS OF PLUTONIUM Narayani P. Singh and McDonald E. Wrenn Environmental Radiation & Toxicology Laboratory University of Utah, School of Medicine 956 West LeVoy Drive, Salt Lake City, Utah 84123 U.S.A. ABSTRACT The biological aspects of plutonium including the metabolic behavior in man and animals, the biochemical behavior, and the toxicological aspects are reviewed. The results obtained from the human and animal studies suggest that the majority of plutonium is accumulated in the liver and the skeleton. However, if the exposure is through inhalation and the plutonium inhaled is highly insoluble PuC>2, a large fraction of Pu remains in the lung for a long time. The toxicity studies suggest that the alpha emitting isotopes of Pu produce bone sarcomas, liver cancers (both carcinomas and sarcomas) and peripheral lung tumors. The results also show that 239pu is 16 to 19 times more efficient in inducing bone sarcomas than is226Ra- Beagle studies show no increased hematopoietic neoplasea (leukemia). (Key Words: Plutonium, Metabolism, Biochemical, Toxicological, Sarcoma, Carcinoma) IT - 24.'I INTRODUCTION The story of plutonium is one of the most dramatic in the history of science. Ed McMillan, the discoverer of eiement number 93 (Neptunium) started looking for element 94, but before he could complete his investigation, he was called away to perform urgent wartime work on radar at MIT. On December 14, 1940, Glenn T. Seaborg's plans (with Ed McMillan's concurrence) were implemented in the search of element 94: "deuteron bombardment of Uranium, using the 60 inch cycletron". Element 94 was born at last on the night of February 23-24, 1941, in a lab known as Room 307, Gilman Hall, University of California, Berkeley. In his article "Plutonium Revisited" Seaborg (3) mentioned the following: "The potential toxicity of Pu as a bone seeker is well known, as is its high radioactivity, amounting to about 140,000,000 alpha-disintegrations per minute per milligram of Pu- 239." Mention these two properties to the radiobiologist, and his face turns pale yellow like the color of plutonium in the +6 oxidation state; mention the enormous sums of money spent world-wide on it's research development and production to the economist, and his face turns green, the color of the hydroxide of the +4 oxidation state of Pu; all color drains from the sociologist's face (and the politician's face) when he must consider the portentous military and political power of weapons using plutonium (Pu in an oxidation +5 state is colorless). Seaborg, in his own statement, at a very early ti.-rse mentioned that "...the physiological hazards oi working with Plutonium and it's compounds may be very great. Due to its alpha radiation and long life it may be that the permanent location in the body of even very small amounts, say one milligram or less, may be very harmful. The ingestion of such extraordinarily small amounts as some few tens of micrograms might be unpleasant. In addition to helping set up handling measures, so as to prevent the occurrence of such accidents, I would like to suggest that a program to trace the course of plutonium in the body be introduced as soon as possible. In my opinion such a program should have the very highest priority." The present review paper describes the biological aspects of plutonium, including the metabolic behavior in animals and humans, the biochemical behavior, and the toxicologicai aspects. IT - 24.2 Metabolic Studies Early works were done at Berkeley (University of California), where plutonium was discovered; at the Metallurgical Laboratory (University of Chicago), where the separations process for the industrial scale production of plutonium was developed; at Los Alamos National Laboratory, where fabrication of plutonium for weaponry occurred; and at the Hanford Works where the large scale separation of plutonium was done. On October 15, 1942, a program under the direction of Joseph G. Hamilton was initiated to determine the metabolic properties of plutonium, other actinide elements and Mission products. Hamilton pointed out that prior to World War II, the world wide total of only about 1 kilogram of radium had been separated from uranium ores, and yet radium poisoning had become a serious problem in the luminous dial industry (2). Thus plans to produce plutonium in kilogram quantities, accompanied unavoidably by production of mega-curie amounts of fission products, presented an awesome increase in man's total experience with radioactive materials. From c;irly studies it became evident that the distribution pattern of plutonium in the body was affected by its chemical form, the route of its administration, the species of animal exposed, and its age at exposure. The second and highly important aspect was the comparison of the metabolism of plutonium in man and experimental animals. This research was initiated April 10, 1945, when the first of twelve terminal patients was injected as a joint effort of the Los Alamos National Laboratory and the Atomic Energy Project of the University of Rochester, School of Medicine and Dentistry (3). Three patients were studied by the Chicago group and one by the Berkeley group. It was determined that in m;sn, as in experimental animals.the skeleton and the liver were the principal deposition sites for plutonium given intravenously as a ^39]^ jnc jtrate complex. Poth skeletal and hepatic retention decreased slowly in man, while plutonium in the rodent liver decreased more rapidly. Careful analysis of urine and feces revealed that less than 10% of the Pu was excreted during the first five years after injection. On July 1, 1950, the University of Utah started the first contract with the U.S. Atomic Energy Commission to initiate the metabolic and toxicity studies of plutonium, other actinide elements and fission products, using beagles as the experimental animal. The primary route of intake of Pu in these dogs was by injection, whereas the studies performed at the In. 'lation Toxicology Research Institute, Albuquerque, New Mexico and Battelle Pacific National Laboratories, Richland, Washington were through inhalation, also using beagles as the experimental animals. In August, 1968, the United States Transuranium Registry (USTR) was established to study the behavior of transuranic elements in humans. The distribution of Pu in human tissues of the general population IT - 24.3 exposed to fallout plutonium has been investigated in a number cf laboratories of various countries such as the United States, the United Kingdom, the Federal Republic of Germany, Japan, etc. (4, 5). The animal studies performed at the University of Utah, Padiobiology Laboratory (6) reveal that the target organs for plutonium in beagles receiving a single intravenous injection of 239pu are the liver and skeleton irrespective of their age group (2 days, neonate; 3 months, juveniles; 12-19 months, young adults). The relative size of the organ or its functional entities at the time of exposure, the state of organ development, and the rate of growth and tissue renewal plays an important role in controlling the fraction of injected plutonium and anatomical distribution wilhin each organ. The immalure liver deposited a smaller fraction of plutonium than the liver of the young adult or mature dogs. A review of the animal works suggest that in dogs injected with plutonium, the organ distribution patterns are affected by the mass of injtcted plutonium and the chemical form of the plutonium injected. However, for inhaled plutonium there is a larger number of variables such as aerosol particle size, particle specific activity, chemical form of the particle, mixed elements effects, and lung burden. The comparative distribution of inhaled 238PuO2 and 239PuC>2 in beagles was studied by Park et al. (7). The whole body retention half-time of Pu was similar for both 2 isotopes, about 12 yefirs. The 39pu was retained in the lungs with a mean retention half- time of 3 years and slowly translocated to the tracheobronchial lymph nodes. In comparison, 238Pu was retained in the lungs with a mean retention half-time of 1 year. The early human studies (1945-1946) involving 18 persons (15 over the age of 45), who were injected with tracer doses of plutonium, revealed that a few days after injection, human soft tissues (other than blood and liver) contained as much as 20% of the Pu dose; 5 to 15 months after injection the average liver Pu content was 31% of the dose for three cases with presumably normal liver function; and four to 457 days after injection the mean total skeleton Pu was 49% for the seven cases judged to have most nearly normal liver and skeleton (7). Two of these patients were hospitalized in a metabolic ward in 1973. All urine and feces were collected and analyzed for Pu by alpha-spectrometric technique and the results were utilized by Rundo in determining the late urinary and fecal excretions of Pu in persons exposed to Pu. The results suggest that at 10,000 days after exposure, the urinary excretions were 2.52 x 10-''% and 1.41 x 10~5%» of injected doses and ihe fecal excretions were 1.05 x 10^% and 0.53 x 10-5% of the injected doses (9). The concentration and metabolic behavior of Pu in the general population was determined by a number of investigators throughout the world after the fallout of Pu from open air testing of nuclear weapons. Thesi results also suggested thai the majority of IT - 24.4 plutonium is in the liver and the skeleton. The liver burden of Pu was lowest in the Finnish population at 12.2 mBq and highest in Salt Lake City (Utah) population at 54 mBq. The average liver burden of Pu for all the populations was 37± 10 mBq. The skeletal burden of Pu was lowest in the Finnish population at ~15 mBq and highest in the Tokyo population ;it ~207 mBq. The ratios of the skeletal burden to liver burden in these populations ringed from 0.7 to 5.9 whereas International Commission on Radiological Protection (ICRP) has clearly stated that the plutonium retained in the skeleton and the liver are equal, each containing 45% of the total systemic body burden. Mclnroy et al. (10) studied the 2 distribution of 39pu a;,d 238pu jn occupationally exposed persons. In five cases the exposure was through inhalation and in one case via a wound which had occurred 20 years or more prior to death. The distribution of 239pu fa four of the inhalation cases suggested that 45% of the total body burden was in the lung even af*er 20 years or more post exposure with only 26% in the liver, 32% in the skeleton and 7% in all other tissues. The wound exposed person had only 1% of the retained Pu in the lung. The systemic body burden of Pu in 5 cases (excluding respiratory tract) was as follows: liver, 35.4 ± 12.5%; skeleton, 53.7 ± 12.5%; striated muscle, 6.5 ± 1.8%; and all other organs and tissues, 4.4 ± 1.7%. The 2-^Pu distribution in the whole body of a person exposed previously through inhalation was 9% in the lung, 52% in the liver and 35% in the skeleton. These results of animal and human experiments suggest that high fired PuO2 remain in the lung for a long time after exposure, the other major organs which accumulate Pu are the liver and the skeleton. Biochemical Behavior of Plutonium Once plutonium is translocated to the blood stream either from the lung (inhalation exposure) or puncture wound, the circulating plutonium is mainly bound to the iron carrier protein Lransferriii (TO, which is a strong complex relative to metals other than iron. It is assumed that circulatory transport of Pu, delivery of Pu to cell surfaces, intracellular Pu transport, and probably release of Pu from cells as well, are mediated by iron-carrier proteins. H is important to note that many extracellular Tf-like proteins exist, differing slightly in protein structure but with similar iron-binding sites, such as Sero-Tf plasma, Lacio-Tf in milk, Ova-Tf in egg yolk. Other Tf exists in amniotic fluid, cerebrospinal fluid, lymph, colostrum and bile. Cell membrane receptors for Tf are found or arc assumed to exist on all iron dependent cells, for example reticulocytes, hepatocytes, ftbroblasts, placental cells, macrophaytos, renal epithelium and tumor cells. Plutonium circulating in the blood stream is mostly accumulated by the skeleton and the liver. In bone, plutonium accumulates at the organic-mineral interlace (MBS) of the IT - 24.5 bone surfaces. Plutonium deposition is greater on the well-vascularized surfaces of trabeocular bone that are surrounded by active orythropoietic marrow (RM). In RM, transferrin moves freely in and ot't of the blood senusoids carrying iron to developing red cells, and in those locations Pu-Tf has a greater probability of contacting or closely approaching MBS. It is interesting to note that in the same kinds of MBS, Pu deportation is greatest on exposed resorbing surfaces, and least on growing surfaces covered by osteoid and proliferating cells. Plutonium accumulates on calcified non-skeletal structures, eg. renal calculli, calcified tracheal and laryngeal cartilages. Toxicological Behavior of Pu From the animal work it is clear ihal line ilpha euiHuag 'wncpcs o} pluiortiuii* produce boat, aaiiAjuias, li'.x. i_ar.«;tj'i (bi«:'-j. rt: r-iAi aint YJ.U •,nn..jj ;;id •••ei"i|jheii«! long tumors. The plutonium induced pulmonary cancers are essentially all carcinomas arriving from alveoler and small airway epithlium. 'Hie bone cancers are mostly osteogenic sarcomas, a majority of which originate in trabecular bones having a high percentage of red marrow and a high turnover rate, such as exists in the vertebrae and proxin ia! femur and humerus. The liver cancers, including both carcinomas and sarcomas, occur at late times after exposure, presumably because of the prolonged lifespan and low cell turnover of hepatocytes. Pathological changes such as prolonged liver enzyme changes, have been noted after relatively low Pu doses to the liver. Pulmonary functional changes have been noted many years after initial exposure (i 1). In dogs injected with tetravalem plutonium (single intravenous injection) some of the less serious end points were pathological fractures, dental changes, aad atrophy of the turbinates. These latter conditions produced functional impairments in only part of the dogs and principally at the higher dose levels. In the soft tissues, moderate thyroid atrophy and very focai kidney degeneration were induced by the 2.9u.Ci 2-^9Pu/kg dose, but neither of these changes were detectable clinically. An increased incidence of soft tissue neoplasma was also noted in die irradiated groups, but this was not unequivocally established as a 239pu induced factor (12). In beagles injected with 23yPu (single intravenous) at different age groups such as 2 days, 3 months and 17-19 months, the highest bone tumor incidence was seen in dogs injected as young adults (17-19 months). Differences in bone tumors were primarily due to age dependent local nuclide distribution (6). The comparative toxicity effects of 239Pu and 238PuO2 were investigated by Pask et al. (7) by exposing beagles through inhalaticn. The 239Pu exposure-related effects included respiratoiy insufficiency, lung cancer, trachiobronchial iymph nodes sclerosis and IT - 24.6 lymphoponia in the four highest dose levels (40,150,700 and 2800 times the current minimum permissible lung doses for a plutonium worker). Whereas 239pu exposure related effects included lung cancer and /or bone cancer, lymph node sclerosis, lymphopenia and liver damage in five highest dosage-level group including 8,40, 150, 700 and 2800 times the current maximum permissible lung dose for a plutonium worker. Beagle studies show no increased hematopoietic neoplasea (leukemia) from internally deposited plutonium. Beagle studies also show that 239Pu is 16 to 19 times more efficient in inducing bone sarcomas than is 226Ra, based on a comparison of average doses to skeleton. This exceeds the ratio of 5 originally used to set standards for h>.man protection (13). Epidemiological studies involving persons exposed to Pu between the periods of 1944 and 1970, when the exposures were larger and more frequent, are being carried out by Voelz (14). These studies also point up the need to pay close attention to such serious compounding factors such as smoking histories, potential chemical exposures and external radiation including reaction exposures. In a study of white male workers at the Rocky Flat plant (Colorado, USA), there was a two to three-fold excess of intracranial (brain) tumors which was clearly shown not to be related to plutonium exposure, but the mortality rates for all causes of death and all cancers were low compared with those of US white males. Bone cancer, observed in plutonium exposed animals, was not observed in these workers, although one in a group of 26 Manhattan Project plutonium workers has developed an osteogenic sarcoma of the carum. The observation of one bone sarcoma is highly suggestive of a radiobiological origin. REFERENCES 1. G.T. Seaborg, "Plutonium Revisited", Radiobiology of Plutonium, Edited by Steven and Jee, University of Utah Printing Services, Salt Lake City, 1-21, (1972). 2. J.G. Hamilton, The Metabolic Properties of the Fission Products and Actinide Elements, Rev. Mod. Phys., 20, 718-728 (1948). 3. W.H. Laryharn, S.H. Bassett, P.S. Harris, and R.E. Carter, Distribution and Excretion of Plutonium Administered Intravenously to Man, Document LA - 1151, (1950). IT - 24.7 4. N.P. Singh, Concentrations and Organ Contents of Plutonium in World-Wide Genera] Populations, Health Phys., 58: Supple 1, S6, (1990). 5. N.P. Singh, Thorium, Uranium and Plutonium in Human Tissues of World-Wide General Population, Jour. Radioanaly. & Nucl. Chem., 138, 347-364, (1990). 6. F.W. Bruenger, R.D. Lloyd and S.C. Miller, Distribution and Retention of Plutonium in Beagle Dogs of Four Age Groups: A comparison. Health Phys., 58: Supple 1., S 25, (1990). 7. J.F. Park, G.E. Dagle, RE. Weller, R. L. Buschbom and G.J. Powers, Comparative 238 2 Distribution of Inhaled PuC«2 and 39puO2 in Beagles, Health Phys., 58: Supple 1, S 39, (1990). 8. P.W. Durbin, "Plutonium in Man: A new look at the old data", Radiobiology of Plutonium, Edited by Steven ;ind Jee, University of Utah Printing Services, Salt Lake City, 468-530, (1972) 9. J. Rundo, "The Late Excretion of Plutonium Following Aquisition of Known Amounts", Actinides in Man and Animals. Edited by M.E. Wrenn (Radiobiology Division, University of Utah), R.D. Press, Salt Lake City, 253-260, (1981). 10. J.F. Mclnroy, Distribution of Plutonium in Man, Health Phys, 58: Supple 1, S 6, (1990). 11. F.F. Harm, N.A. Gillett, S.C. Miller and G.N. Taylor, Pathological Observations in Dogs Exposed to Plutonium, Health Phys., 58: Supple 1, S 40, (1990). 12. G.N. Taylor, W.R. Christensen, L. Shabestari and W.S.S. Jee, "The General Syndrome Induced by 239Pu in the Beagle", Radiobiology of Plutonium, Edited by Steven and Jee, University of Utah Printing Services, Salt Lake City, 59-79, (1972). 13. M.E. Wrenn, Biological Behavior of Pu in Man and Experimental Animals: A Summary, Health Phys, 58: Supple 1, S 40, (1990). 14. G.W. Voelz, Epidemiology Studies of Plutonium Workers, Health Phys., 58: Supplel,S25, (1990). IT - 24.8 INTERNATIONAL SYMPOSIUM ON RADIOCHEMISTRY India, February 1991 METHODS FOR ANALYSIS OF URANIUM-PLUTONIUM MIXED FUEL AND TRANSpLUTONIUM ELEMENTS AT RIAR G.A.Timofeev V.I.Lenin Research Institute of Atomic Reactors, 433510 Dimitrovgrad, Ulyanovsk Region, USSR Different methods for analysis of the uranium-plu+onium mixed nuclear fuel and transr1utonium elements are briefly dis- cussed in this paper: coulometry, radiometric techniques, emission spectrograpby, mass-spectrometry, chromatography, spectrophotometry, The main analytical characteristics of the methods developed are given. Key words: transplutoniuni elements, uranium-plutonium oxides, chemical analysis. IT - 25.1 1. INTRODUCTION At the V.I.Lenin Research Institute of Atomic Reactors (RIAR) the process of complex extraction isolation and purification of tranaplutonium elements (TPE) from plutonium to californium has been developed and successfully applied ,'1,2/. The technology of TPE extraction and purification is complicated and requires a care- ful analytical control. At the Institute the works on the fast- neutron reactors fuel cycle are also being carried out. A pilot- plant for refabrication of oxide fuel elements, including uranium plutonium oxides, of the BOR-6O reactor has been set up /'i/. This paper deals in brief with the methods for analysis of the uranium-plutonium mixed fuel and TPE applied at the RIAH Analytical Lepartment. 2. METHODS FOR TPE DETERMINATION Quantitative determination of TPE arid plutonium is often bound up with the necessity of isolation of these elements from a solution. For successful employment in the technology control the extraction methods should not be too complex and various but at the same time should provide a fairly high extraction and puri- fication from the accompanying actinides, rare-earth and other elements. For plutonium extracting the use of 2 methods is most often made . One of those is based on the plutonium (IV) extraction with di-2-ethylhexylphosphoric acid (D2EHPA) in decane from nitrate solution. The other method is based on the plutonium(VI) extrac- tiu.:-chromatographic isolation and its separation from tri-and tetravalent elements with the help of 5-10 M HC1 A/. Americium separation from curium and other TPE is one of the most difficult problems in practice of the americium analytical determinat'. on in technological solutions. There are several methodB used for i,nis purpose. One of those described in Reffs. /5,6/ is baaed on the americium oxidation to hexavalent state and its fur- ther separation from curium and other elements deposited in the fo~n> of hardly dissolved fluorides. The separation method of americium (VI) and curium (III) by the hexavalent americium extraotion with D2EHPA in decane or In another non-polar diluent IT - 25.2 is widely adopted /6/. Isolation and purification of berkelium are accomplished by means of its tetravalent state extraction with the D2EHPA solution in decane from 6 M HNO-,. Californium, arnericium, curium are separated by the extraction-chromatographic method using D2EHPA: americium and curium are washed off from a column with 0.2 M HN0-, californium - with 4 M HN0~ /?/. Alpha-, beta-, gamma- and neutron radiation measurement methods remain some of the main techniques both for the quantita- tive determination of TPE in various technological solutions and with the estimation of radiochemical purity of their preparations /8-10/. For the last 10 years thanks to extensive application of semiconductor radiation detectors, improvement of instrumentation, computer facilities the opportunities for the quantitative analy- sis have been considerably expanded by means of radiometrlc methods. The careful sources preparation for measuring (in particular, applying of electrochemical deposition of radioactive samples on the substrate), the close adherence to measurement conditions, the correct energetic calibration of spectrometers, the nucleai— physical isotope constants revision have made it possible to significantly increase the precision and accuracy of TPE radio- metrical determination. The integral d-count methods are widely employed for deter- mination of the total (volume) alpha-radiation activity of the solutions as well as for estimation of plutonium, americium, cu- rium, californium, berkeliuir. (with reference to the daughter 249 Cf accumulation) upon separating these elements from other alpha-emitters. To measure the alpha-radiation use is made of the proportional flow counters in 25"- or 40T -geometry as well as those on the base of camera with a definite solid angle. As a rule to determine curium its preliminary chemical extraction is not required as alpha-radiation of its nuclides makes the main con- tribution into the total alpha-emission activity of the solutions. In determining californium by means of alpha-radiation measuring 252 the contribution due to the recorded Cf spontaneous fission fragments is taken into account. To analyse the alpha-emitter mixturefj the alpha-spectrometric method is used. The gamma-spectrometric method is widely applied to control a decontamination from fission products and radiochemical purity IT - 25.3 of finished TPE preparations. Ge- (Li)-detectors with multichannel analysers are used. The intensity measurement of JT- and X-ray radiation is also employed for TPE quantitative determination and analysis (americium in curium, in mixture of Pu+Cm, plutonium in Am+Cm mixture, americium and curium additions in berkelium and californium preparations). Use is made of the alpha-spectrometric method in combination with gainma-spectrometry to measure isotopic ratios. The measurement of ^-radiation rate is applied for berke.i.ium determination after its extraction and thorough purification from other ^-emitters involving cerium. To I1*-.termine californium amounts in various solutions and neutron sources use is made of measuring and comparing of neutron emission with that of the reference source under rigid identical geometrical conditions. The gamma-scanning method to determine and control the uniform californium deposition along the medical source is successfully designed and applied /11/. The extensive use has been made of the mass-spectrometric method for the determination of plutonium and TPE isotopic com- position as well as for the exact estimation of these elements in the analysed samples /12, 13/• The mass-spectrometric method for version of an isotopic dilution is one of the main techniques used to determine precisely plutonium and americium in the initial solutions obtained upon dissolving the irradiated targets with TPE. For the isotopic analysis upon chemical cleaning from non- r&aioactive impurities, a sample containing 0.1-0.5 J*-B element (0.01-0.05 f*g for californium) is pipetted as a nitrate solution on an evaporator strip. Usually no less than 20 mass-spectra are taken in the analysis of every sample. By analysing the relative mean-square deviation, S , it depends on the isotopic ratio in mixture and changes as follows: in ratio 1:1 S =0.002 r 1:10 S =0.01 r 1:50 Sr=0.07 Applying of the highly effective ion source makes it possible to increase essentially the sensitivity of mass-spectrometric measure- ment and to analyse up to 10" g element /14/. The procedures for mass-spectrometric determination of plutonium, TPE and rare earths (RE) in uranium, plutonium, americium, curium, berkelium IT - 25.4 and californium samples have been developed /15-19/. Plutonium and amerieium in curium to 10 %, plutonium, americium and curium in berkelium and californium to 0.3% and rare-earths in berkelium _2 and californium to 3 10 % can be determined with 0.1-10 Mg by mass. The root mean-square deviation ranges from 0.1 to 0.2. In the analysis of pure plutonium solutions use ie made of the own plutonium (VI) light absorption at 803 nm for the plutjnium determination. In case of americium use is made of specific band of americium (III) absorption at a maximum of 503 nm. In particular, spectrophotometric americiuir. identification in its curium and RE mixture has been described /20/. For the present the epectrophoto- metric methods are not employed in technology control of heavier TPE extraction. The extensive use is made of complexonometric methods with diethylenetriaminepentacetlc acid (DTPA) application to determine plutonium (IV) in presence of trivalent HE and TPE as well as americium and curium in solutions and preparations. Plutonium can be determined by DTPA complexonometric titration with xylenol orange as an Indicator in 0.3-0.4 M HN0-. (HC10.) to the accuracy of 0.5-1-5% at 0.2-0.8 mg of its content in the sample /21/. The methods have been developed to determine tracer amounts of americium and curium by DTPA solution titration with visual spectrophotometric or potent:.ometric indicating of the end-point of titration /20/. The estimation error is 0.5-2.0%. In coulo-complexonometric measurement the electrochemically generated complex of Fe (II) in combination with DTPA is used as a titrant 1221. This makes it possible to carry out the analysis with the help of an automatic coulometric titrator. Coulometric methods are some of the most precise techniques for the plutonium and TPE determination. The automatic coulometric titrating method is successfully used for the determination of plutonium in solutions, oxide and metal. ?• n this method plutoni- um is quantitativel.7 oxidized to the hexavalent state with argentic oxide (II). Then plutonium (VI) is reduced by the electrically generated Fe (II) ions at constant current in sulfur-phosphate solution. Plutonium (VI) amount in the solution is estimated by the energy consumption when generating ferrum (II) /23/. The described method allows to determine 2-5 mg plutonium in a sample with an error up to 0.1%. IT - 25.5 The method of americiun; coulometric determination in 2 M H^PO^+0.1 M HCIO^ solution allows to identify up to 5-10 of americium in a sample /24/. Americium (III) is oxidized to americium (IV) at the anode potential 2.0 V and then it is reduced to arnericium (V) at the potential 1.27 V. The error of americium determining,when its content is more than 30 mg7is 1.5%. For the coulometric berkslium determination 1 M HpSO, solu- tion or 1 M HNO^fO.I M HpSO^ mixture are the most suitable electro- lytes. Berkellum is oxidized at the 1.85 V anode potential and then reduced at 0.70 V potential and whan reducing the energy consump- tion is estimated whereby berkellum content is determined in the sample to be analysed. With tOj^g/ml berkelium concentration the error is about 1-2% /25/. There are two versions of emission-spectral methods to deter- mine impurities in plutonium, americiun], curium: direct spectro— graphic and chemical-spectral determination with preliminary matrix separating (for americium and curium) that allows to de- crease essentially the detection limits /26,27/. In the first version the 0.5 ml sample containing 0.2 mg of Pu, Am or Cm is pipetted on a carbon electrode and emission spectrum is taken exciting it in an alternating current arc within 250-450 nm. The elements being determined and their detection lii. its are pre- sented in Table 1. In the 3econd version 121J the impurities from analysed samples (5-10 m;s by mass) are isolated by the extraction- chromatographic method on silica gel with D2EHPA. TPE separation from RE is performed by washing of the column with 0.03 M DTPA+ +0.3 M citric acid solution at pH=2.9. Fractions containing im- purities are evaporated and spectrographically analysed. The de- tection limits of this method are lower than those of the direct technique by a factor of 5 and in addition to the elements listed in Table 1 lithium, potassium, rubidium, cesium, strontium, barium, beryllium, yttrium, lanthanum, manganese, cobalt are determined. The investigation of berkelium and californium emission spectra excited in the alternating current arc made it possible to develop the method of direct spectrographic determining of Al, B, Ca, Fe, Mg, Mn, Mo, Na, Ni, Pb, Si, Ti impurities in ber- kelium and californium solutions /28/. The spectra of dry sample residues are excited in the alternating current arc and recorded IT - 25.6 at two speotrographs simultaneously with wave lengths 250-336 and 504-675 nm. The absolute impurity detection limits are —8 —Q within the range of 10 - 10 y g. Table 1 Analytical lines and detection limits of direct "jpectrographic Am and Cm oxides analysis Elements Wave length of Detection limit,% analytical line, nm Aluminium 308.21 0.0025 Silicon 288.16 0.005 Cadmium 326.10 0.025 Boron 249.68 0.006 Natrium 330.23 0.025 Calcium 422.67 0.005 Magnesium 285.20 0.002 Nickel 305.08 0.008 Chromium 284.33 0.008 Ferrum 259.95 0.008 Cerium 422.20 0.08 Praseodymium 422.29 0.0? Neodymium 424.73 0.05 Samarium 442.43 0.07 Europium 290.66 0.02 Gadolinium 303.28 0.008 3. METHODS FOR URANIUM-PLUTONIUM MIXED FUEL ANALYSIS After quantitative Pu0« or (U,Pu)0p dissolution, plutonlum or uranium is determined by automatic coulometric titratlon /23/. The plutonium and uranium-plutonium mixed oxides are analysed for metal impurities spectrographically using the fractional dis- tillation method /29/. 30 mg of the ground oxide sample are mixed with a carrier-In or Ga oxide- and placed into a carbon electrode orater. Antimony, germanium and silver are added to the samples and references as inner standards. References are based on uranous- uranic oxide. The alternating current arc is applied for exciting the samples and references spectra. Table 2 presents impurity IT - 25.7 deteotion limits. The relative standard deviation during the element determination on the detection limits level in Table 2 ia 0.25-0.30. Table 2 Impurity detection limits for (U,Pu)O? by fractional distillation with a carrier in 30 mg samples, % Element Concentration^ Element Concentration,% Fe 1 10~3 Ba 1 10~3 Mo 5 10~3 K 1 10"3 B 6 10~"5 Na 1 10-3 Ni 1 10"3 Li 3 10 Cr 1 10"3 Ca 1 10"3 Al 1 10~3 Si 3 10 Be 3 10~5 Mn 1 10~4 Mg 1 10~3 Cu 3 10~ Pb 1 10~3 Ti 5 10~3 Cs 1 10~3 The carbon content in PuOp and (U,Pu)0p ia determined chro- matographically in the form of COp following the sample pyrolysis at 1100-1200 C in a quartz tube under oxygen. Carbon dioxide enters the gaseous chromatograph detector. The relative standard deviation of this method in the concentration range from 1-10 to 1% is 0.05-0.003. The analysed sample mass is about 1 g. The detection limit is 1^*g. Chlorine and fluorine in uranium-plutonium mixed oxides are determined aft^r pyrohydrolysis of the sample at 85O-9OO°C in a quartz tube under water vapours and oxygen /23/. Chlorine and fluorine in tha form of volatile acids are trapped in a distilled water receiving tank. The chlorine cone^itration is determined by the spectrophotometric method using the ferrous thiocyanate complex; fluorine is determined from lanthanum-alizarine complexon. The relative standard deviation with a chlorine concentration of 3-1O~3fl is 0.07 and with a fluorine concentration of 1«10~3% is not higher than 0.05- The analysed sample mass is about 1 g. IT - 25.8 The lower detection limit is 1 ug for chlorine, 0.5 J IT - 25.9 REFERENCES 1. Yu.S.Zamyatnin, A.V.Klinov, A.G.Rykov. Radiokhimiya, V.19, N 4, P.421-425, 1977. 2. V.M.Nikolaev, Ya.N.Gordeev et al. Journ. of the Less-Common Metals, V.122, P.401-410, 1986. 3. H.Steii.kopff, R.Krompafl, K.Shumann, V.A.Tsykanov, O.V.Sklba, P.T.Porodnov, Kerntechnik, V.55, N 4, P.243-246, 1990. 4. B.I.Levakov, u. A. Timof eev. Zhurn. anal, khioi. , V.29, N 5, P.1023-1025, 1974. 5. P.L.Moore. Analyt. Chem., V.35, N 5, P.715, 1963. 6. P. Penneman , T.Kinen. Americium and curium radiokhiniiya, Izdat. Inoctr. Lit., M., 1961, P.59. 7. B.I.Levakov, V.B.Mishenev, G.A.Timofeev. Theses of the All- Union conference reports on extraction and sorption methods application for actinoids and lanthanoids extraction and separation. M., Nauka, P.11-12, 1984. 8. B.F.Myasoedov, L.I.Guaeva, I.A.Lebedev et al.Analytical chemistry of transplutonium elements. M., Nauka, 1972. 9. G.A.Timofeev. Methods of analytical control in transplutonium elements technology. In book: Modern methods of radioactive elements separation and determination. M., Nauka, P.22-29, 1989. 10. G.A.Timofeev. Radiokhimiya, V.28, N 2, P.232-234, 1986. 11. V.V.Pevtsov, Yu.S.Popov, N.S-Kurochkin et al. Hadiokhimiya, V.21 , N 1, P.108, 1979. 12. V.Ya.Gabeskiriya, Yu.S.Popov, V.B.Mishenev, G.A.Timofeev. ktomnaya energiya, V.55, N 3, P.175-176, 1983. 13. V.I.Boriaenkov, A.M.Mitin et al. Radiokhimiya, V.32, N 1, P.105-107, 1990. 14. V.V.Kalygin, V.Ya.Gabeskiriya, V.I.Raiko et al. PTE, N 4, P.181, 1980. 15. A.P.CheLverikov, V.Ya.Gabeakiriya, V.V.Tikhomirov. hadiokhimiya, V.19, N 4, P.447-451, 1977. 16. A.P.Chetverikov, V.Ya.Gabeskiriya, V.V.Tikhomirov. Radiokhimiya, V.21, N 1, P.132, 1979. 17. V.V.Tikhomirov, A.P.Chetverikov, V.Ya.Gabeakiriya. Radiokhim..ya, V.22, N 3, P.435, 1980. IT - 25.10 18. V.V.Tikhomirov, V.Ya.Gabeskiriya, A.P.Chetverikov. Radiokhimlya, V.23, N 4, P.595, 1981. 19. V.V.Tikhomirov, A.P.Chetverikov, V.Ya.Gabeskiriya. Radiokhimiya, V.23, N 6, P.896, 1981. 20. G.A.Timofeev, G.A.Simakin, P.F.Baklanova et al. Zhurn. anal. khim., V.31, N 12, P.2337, 1976. 21. A.G.Rykov, E.M.Piskunov, G.A.Timofeev. Zhurn. Anal, khim., V.30, N 4, P.713, 1975. 22. G.A.Simakin, G.A.Timofeev, N.A.Vladimirova. Radiokhimiya, V.19, N 4, P.560, 1977. 23. V.M.Barinov, V.M.Chistyakov, G.A.Timofeev. Radiokhimiya, V.25, N 6, P.734, 1983. 24. V.Ya.Frenkel, Yu.M.Kulyako, I.A.Lebedev et al. Zhurn. anal, khim., V.35, N 9, P.1759, 1980. 25- G.A.Timofeev, V.M.Chistyakov, E.A.Erin. Radiokhimiya, V.28, N 4, P.498, 1986. 26. V.M.Barinov, N.A.Adaikin, E.V.Savchenko, V.T.Filimonov. Radiokhimiya, V.21, N 1, P.121, 1979. 27. V.M.Barinov, T.Ya.Vereshchagina, Yu.I.Korovin, G.A.Timofeev, V.T.Filimonov. Spectrochim. Acta, V.36B, N 12, P.1215, 1981. 28. V.M.Barinov, V.I.Konovalov, E.A.Erin, G.A.Timofeev. Theses of the 2nd Ail-Union conference on transplutonium elements chemistry. Dimitrovgrad, P.95, 1983. 29. A.N.Zaidel, N.I.Kakiteevsky, L.V.Lipis, M.P.Chaika. Emission spectral analysis of atomic materials. Gos. Izdat. FML, Leningrad-Moscow, P.314, 1960. 30. K.Teske, C.Nebelung, I.I.Kapshukov et al. Journal of Nuclear Materials, V.168, P.97, 1989. IT - 25.11 ANALYTICAL CHEMISTRY OF PLUTONIUM H.C.Jain Fuel Chemistry Division Bh.ibha Atomic Research Centre Troathay, Bomba.y-4 00 0H5 I.INTRODUCTION Analytical chemistry of plutoniuin is quite fascinating as well as challenging. Fver since the discovery of plutonium as the second transuranium element in 1940, formed by bombarding uranium with deuterons in the cyclotron at BerVeley, the different analytical aspects of plutonium are being investigated continuously. These include the physico-chemical properties, nuclear parameters and methods for the determination of plutoniuin ranging f roin one million atoms in the environment to multi- kilograms at the reprocessing plants. The analytical chemistry of plutonium is challenging owing to a 'Uiiiber of reasons. For example, it is known that different oxidation states of plutoniuni, Pu(111) , Pu(IV) and Pu(Vl), can coexist in aqueous solutions. Attempts have also been maiie to prepare higher oxidation states of plutonium in the alkaline medium. A Pu(II) oxidation stcjte has also-been identified in solid melts. Fifteen isotopes of plutonium, Pu to Pu, have been synthesized and characterized for their nuclear properties. The different radiations like alpha particles, famma rays, X-rays and spontaneous fission neutrons emitted by various plutoniuin isotopes have formed the basis of a number of methods for pi utoni um del erini na t ion . Analytical chemistry of plutonium encompasses different aspects e.g. solution chemistry, solid state chemistry, nuclear chemistry aiid methods for its determination. In this paper, the deterini na t i on of plutoniuni [1] which is one of the important aspects of analytical chemistry is deal? with. This involves the determination of isotopio composition and c< ncen+rat f on/amount of plut.oni urn at different levels in a variety of matrices. Because of its strategic importance, highly toxic nature, and also from the critical jt.y point of view, sensitive, precise and accurate methods are essential. The selection of the method is governed by the type of matrix, the amount of plutonium present and the precision and accuracy requirements. This paper reviews the different methodologies available for these measurements with emphasis on the work performed in our laboratory. 2. SAMPLE TREATMENT Quantitative determination of plutonium in a sample may involve sample dissolution, aliquoting, redox treatment, separation and purification, and use of a tracer for chemical yield. The dissoltit i-on method is dictated by the type of matrix. For fuel materials, different combinations of acid, mixtures have been tried. For oxide sample^, HNO3 with traces of HF and digestion under an infra-red lamp is commonly used.. A mixture of IT - 26.1 HNO3 and H2SO4 with refluxing has been developed in this laboratory for oxides, carbides and samples containing free carbon [2J. Alloy samples like Pu-Zr-AI are dissolved in HC1 or in cone. HNOcj containing traces of Hg . A method employing H2SO4—dil. HNOij mixture has also been tried for the alloy samples [3J. In environmental and biological samples, acid leaching and/or wet digestion and dry ashing methods with or without a tracer are employed. For carrying out quantitative work, an accurately weighed sol id sample is dissolved and made up in a pre—weighed volumetric flask. The aliquots for plutoniurn det erm i rui t ion are taken on weight basis using polyethylene weight burettes. This eliminates the volume calibration and pipetting errors and assures an ."curacy of better than 0.005% in aliquot ing. For methods which cannot tolerate bu1k of matrix, the separation and purification of p 1111 oni uni should be performed after the addition of a known amount of tracer. Treatment with cone. HNO3 and redox reagents for depo lynier i za t i on and chemical exchange of plutonium isotopes in the sample and the tracer is also essential. Solvent extraction and ion exchange methods are used for the separation and purification. 3. RAIHOMETRY 3.1 Gross A1pha Counting Gross alpha counting of purified plutonium is commonly performed in a gas flow proportional counter or a low geometry counter with semiconductor detector. Liquid scintillation counting is also sometimes employed. For determining the amount of plutonium, different parameters like specific activity of plutonium, counter efficiency,, quantitative source preparation on a weight basis, backing material, aqueous or organic medium, evaporation ;>f the solution and ignition of the source at controlled temperature must be kept in mind to get the desired rtccu racy of" 0.5 to 1%. AH an ex«ni[>le, I he specific activity of plutonium can vary by a factor of 5 depending upon the origin of p I u I on i uiii. Oilier parame I er s can give rise to errors ranging from 1 to 30%. Mi.ny of the above mentioned parame t ers were studied as early as in 1966 in our laboratory [4J- For example, the specific activity of p 1 u t on i tun available at that t i me was det errni ned [ 5 ] using (i) e 1 ec t rochemi ca 1 method ( pot en t i ome t ry ) for p I u t 011 i urn amount; the method was checked through the use of anhydrous plutouiuin sulphate standard prepared in our 1aboratory, (ii) gross a 1 pba . count, i rig with counter" efficiency detensined using a standard Am source qu polished stainless steel planchet as well as by using Am standard solution from IAKA, til e<; t opol i shed stainless steel planchets and a 1 pha liquid scintillation counter, (iii) alpha spectrometry tor 2.JHPu/( * Pu+ Pu.) alpha activity ratio and (iv) subtracting the Pu alpha activity froin_gcoss alpha counts and determining the percentage of Pu and Pu from,the remaining alpha specific activity. The contributions of Pu and Pu to the alpha specific activity were assumed to be negligible as plutonium was from a low burnt fuel. The Pu content was obtained from the ingrowth of Am. Finally the isotopic composition including the IT - 26.2 percentage of 242Pu taken from the published literature on a CANDU type reactor was arrived at. The isotopic composition and specific activity data agreed well with i-Iie results obtained later in 1970 using a thermal ionibation mass spectrometer. 3. 2 AJjjha Spectrometry In alpha spectrometry, there is an inherent problem of tail contribution at the low energy peak due to energy degradation of high energy alpha particles. This governs the grgc i si i go- 2§H^ accuracy achievable in de t errni ni ng Pu/( ' Pu+ Pu) alpha activity ratio. Our laboratory participated in two international int.ercoiwparison experiments, ASfiT-74 and AS-76, conducted by KfK, Karlsruhe for the..eva 1 uat ion of alpha spectra and the determination of ' P"/( ' Pu+ Pu) alpha activity ratios. The results reported by our laboratory were in excellent agreement with the mean of means. In uur laboratory, a method based on the geometric progrtcriioi; (G.P.) decrease of the counts in the low energy tail of the alpha spectrum has been MStfd f°] to correct for the tail contribution of_the higb_energy Pu peak (5.499 lieV) at the low energy peak ot Pu + Pu (5.155 and 5.168 MeV). Precision and accuracy in the determination of alpha activity ratios were evaluated by preparing synthetic mixtures from solutions ot Pn and Pu isotopes and covering a range of 0.01 to 10 for the Pu/( Pu+ Pu) alpha activity ratio. The G.P. method has been adopted by a number of other international laboratories for the precise and accurate determiwation^of alpha activity ratios. Alpha spectrometry using Pu and " Pu as spikes (IDAS and R-IDAS respectively) Cor low and high burn-up plutonium samples, respectively, in isotope dilution, has also been developed in our laboratory and is capable of giving precision and accuracy of 0.5% in the concentration determinations [7]. The data on the t. so topic composition of plutonium needed in IDAS and S 73ge PJ-JPXided by mass r.pectrornetry. A correlation between Pu/( Pu+ Pu) alpha activity ratio and alpha specific activity or plutonium has been investigated in fuel from CANDU type reactor [8]. This eliminates the need al isotopic composition data and can be used for sample* 1 r uiii waste streams in the reprocessing plants. A number of other international 1 aboia tor i HS , have recently reported t lie use of IDAS for pJutonium concentration measurements. The IDAS technique is now becoming popular and IAEA in collaboration with USSR is conducting an international inter-comparison experiment (IDAS—90) to evaluate the performance of -he methodology. We have already received the samples for measurements. Another technique using "U .as a non-isotopic diluent has also been reported from our laboratory for determining plutonium concentration [9]. Recently, the passivated ion implanted silicon detectors (IPK) have been introduced for carrying out high resolution alpha spectrometry. These detectors have the capability of giving better resolution of about 10 keV (FWHM) at 5.5 MeV compared to a resolution of 15 keV obtained by conventional silicon surface barrier detectors. They are attractive for partially resolving close lying alpha energies e.g. Pu at 5.155 MeV and Pu at 5.168 MeV and apptar IT - 26.3 promising for determining 240Pu/239Pu ratio by alpha spect lometry. A few experiments have been conducted in our laboratory to use the 1PR detector for this purpose [10]. 3.3 Gamma Spectrometry Gamma spectrometry using high resolution, high purity germanium (HPGe) detector is useful for determining " Pu to Pu isotopes. The method requires at least about 500 micrograras of plutonium and cannot provide any information on Pu content [11]. The abundance of Pu can be known using a suitable isotope correlation. Precision and accuracy values of 2 to e>% can be achieved in determining the isotopic composition. Gamma spectrometry is extremely useful as a non-destructive technique for verifying the isotopic composition of plutonium. Energy region of 94 to 122 keV has been recommended for use in plant conditions for plutonium isotopic composition measurements within a short time. A fuel pin scanner based on gamma spectrometry has been developed in the Radiochemistry Division [12] and used for the assay of plutonium and quality control regarding active length and presence of non-standard pellets. 3^4 Neutron Assay Techniques Neutron coincidence counting techniques are useful for t lie determination of plutoniuin in sealed containers. The method makes use of the multiplicity of the prompt neutrons, which, ..are emitted within 10 sec of the spontaneous fission, of Pu. The fission rale information along with the isotopic composition data is useful to assay plutonium. Neutron well ..coincidence counter generally consists of an annular array of "He or BF3 detectors in a moderator assembly with a central well for keeping the sample. The sensitivity, precision and accuracy in these measurements are being improved by devising different methods. These counting systems have been developed in the Radi ochemi^try Division and are being used [13]. For active neutron assay, ' Cf or Am—fte neutron sources have been investigated. Am—Be neutron sources appe.ii lo be competitive.' with ' Cf sources from the point o t view of shielding requ i reinents and the neutron flux i i ri I) i i i t y . 3_.J5 .Kr-Kdge Dens i tome try In K-edge denis i t oinet ry , the amount of pi is ton i urn in the sample in related to the measured transmission ratio determined directly above and be low-the absorption edge energy 121.8 keV of p!'.ito^JLurn. Sources of Co (Tj/2 = 271.3 days, Er =122.1 keV) and "" Se (J\l/2 = 120 days, Er = 121.1 keV) have been used for this purpose [14J. Encouraging results from the field test at the reprocessing plants [ 1 5 J have shown that this method along with gt.mma speotrorne t. ry ' for- isolopi" compos it ion may replace the conventionally used redox titration and mass spectrometric methods. A hybri 1 K-edge densitometer / X-ray fluorimeter ha.s been constructed at KfK, Karlsruhe [16] and tested under the field conditions for determining plutonium in the highly radioactive input solutions. K-edge densitometry was used for uranium determination and U/Pu ratio (range 100 to 120) was determined by X-ray fluorescence. This approach also appears IT - 26.4 attract ive from the point of view of obtaining qui.-k rcssil at the reprocessing plant. A nev method called Thr^e Energy Way Absof|)l iomftry [17] . rhi»A) using mixed sources of He (121.1 and 198.6 keV) and Ba (81 keV) and measuring ratios of the transmitted peaks at 121 and 198 keV for uranium end 121 and HI keV for- p 1 uton i urn has been used for- simultaneous and independent determination of uranium and plutonium. The techni que is in a preliminary stage of development and is sens i t i vc to matrix material e.g. acid molarity. 4. RI.KOTROCHfirtlOAI. MKTKODS Electrochemical methods based on redox titrations find applications for the detern.inat i on of plutoriium in fuel samples. The methods require 2 to 5 nig of plutonium and detect the end point using potent, iometrj i amperomei. ry or cou ! orne t ry . Since p I u ton i utii i egenerally present jD the pure form, these methods are caoabIe of providing precision and accuracy of 0.1%. The potent i ometrie and amperomet ri c methods involve the .ise of AgO for the oxidation of Pu to Pu (VI) , fiiilptwrnic acid for destroying the excess of AgO, addition of a known excess of Fe(II) for the r educt i on if <>u(V!) to F'u(lV) and ti (rat ion of the excess Fe(II) with Cr(VI) UK ing a platinum indicator electrode and a reference Srt t iifri t ed c« I ome I electrode. The potentiometric method was developed in 1968 [18] after overcoming the limitations of methods reporied by Druimnond and Grant [19] , and Milner et al [20] and is in routine use along with the amperometric method. Alternative oxidanls (Co(III), Ni(lII)) and reductants (Cu(l), Ti (I I I ) ) have been developed and used for plutonium determination l>y the electrochemical methods [21-25J. In the controlled potent ial eon 1 oinet r i c method, the problems ot organics generated during the dissolution of the sample arid the accumulation of iron due to the use of large volumes of HNO-j in dissolution have been overcome. Another innovation introduced in the coulometric method was the use of successive addition approach whereby 10 to 1% aliquots could be analysed. This reduced ttie volume of tht analytic-si waste and t lie analysis time without degrading tiw quality of results 5. MASS SPKCTROMHTBY 5.1 Ijs<2_top_-Lc^ Compos i tion Thermal ionization mass spec t.r omet ry (TIMS) is a recognized technique for the determination of isotopic composition of plutonium. The method requires sub-mi crogram amounts of plutonium and has the potential of providing the highest precision and accuracy. New developments have recently been introduced in TIMS. The advent of extended geometry, high sensitivity (1 ion per 25 atoms of piutoniurn), fully automated thermal ionization mass spectrometers coupled with the variable multi-col?ector Faraday Cup Detectors has considerably improved the sensitivity, reduci ,i the time of data acquisition and enhanced the precisic ii and accuracy of isotope ratio measurements. The major isotopes (abundance more than 10 atomX) can be deteimined with a precision and accuracy of better than IT - 26.5 0.01%. The lower abundance isotopes can be determined with precision and accuracy values of 0.1 to 1% depending upon their abundances. 5.2 Concentration by IPMS Isotope dilution using thermal ionization masn spectrometry (ID-TIMS) is well recognized tor determining plutonium concentration in the highly radioactive dissolver solution of irradiated fuel. It does not require quantitative recovery of plutonium and sub-microgratn amounts of piutonium are sufficient. Enriched Pu is commonly used as a epike (tracer) for determining the plutonium concentration by ID-rrplS [27\. Pu has also been used as a spike and it offers the advantage of providing the isotopic composition data simultaneously from a single mass spectroirmtr i c analysis. ^Iq our laboratory, due to the difficulties in getting enriched Pu and Pu, alternative spike ~ Pu has been investigated and developed [28] . Pu ( > tyQq atomS) is used as n spike for plutonium samples containing Pu ( < 70 atom%) and vice-verfia. The ID-TIMS technique holds the potential of giving precision and accuracy of 0.1% in Plutonium concentration determinations in discolver solution. Recently, 1arge batches of * Pu spikes have been prepared by the Safeguards Laboratory of IAEA to be used for plutonium concentration determination [29]. An improved surface ionisation diffusion thermal ionisation source has been used in TIMS for the determination of 10 atoms of Pu in samples obtained from nuclear debris and in human tissues, with a precision of about 10% [30]. 5.3 Tracer Techniques for Input Accountabi1ity Determination of total plutonium in the input accountability tank of a reprocessing plant is one of the most important measurements and demands the highest precision and accuracy. From our laboratory, we have earlier shown the possibility of using Magnesium and Lead as tracers (MAGTRAP and LEADTRAP) for determining the volume of solution in the input accountability tank as well as for determining the total amount of p 11: toni um/urani um in the tank without invoking the knowledge of aliquot size and volume of solution in the tank [31]. Precision and accuracy values of 0.5% were demonstrated. The MAGTRAP and LEADTRAP experiments have been followed by a Reprocessing Input Tank Calibration Exercise (RITCEX) [32] by EURATQM countries using Lead (Pb), Neodymium (Nd) and Lutetium (l.«i) as tracers. Recently, 3nother element Erbium (Er) has been proposed as an equally useful tracer [33]. 6.OTHER TECHNIQUES *JLi! Resonance Ioni zation Hasa Spectrome try Resonance ionization mass spectrometry (RIMS) using lasers has been reported tor the determination of isotopic composition and concentration of plutonium at 10 atoms of Pu for the environmental studies [34]. During thede measurements, plutonium was sandwiched between a rhenium filament and a thin rhenium layer by art electrolysis procedure, and evaporation of IT - 26.6 the atoms was achieved by heating the filament to about 1BOOo. The ions were detected by a channel plate detector after a drift length of 2 meters in a t ime-of - f 1 j ght mast; spectrometer. Isotopic composition of two plutoniuin samples as determined by a series of RIMS measurements and by TIMS agreed within their limits of errors. 6. 2 Induct, i vely Coujpjj^d P) asma Hasa Spec trometry With the increased availability ot Inductively Coupled Plasma Mass Spectrometers, the^techniqye has recently been ufc-ed for obtaining data on the Pu/ " Pu atom ratio in the environmental samples [ 3 *5 J . 6.3 Ly/Aipha Ray Kaii ssion Ratio ~ This techni qug-has^beeri used by a group from Japan for the determination of Pu/ Pu atom ratio and..is based on r,the differences in Lx/alpha ray emission ratjo for Pu and Pu [36] . 6.4 Fission Track Method The fission track method is tidsed on Lhe fact that 239 Pu undergoes fission by thermal neutrons while Pu does not. The amount of plutnniuin is determined by comparing the track density of the .sample with that, of the standard |35,37]. b. 5 Laser Raman Laser Raman Spect romei.ry has been reported [38] for the determination of Pu(IV) in HNO3 solutions, using U(VI) as an internal standard. Concentrations of Pu(VI) and Pu(V) as a function of time were measured in Pu(VI) solutions of different acidities containing H2O2 as a reducing agent. 6.6 Spect ropho tome I r_y A d i. f f eren t i % in the concentration range ot 0.1 to 0.3 mg/g was obtained. 6.7 X—ray Fluorescence X-ray fluorescence technique using thorium as an internal standard has been tried in our laboratory on puie plutoniuin solutions and on (U,Pu)t)2 solid samples [40]. 6. 8 Neutron, Act Wat. j on Plutonium in the environmental and biological samples can be determined using neutron activation and high resolution gamma spectrometry. 7. CONCLUSIONS A number of chemical and instrumental methods are available for the precise and accurate determination of plutonium. Various techniques like mass spectrometry, alpha spectrometry, gamma spectrometry and electrochemical methods are routinely employed in nuclear technology fox- the determination of IT - 26.7 pi u t cm i urn, depend j ng upon the type and nature of the sample arid 1 'in aiiioimt of [> I II t on i um present . Existing methodologies are constantly being improved to meet the requirements. Lasers are being employed for the determination o* piutoniinn at extremely low levels (10 atoms). 8. ACKNOWLEDGEMENTS The work reported in this paper has been carried out by different groups in the Fuel Chemistry and R&dioehemistry Divisions. The author is tharikfu! to Dr S. K. Aggarwa 1 and iJr K . L . Raniakumar for their help in preparing the manuscript. M.V.Ramaniah, Artificial Radioactivity, Edited by K.N.Rao and H.J.Arnikar, Tata McGraw Hill, 1985, p.191. Keshav Chander et al, Nuc:l. Technol., 78, 69 (1987). S.P.Hasilkar et al, Preprints Volume of Symposium on Rad i ocheini st ry and Radiation Chemistry, IGCAR, Kalpakkam, Jan. 4-7, 1989, Paper No. RA-21. N.Sririivascin et al, AEKT-265 (1966). 5. N.S r i n ivasau et al, BARC Report. BARC/I-87 (1970). 6. S.K.Aggarwa1 et al, Radiorhicn. Acta, 27, 1 (1980). •7 M.V.Raman i ah et al, Nucl. Technol., 49, 121 (1980). 8. G.Chouras i ya and H.C.Jain, Pr •. prints Vol. of Symposium on Radi oehemi s t ry and Radiation Chemistry, IGCAR, Kalpakkatn, Jan. 4-7, 1989, Paper No. RA-33. 9. S.K.Aggarwrtl et al, Radiochim. Act a, 41, 23(1987). 10. S.K.Aggarwa1 et al, Preprints Vol. of Symposium on Had i oeheni i . t ry and Radiation Chemistry, Nagpur, Feb. 5- 8, 1990, Pcp«r No. NC-8. I 1 S.fl.Maiiohar et al, J. Radioanal. Chern., 6_3, 145 (1981). 1 2. P.P.Burte et al, Preprints Volume I of the Seminar on Fast Reactor Fuel Cycle, IGCAR, Kalpakkarn, Feb. 10-12, 1986, p.207. 13. S.B.Manohar, Preprints Vol. of Radiochemistry and Radiation Chemistry Symposium, I IT, Kanpur, Dec. 9-13, 1985, Invited Talk No. IT-9. A. Ramaswami at. al, Preprints Vol. of Symposiujn on Rrid i ocheini st ry and Radiation Chemistry, IGCAR.. Kalpakkam, Jan. 4-7, 1989, Paper No. NT-05. t r>. B . 1.. Tay 1 or and l,,('.Russen, Nuclear Safeguards Technology, Paper No. IAKA-SM-293/25 (1986). 16, H.Ottiiiar et ai, Ibid, Paper No. IAEA-SM-293/48. 17. M.Aparo et a I, Ibid, Paper No. IAEA-SM-293/80. 18, C.I,. Rao et al, Fres. Z. Anal. Chem., 254, 126 (1971). 19. J. I.. Druimnoiid and R. A. Giant, Talanta, J3, 477 (1966). 20. C.W.C.MiIner, A.J.Wood and G.E.Casie, AERE-R-4975 (1965). 21 . N. Gopinath and G.A.Rama Rao, J. Rftdioanal. Nucl. Chem. Let t . , 104, 7 (1986). 22. N. Gopinath et al, ll>id, 105, 7 (1986). 23. P.R.Nair et al, Fres. 7.. Anal. Chem., 315, 355 (1983). 24. P.R.Nair et al, HARC Report, BAHC/1-822 (19H6). IT - 26.8 25. P.R.Nair et al, J. Radionnal. Nu<;l. Chem. , Articles, 1_22, 1* (1988). 26. H.S.Sharma et al, Nucl. Techno 1., 89, 399 (1990). 27. S.A.Chitambar et al, BARC Report, RANC-865 (1976). 28. S.K.Aggarwal el al, Int. J. Mass Spectrom. Ion Proc., 69, 137 (1986). 29. G.Jammet et al, Report. No. IAfiA/Al/029 (1990). 30. R.E.Perrin et al, Int. J. Mass Spectrom. Ion Proc, 64, 17 (1985) . 31. C.K.MathewB et al, Nucl. Technol., 42, 297 (1979). 32. R.Carchon et al, Nuclear Safeguards Technology, Paper No. IAEA-SM-293/168 P (1986). 33. P.R.Trincherini et al, ESARDA Bulletin, No. 10, p. 10, 1989. 34. W.Ruster et al, Nucl. Instr. Methods in Phys. Res., A 281, 547 (1989). 35. C.K.Kirn et al, J. Radioanal. Nucl. Chem., Letters, K36, 353 (1989 ) . 36. K.Komura, Health Physics, 46, 1213 (1984). 37. A.R.Moor Ihy et al, Anal. Chem., 60, 857 A (1988). 38. E.Gantner et al, KfK-4219 (1987)". 39. V. K . Hhrti ga vf. et al, J. Radioanal. Nu<: 1 . (Them., Articles, 132, 179 (I 989) . 40. N.C.Jayadevan et al, Ibid, 82, 125 (1984). IT - 26.9 CURRENT STATUS OF RADIATION CHEMISTRY AND RADIATION PROCESSING IN THE U.S.S.R. A.K.Pikaev Institute of Physical Chemistry of Academy of Sciences of the U.S.S.R., Leninsky Prospect, 31, Moscow 117915 GSP-1, U.S.S.R. ABSTRACT - The paper is a brief review of current status of radiation chemistry and radiation processing in the U.S.S.R. The most important recent achievements in these fields are discussed. The trends of further development are considered. Key words: 7-ray sources, electron accelerators, pulse radiolysis, radiation chemistry, radiation processing. 1. INTRODUCTION The systematic studies in radiation chemistry in the U.S.S.R were started only 45 years ago in connection with arose requirements of nuclear power. During such a short period it was turned into an important division of physical chemistry having an influence on electrochemistry, radiochemistry, chemical kinetics, photochemistry etc. At the end of the fifties the birth of radiation processing took place. At present the studies in radiation chemistry and radiation processing in the U.S.S.R. are performed in the following main directions: mechanism and kinetics of radiation-chemical processes, radiolysis of inorganic and organic substances (inoluding biologically important), radiation chemistry of polymers, radiation polymerization and copolymerization (including graft-polymerization and curing of coatings), radiation modification of materials (mainly polymers), radiation-chemical syntheses based on chain reactions, radiation destruction of polymer materials, radiation sterilization, radiation treatment of food, application of radiation processing for the solution of ecological problems, technological dosimetry, radiation engineering (construction of radiation sources, radiation plants etc.). The present paper is a brief review of current status of radiation chemistry and radiation processing in the U.S.S.R. It is based on the material reported in a number of books (see e.g. [1-51 )• various reviews (see e.g. [6-9]), proceedings of conferences [10-12] and numerous articles. Technological IT - 27.1 dosimetry in the U.S.S.R. was described earlier [:3K 2.RADIATION CHEMISTRY Different sources of ionizing radi.ition are used in radiation chemistry in the U.S.S.R. They are powerful sources of Co 7-radiation, electron accelerators, cyclotrons, research nuclear reactors etc. As an example, Table 1 shows parameters of linear electron accelerators, which are located in the Institute of Physical Chemistry of the U.S.S.R. Academy of Sciences. Table 1. Parameters of linear electron accelerators Type of ac- Energy, Duration of Maximum Maximum dose oelerator MeV pulse, s current in per pulse, pulse, A Gy U-12 5 2.3x10 6 0.2 150 "Elektronika" 8 5x10"^ 15 20-30 The studies are based on modern physical--chemical methods: pulse radiolysis (with optical, conductometric and ESR registration), ESR, positronium etc. Several pulse radiolysis facilities with micro-, nano- and picosecond time resolution are functioning. Picosecond pulse radiolysis facility registers the luminescence of irradiated systems [14]. At present the high-sensitive method of magnetic resonance detected from the yield of reaction including optically detected (OD ESR) is widely applied [15,16]. The teclinique of quantum beats at recombination luminescence of irradiated systems (see e.g. [17]) was developed. The important results on primary products of radiolysis were obtained by means of electron spin echo [18], ESR spectroscopy of high resolution [19] and ESR tomography [20]. The efforts of many researchers are concentrated on the investigations of the formation and properties of primary products of radiolysis of various systems: water, aqueous solutions, organic liquids, polymers, solids, heterogeneous systems etc. For example, recently the formation of solvated electrons (e ) was detected in the case of organic phosphates [21, 22] and melts of fluorides of alkaline metals [23]. According to Ref.[21], the peaks of optical absorption bands of e~ in tri--n-butyl phosphate (198 K) and trimethyl phosphate (233 K) are located at 1580 and 1280 nm, respectively. Tn tri-n-butyl phosphate the yields of free IT - c17 . 2 and geminate ionic pairs at 208 K are equal to 0.84 and 4.05, respectively [22]. As it follows from Ref.[23L optical absorption band of e~g in melts of eutectic mixtures of LiP-KP (62-48 mole %) and LtF-NaF-KF (46.5-11-5-42 mole %) has a peak at 680 nm (half-width of the band is 1.3-1.5 eV). The yield of e~ in these melts is equal to ~2 electrcn/100 eV. Nowadays there is a permanent improvement of computer simulation of primary radiolytical processes. For example, recently a stochastic model of water radiolysis was developed [24]. The calculation of' kinetics of water radiolysis products within the picosecond range* by the Monte-Carlo method was performed. It was found that ion-electron recombination (H..0 + e~) plays an important role not only in hydrocarbons but also in water. At present the special attention is paid to high temperature radiolysis of water and aqueous solutions [25]. It is due to the use of such systems as coolants in nuclear reactors. The respective systems are water, diluted aqueous solutions of ammonia, hydrazine, phosphates, chlorides etc. The other topics under investigation, which are connected with problems of nuclear power. are radiolysis of aqueous solutions of lanthanides and actinides [26,27] and i-adiolysis of extraction systems [28]. The other important direction is the application of pulse radiolysis for the solution of different problems of general chemistry [29]. The last contribution of this method is the thorough study of reactivity of inorganic free radicals (S0~, NO-, Gl2, COl etc.) towards metal ions in aqueous solutions. 3.RADIATION PROCESSING Nowadays in the U.S.S.R. sources of 7-radiation of Co and 137 Cs (to a considerably lesser extent) and electron aooelerators are used in radiation processing. Sources of p-radia- tion of ' Sr +^ Y are sometimes applicable for radiation treatment of semiconductors. Besides, ion beams are employed tor the modification of surfaces of solids. The total number of plants on the basis of 7-radiation sources including research facilities exoeeds 200, more than 30 of them being industrial and pilot plants. The activity of radionuolides in them is varied from a few hundreds of curies to ca. 1 MCi. The total number of electron accelerators, used in radiation IT - 27.3 processing in the U.S.S.R., is more than 50. Their total power is more than 1 MW. The greater part of accelerators have the energy from 0.3 to 2.5 MeV and power from a few kilowatts to 90 kW (see Table 2). At present there are projects to construct the accelerators with a power of 300 and even 500 kW. Linear electron accelerators are mainly used in research work. However, two linear accelerators with an energy of 8 MeV are employed for radiation sterilization. Let me begin the consideration from the pilot and industrial plants with electron accelerators. The largest industrial radiation process in the u.L.G.R is radiation cross-linking of polyethylene. About 20 industrial linus with electron accelerators manufacturing cables and wires with radiation cross-linked polyethylene insulation are working in this country. The total power of the accelerators is ca. 0.4 MW. Moreover, there is a tendency to expand considerably the process under consideration. It is necessary to mention a new self-shielded compact plant for radiation cross-linking of polyethylene and polyvinylchloride insulation of cables. The plant functions on the basis of an ILU-8 electron accelerator. The plant is installed in a common production building. The second by its capacity industrial process is curing of coatings by electron beams. At several TV-plants (in Lvov, Simferopol and others) there are or were during some years commercial lines for radiation curing of paint coatings on wooden parts of TV-sets. Electron accelerators "Elektron" and "Avrora" are or were used as radiation sources. Now the interest in this process has become less, because wooden parts of TV-sets were subsituted by plastic ones. Simultaneously the growth of the application of radiation curing takes place in civil engineering. For example, in 1987 a commercial line for finishing the furniture boards by radiation curing of coatings was put into operation at the Kiev furniture factory. Two "Avrora-2" electron accelerators are used as radiation sources. Curing is carried out in the air. The output of the facility is 7-8 m/min. The third by its capacity radiation process is the production of heat-shrinkable items from polyethylene (films, tubes etc.). In the U.S.S.R. 4 commercial plants are working. For example, since 1983 the industrial facility for production of heat-shrinkable polyethylene insulation tubes on the basjs of two electron IT - 2 7. i* Table 2. Parameters of Soviet industrial accelerators Type Trade Energy, MeV Power, kW mark a>b Eleotron transformer ELT-1.5 1.5 15(25) The same ELT-2.5 2.2 15(20)a Name of Location Activity, Sterilized products sterilizer "Sterilization-?" Leningrad 930 Devices for blood transfusion, disposable syringes "Sterilization-3" Belgorod- Dnestrovsky 360 Devices for blood transfusion, dialyzers of an artificial kidney and others Parcha" Moscow Region 120 Bandaging material "Paket" Kaluga Region 670 Obstetrlo kits "Pintset" Kazan 170 Surgical material "Palata" Tyumen 470 Injection needles It is necessary to emphasize the Increased Interest In the U.S.S.R. to radiation-physical technology with using different types of ionizing radiation: electron and ion beams, 7- and p-radiations. It is related to the modification of semiconducting devioes, ion implantation, deposition of metal layers, hardening }f surfaces etc. The fact, that In the U.S.S.R. 5 facilities with 1 "YT (Cs 7-radiation sources (each having the activity from 5 to 45 kCi) and more than 10 facilities with 90Sr + 90Y p-radiation sources (each Having the aotivity ca. 7 kCi) ar0 used for radiation treatment of semiconducting devices, may serve as an example. To conclude this section 1st me enumerate the most important radiation processes in the U.S.S.R., which have technology tested by means of enlarged laboratory facilities and which are prepared for large scale realization. Suoh processes are radiation-flotation method for purification of aqueous wastes (from the electrolytical production of chlorine) from mercury, radiation disinfection of dung wastes from stock-breeding farms IT - 27.7 and sewage sludges (to be further applied as fertilizers), radiation destruction of cellulose-containing wastes (sawdust, chips, straw etc.) to obtain nutritous additions for fodder, radiation grafting of acrylamide onto polyethylene and silicon rubber for preparation of some medical devices and acrylic monomers onto butadiene-styrene thermoelastoplastio and ethylene-propylene terpolymer to produce oil resistant rubber, radiation-chemical synthesis of organochlorosilanes by vapour condensation of hydrochlorosilanes with halogen substituted alkyls, aryls and heterocycles at 2OO-45O°C. 4.CONCLUSION analysis of current status of radiation chcruistry and radiation processing in the U.S.S.R. allows the following conclusions on the trends of these fields to be made. In radiation chemistry it is possible to expect further progress in the study of initial stages of radiolysis and primary products of radiolytical transfomationB by means of the improvements of experimental technique and computerized calculations. In radiation processing one can expect the expansion of traditional directions: cross-linking of polyethylene and polyvinylchloride cable insulation, onrJjig of coatings on various surfaces, manufacturing of heat-shrinkable polymer products etc. Radiation sterilization of medical products will be further developed. Radiation treatment of food and application of ionizing radiation for the solution of ecological problems will matters of utmost interest. There is also a tendency towards a more intensive development of radiation-physicu. technology. There will still bs a trend to a faster development c. radiation processes on the basis of electron accelerators ae compared to those involving 7-radiation sources. REFERENCES 1. 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All-Union Conference on Pure and Applied Radiation Chemistry (Obninsk, October 1984). Abstracts of Papers, Ed.by A.K.Pikaev, Hauka, Moscow (1984). 12. Second Ail-Union Conference on Pure and Applied Radiation Chemistry (Obninsk, October 199QU Abstracts of Papers» NIIT.EKh.im, Moscow (1990). 13. V.V.Generalova, M.N.Gurslcii and A.K.Pikaev, Radiat.Phys. and Chetn., 3_1. 449 (1988). 14- V.M.Grigor'yants, V.V.Lozovoy, Yu.D.Chernousov, I.V.Shefo- laev, V.A.Arutyunov, O.A.Anisimov and Yu.N.Molin, Pokl.AN SSSR, 299. 1366 (1988). 15. E.L.Frankevich and S.I.Kubarev, Triplet State and ODMR Speotroscopy, Ed.by R.H.Clark, Wiley, N.Y., p.187 (1982). 16. Yu.N.Molin and O.A.Anisimov, Radlat. Phys. and Chem., 21_, 77 (1983). M. A.^.teselov, ^.L.Bizyaev, ^.I.iaele^Khov, O.A.Anislrnov and Yu.N.Molin, Radiat.Phys. and Chem., 34, 567 (1989). 18. K.M.Salikhov, A.G.Semenov and Yu.D.Tsvetkov, Electron Spin Eoho, Nauka, Novosibirsk (1976). 19. O.Ya.CJrinberg, A.A.Dubinskii, V.P.Shuvalov, O.E.Yakimchenko and Ya.S.Lebedev, Dokl.AN SSSR, 230, 884 (1976). 20. O.E.Yakimchenko and Ya.S.Lebedev, Khim.fizika, 2, 445 (1983). 21. V.D.Zaitsev and G.I.Khikin, Khirn.vys.energ., 23, 99 (1989). 22. V.D.Zaitsev, E.L.Protasova and G.I.Khaikin, Khim.vys. energ., 24, 414 (1990). 23. T.N.Zhukova, I.E.Makarov and A.K.Pikaev, Izv.AN SSSR. Ser.khim., 495 (1988). 24. V.Ya. Sukhonosov and I.G.Kaplan, Khim.vys.energ., 24, 404 (1990). IT - 27.9 25. A.K.Pilcaev,, S.A.Kabakchi and I.E.Makarov, High Temperature Radiolysis of Water and Aqueous Solutions, Energoatomizdat, Moscow (1988). 26. A.K.Pikaev, Y.P.Shilov and V.I.Spitsyn, Radio 'y.H.8 of Aqueous Solutions of Lanthanides and Actinides, Nauta.., *oeoow (1983). 27. M.V.Vladimirova, Radiation Chemistry of Actinides. Energoatomizdat, Moscow (1983). 28. G.F.Egorov, Radiation Chemistry of Extraction Systems, Energoatomizdat, Moscow (1986). 29. A.K.Pikaev, S.A.Kabakohi, I.E.Makaraov and A.V.Gogolev, New Trends and Developments in Radiation Chemistry, IAEA, Vienna, p.23 (1989K 30. A.P.Voronin, N.Z.Lyakhov, P.A.Salimov and G.A.Spiridonov, Zhurn.Vsesoyzn.khiin.ob-va im.D.I.Mendeleeva, 35, 72 (1990). IT - 27.10 CHEMICAL EFFECTS OF THE 8 7C1 (n, Y)38C1 NUCLEAR REACTION IN MIXED cis/trans- AND fac/mer-POTASSIUM-HEXAFLUOROCHLORO- OSMATES(IV), K2OsFnCl«-n ( n = 2, 3, 4 ). Horst Muller and P. ObergfelT Institute of Inorganic and Analytical Chemistry, Section Radiochemistry, University of Freiburg, Albertstr. 21, D-7800 Freiburg i. Br., Germany Different reaction channels are possible for ligand recoil 3toms in target substances belonging to the K2PtClo structure. A model "Imp^ct-induced Multiple Ligand Abstraction, IMULA" was developed which considers the different reaction channels 1.) capture at interstitial vacancies 2.) primary retention 3 ) billiard-ball reaction = substitution, which is the most important channel 4.) substitution reaction with additional formation of an extra vacancy which may by populated by adjacent halide ions In a large number of investigations using different target sub- stances it was shown, that those four reaction channels are suffi- cient for a quantitative description of all experimental results including the evaluation of the different reaction channels. When using geometrical isomers of mixed potassiumhexafluoro- chloro-osmates the IMULA model prohibits for S8C1 recoil atoms the formation of fac-OsF338ClCl2z- from trans-K2OsF4Cl2 as well as the formation of cis-OsF238ClCl32- and cis-0sF438C1C1Z- from mer- K?OsFsCl3, all these species, however, are found although in small quantities of 6%, 356, and 1%, respectively. - For that reason it must be assumed that the intermediate species with the additional ligand vacancy, in the case of trans-Ks0sF4CI2 cis- and trans- 0&F238C1C12 • z" (when F ligands are entering the vacancy) and mer-OsF338ClClD2~ (when Cl ligands are entering), are geometri- cally not inert but isomerise by an intra-anionic 1igand-vacancy IT - 28.1 exchange which destroys the geometrical orientation. This investigation was enabled by using specifically designed preparative and separation procedures. The cis/trans- and fac/mer-isomers were prepared by thermal exchange of K*OsF«- KzOsCla mixtures, the analytical separation was performed by HPLC using a specifically designed stationary phase [H. Muller, P. Obergfell, Fresenius Z. Anal. Chem. 3_£&, 242 (1987)]. This investigation was supported by * the Bundesministerium fur Forschung and Technologie. CHEMICAL EFFECTS OF THE 37Cl(n,Y)38Cl NUCLEAR REACTION IN K2SnCle-K2OsBrB MIXED CRYSTALS Horst Muller and I. Hagenlochor Institute of Inorganic and Analytical Chemistry, Section Radiochemistry, University of Freiburg, Albertstr. 21, D-7800 Freiburg i. fir., Germany *X recoil atoms produced by the nuclear reaction X(n,-y)*X in mixed crystals of the type K2AXB-K2BY6 preferentially react in four different ways: (1) primary retention (L3(0) reaction) (2) capture at interstitial vacancies (3) billiard-ball reaction = substitution, which is the most important channel (Lr[1] reaction) (4) substitution reaction with additional formation of an extra vacancy which may be filled by adjacent halide ions (Lf[2] reaction) Reactions 3 and 4 may occur with AXo ions and with BYa ions. Usually one of the components is of suc.i a kind, that the primary retention is preserved upon dissolution, e.g. in the system K2ReBre-K2SnCU (X=Br). The system K2 SnCla -K2OsBr8 is the first example of an investigation in which only the L* substitution products may survive the dissolution. - The following reactions occur IT - 28.2 (1) K2SnCle(n,Y)K2SnC79»«C! (dissolution)—-> >«ci (2) KzSnC1e(n,Y)38Cl —> »«C1 (interstitial ) — > > (dissolution) > »«ci- (3a) »»C1 + OsBre2' > OsBrs^'Cl2- + Br (3b) »«C1 + SnCla2- > SnClss»Cl2- + Cl 2 SnCl3s«Cl - (dissolution) —> »ec]- (4a) a«C7 + OsBra2- — > OsBr4»8ClD2- (intermediate spec.) + 2 Br (4b) 3«C1 + SnCla2- ~-> SnCi43<>Cia2- (intermediate spec.) + 2 Cl (4c) 0sBr43aClO2- + Br > OsBrs38Clz- (4d) SnCl438ClDz- + Br > SnBrCl438Cl2- -(dissolution)--) S«C1- (4e) OsBr43»ClD2- + Cl > OsBr4 C138C12- C4f) SnCl438ClD2- + Cl > SnCls38Cl2- (dissolution) — > 38C1- The reaction products have been determined quantitatively for approximately twenty mixed crystals of different composition from which the portions of the reaction channels could be evaluated: (1 +2): 31 X; (3a + 3b): 44 X; (4c + 4d + 4e + 4f): 25 58. The analytical separation was performed by HPLC using a specifically designed stationary phase [H. Muller, P. Obergfell, I. Hagenlocher, J. Phys. Chem. SSL, 3418 (1986)]. This investigation was supported by the Bundesministerium fur Forschung and Technologie. IT - 28.3 Aspects of Formation and Chemistry of Post troniuia T. DAT? A Radiochemistry Division, Bhabha Atomic Research Centre, Trombay, Bombay - 400 085 SUMMARY : This paper describes the aspects of formation probability, inhibition and enhancement of pO3itronium species, physico-chemical interactions of various types of positronium species, experimental techniques and the major areas of application of positron annihilation spectroscopy* Invest! Positron and Positronium - Positron (e+) is the antiparticle of electron (e~) with same mass Cm«), equal but opposite change (+1e) spin (£t and nearly equal and opposite magnetic moment- A positron either directly annihilates or form a bound t-ate called positroniuia (Ps) with an electron. A Ps - species is called a para-positroniura (p - Ps) or ortho-poaitronium (o-Ps) if the positron and electron spins are antiparallel (sincrlet state) or parallel (triplet state) in ic respectively. Although a Ps atom apparently resembles a Hydrogen atom, the ionization potential of Pa- atom ia 6.8 ev compared to 13.6 ev in H -. atom and the first Bohr - orbit radius of Pe is twice (1.06A) compared to H - atom (0.53 'A) due to difference in the reduced mass sincere - atom is two centered unlike the H - atom. On the other hand, basic differences between p - Ps ani o - PB states are due to the relative variations in epin-alignment of the constituent e+and e~ Table - 1 shows a comparison of the properties of these isomers/2/. Table - 1 Comparison between the properties of p - Ps and o - Ps states* IT - 29.1 p Ps o - Pa • Ionization Potential (ev) 6.8 £.8 • Atomic fiadius ( *A) 1.06 1.06 • Spin 0 1 * Life M.me. (a see.) 0.125 140.0 * Intrinsic Annihilation 2r 3r • Relative Probillty of Formation (%) 25 75 * Magnetic moment (f)<-) 2 2 (for o » state) M-.e magnetic moment 0 (for ± 1 state) Bras, in the event of formation of Ps - species, longer life- time and higher probability (due to higher spin-multiplicity) it ie o - Fs specie that is susceptible to its physico- chemical enviornment. Fundamentals of Posltronlum Formation - Badionuclides in lighter mass regions having proton/neutron ratios slightly in excess of /3 - stability configuration are positron emitters with convenient half-lives. The emitted positrons of energy of few hundreds Key slow down to thermal energy region in less than few tens of picoseconds by inelastic ( ionizing end exciting collisions) with media molecules or atoms. The Ps formation probabilities in different media are interpreted by two major models viz. the 'Ore' model and the 'Spur1 model /3/ ID both the models ©+ - annihilation or Ps formation are assumed to occur following thermal!zation, for thermal!zation time ie much shorter compared to Ps life-times. According to the Ore model considering energy threshold domains, the raotion (?) of positrons forming Ps - species is qlven by t gex - (I - 6.81 / F / 6.8 (1) B ^ ^ I ex with F (p - Pa) - 0.25 F, F (o - Ps) - 0.75 F. Bfccis the energy of the first excited state of the medium molecule with ionization potential I. This model is seen to be qualitatively valid in many oases except when strong affinities of molecules for e /e~ or even Ps necessitate more detailed energy - balance treatement. IT - 29.2 A more realistic and esseutially radiolytic approach is invoked in the 'spur* model/5/. In this model, during slowing down of e+ of energy o- 100 kev to thermal energy in few 10 p.sec,ene»ry deposition l~ 100 ev steps) occurs in steps of* 10 sec in dis- crete aones of *10 A called 'spur1* Around 10,000 spurs separated by-v 300 A containing excited /ionized molecules, atoms, preadicals, electrons of few *V energy annihilate or form the Ps - species. The Pe formation is <»iven by t . P - 1 - exp ( - rc/rr) (2) fc is the 'Onsaa-ar* distance and rr is the mean thermalization distance for positron* The basic intra-epur reactions explaining the formation of Po as well as the inhibition and enhancement of Ps formation pro- babilities in any molecular medium ares a. Basio spur reaotions - • Ps - formation i e4" + e~-* Ps • Geminate recombination, i 15* + e"-* M* • Capture by Prae Hadical i R' + e~-> R" b» Inhibition of Ps formation - • e~ - scavenging t R * + e~->J8*+X~ • e+ - scavenging » R X + e^R X* o. BoheacemQat of Pe formation - • Hole scavenging : h* + X -> X+ * Shallow e" « trapping i ' with -i~ + s->S" + e% Ps +3 e+ - abstraction * Radical scavenging * R -•• X-> RX •* Experimental. Experiments involve common nuclear spectroscopic techniques for both fast timing coincidence and/or hi IT - 29.3 ( T >/t =» 2.7y) source. 9OV. 22-Na "<. A i 2.7y ^» _I22-NIe In the coincidence measurement with a pair of fast response plastic scintillators the 1.28 Mev gamma ray signals birth of a positron while the 511 kev annihilation gamma ray signals end ot the e or Pa life-time. Typical tins resolution is 240 - 300 psec. The coincidence rate is X 1*iexp ( ~ I-t. is intensity of i, the state with decay probability Ai- . Information on various states intensities momentum distribution of the associated electrons, Fermi surface etc* are obtained both by one and two dimension angular correlation of annihilating radiation (1D. ACAR and 2D. - ACAR) studies. In these techniques both the 511 Hev gamraa - lines from position - emitters eg» 64 21 Cu (T"/i_= 12.8 hr), Na etc are measured in coincidence» as a function of mutual angles ( 9 ). The observed antrular devia- tion from the ideally expected 180° directional, correlation between the annihilation warn na"line (due to momentum conservation) arises due to finite momentum of the annihilating pair of e+ - e~ (i.e. of electron actually since e+ is thermalized)* Resolution of these techniques correspond to 0.2 to 0.3 ev of electron energy. Pyt , ^^---To Ae. &~ Pe /me.c A more rapid* much simpler but lower resolution (5-6 times) technique providing same information as I'D - ACAR in Dopvler - broadened annihilation radiation line-shape analysis. In this technique higher sensitivity is obtained after deconvolution of e-xperinentally obtained annihilation (511 kev) sectrc from detector response A B = + Pe . C/2 - (5) Diract measure of 0 - Ps intensity is obtained from triple gamraa- coincidence or from 2r/3r ratio of annihilation radiation. Com- bined measurements of lifetime and angular correlation in coinci- dence is a recent technique providing indepth understanding of the e+ As states in different media. An entirely new technique IT - 29.A of recent origin Is inverse poeitronium formation apectroaeopy providing information en chemical environment in a aample •* Poeitronlum statea and Chemistry of ps - speciesJ- In addition to the p- Pe and o-Ps basic states, Poeltronium exists in different types of states that display characteristic annihilation patterns in condensed matter particularly Bolide/4/. In Metuls or ionic lattices a thermal e+ is squeezed in the Inter— atical si tea away from the positive ions If there is additional space in the form of vacancy, defects or imperfectiona, e+ as well ae Pe (except in mefcals) occupy such sites. Such trapped e* or Pg show lonw lifetime and narrow momontttm component, due to perturbation of pure Ps - states by the lattice electrons it is difficult to ^et the pure Pa fit ate a and one mi^ht be left with q ii ( - Ps) states that have, largeg r radius lower binding energy and different decay probabilitieslti . In perfectft , defect - free crystals at low temperature delooalized q Ps gtateg have been IdentL fied showing very narrow momentum components. These delocalized q - Ps states- have been established also from magnetic quenching of long—lived state into narrow-momentum components. Lt higher temperature or,in presence of defects, scattering of delocalized Ps - apecies at temperature-induced or exist a? defect sites lead to localized apecies/4/. In solids lllong-livei d PfP l stattt e ttrappe d In vacancy cluBtprlt n hhav e alsl o bbeen identified. Further lon^-lfved (~ 500 psec) Pa." states in massive vacancy clusters or' aurface states has also beea identi- fied in oxides. In solutions p - Pe and o-Ps states are observed. Sometimes, depending on solutes fey oomplex or substituted gtateq are obaarved/S/. In. solutio na, e* or Po trapping occurs due to creation of a Bhallow potential well around Ps (Bubble formation) due to coulomic repulsion between Ps - electron and molecular electrons. Bnera-etics, size etc of the bubble depend on the Ps - zero point kineticenergy, liquid surface tension and external pressure (if any)./6/. Table - 2 shows the typical data on lifetime (1O, Angular correlation PWHM (/») and intrinsic energy line-width (Ti) for iaportqnt Pa - states in solids and liquid!a. Ps - state p - Ps (Bloch) 0.12 0,4 (o.V) p - Pe(Trap) 0.12 3-5 1.0 p - Pe (Bloch) 0.5-1.2 1-5 0.3-2.5 Prea 0.4-0.6 5 1-2 p - Pe (Solu.) 0.12 4 1.0 IT - 29.5 o - Ps (Solu.) 1.8 8--10 2.3-2.6 9 I (Solu.) 1-2 5«9 U5-2.5 Fr*e e (Solu.) 0,4-0.6 9.12 2.5-3.0 Solutions in Water - In pure solvents, in the absence of any chemical reactant the quenching madeof Ps involves a physical (colulombic) interaction of Pe with surround electrons leading to polarisation and eventually 2r - annihilation of o - Ps ( & also of p - Ps). Ifce process is called 'Pick-Off* and the rate of pick off is given t, /1,3 A * PO » 4 "re r? C n - 2 eff (6) Where ro is classical electron radius, C is velocity of li^ht, n in atom density and 2^ is effective electron number of the surrounding atoms. In pure solvents the Ps - formation probability is possible to correlate with lonization potential of molecules while the decay probability (%p.o) is empirically seen to be correlated to bubble size, surface tension, viscosity and even parachor in the case of mixed solvents/3/. In the presence of various inorganic or or J * 1 + kc IT - 29.6 V + 1 - V ) (ii) 1 + KC 1 + or c , (J(_J+ i - f> * +0(c 1 + KC 1 +pc (v) 8 is therefore the total Inhibition constant, K is the limited inhibition constant with *f* fraction of Pe liable to limited inhibition; Ji iescaven^in^ effid ency of the enhancer and <\ Jp is enhencement co-effi d ent (lift*'l*)/9/o Temperatu?« dependence/10/ of . Ps - formation probability in various caaea becomes apparent due to temperature dependence ot the rate - constants K as l K » KQ. ezp ( - B/R?) - ( 7 ) particularly in the case of limited inhibitors suggesting a dependence on solvent viscosity. Therefore limited inhibition ie assigned to the localized fraction (f) of Pa in the solvent temperature effect in water is shown in Table - 4, fable - ?» o-Ps intensity and lifetime in common solvents* Solvent H20 Benzene n-Hexane Cyclo— Pyridine C3 Acetoae hexane o - Pa (#) 27.9 43.0 42o4 37.9 14.3 46.7 17.1 ^C (ne) 1*80 3.15 3.92 3.24 2.68 2.12 3.23 .'able - 4. Temperature effect of Pa - formation in water s- ,' (K) t lo-Ps W» X 294 0.28 27.9 1• 80 323 0.32 28.6 1 .75 363 0.33 29.6 1 .72 B. Chemical quenching of o - Pa can occur in multiple wayB/2,3/ eg,» oxidation, spin-conversion, substitution,addition,complex— ation» donor-acceptor interaction and conversion-chemical process as sketched briefly* IT - 29.7 i) Oxidation s Pa +Fe+**-> e+ + Fe^1" ii) Spin-conversion t o-Fs +M (t) ^ p - PB + H {i ) M » No, O-, Ceo , Ma* , Ni**, Fa*'* iii) Substitution : Ps •*• cl -> Ps«l + d iv) Addition s PB + X-*Ps I v) Complexation : o * + X-»(e4tX), l"« Cl^ BrJ I" vi) Conversion chemical » o-Ps + Ac (*)->( Ps Ac (1V ) ) -^ p-Ps + Ac ( \ ) OR 2r 7ii) Donor-acceptor mechacden : Ps + ii ( • C ) -g Ps - M ->2r M: carbonium. These reaction are not as straightforward as sirailar conventional processes /11/ In oxidation reaction for example kinetic effects are more important than the redoxpotential of the ions, la conversion all paramagnetic ions with unpaired 3d electron are quenchers/12/ due to Ibrmation of intermediate complex. Several organic diama^netic quenchers/7/eg., nitroaromatics, conjugated compounds quinones and certain nitriles function via donor- acceptor scheme. In the area of ccmplexation/13/positron forma bound states with cl^ Br? I" , CN~ etc ions but not with F", 0C(i SO" oHr acetate, oxalate ions. Quenchin? reactions alfio take plafce with complex anions/H/eg-. it is seen that with Fe (CM i; Co (CW/-)V 2n (CN4)V cd (CN,H~ o-Ps quenching occurs followiu^ addition across CN- bond vfiile with Fe (CNg)»"oxidation leads to quenching. The rate constants of complexation reaction or generally thequenchinx? reaction are seen to be diffusion - controlled and thus are susceptible to temperature.. •• Applications :- Positron annihilation spectroscopy (P.A.S.) essentially provide information re^ardingr effective electron - density, polar!zability, electron momentum distribution defect / vacancy types and concen- tration, chemical, bonding types of chemical species or transients and also about structural details of the medium of investigation. Very hi^h potential therefore exists for application of PAS techni- ques in multi-disciplinary areas both for fundamental and applied purposeo/1,29 3,4/« in the area of solid state physics and metallur^ry information about the Fermi surface to test various theories, band structure IT - 29.8 description etc are provided by the PA3 technique. Investigations on lattice defects, vacancy formation enlhalpy, entropy etc phase transition annealing behaviour etc are poeaible to carry out UBingj PAS techniques* Even in the aiea of hi^h Tc superconductors indepth investigations have been carried out "by these techniques/ 1, 4» 15, i6/« New information in the area of radiation chemical studies e^.about precursors of solvated electrons where only sui- table techniques are fast pulseradiolyeie and laser photolysis* are avail d?le from PAS investigations. Finite and diroct corre- lation are seen to exist between electron scavenging ability and the Ps - inhibition constant of several solutea Information about triplet photoexcitec* molecular states are available from studies on o-Ps quenching by spin - conversion process, in photo chemical interaction/17/. A very important area of application io investigation on the nature of* chemical bonding usin •* References - 1. Proceedings of International Symposia on Positron Annihilation - (1979, 1982, 1985 and 1988), 2. Positronium and muonium chemistry - Bd. H, J. Ache, (197S) IT - 29.9 3. Postron and Positroaium Chaaistry - Ed. D.M. Schrader and T.C. Jean, (1988). 4. Hositrons in solids- Ed. P. Hautojarvl, (1979). 5. {J. Duplatre, J.C. Abbe, A.T. Haddock and A* Haesaler - J. Chem. Phys. 72, 89, (1988). 6. p«Jansen, Ria ^Report No. - 333, Rist* National lab., Denmark, (1976). 7. V.I. OddansMi and V.P. Shantaro vich - App. Phys. 3J5_, 8. B* Leavay, S.J.s, Lund and O.E. Mo?ensin - Chem. Phys., 48, 97, (1980. _ 9. J.C. Abbe, (J. Duplatre, A. Haeasler and J. Talamonl - Radiat. Phys. Chem 28,19.(1986). 10. J. Talamoni, J.C. Abbe, 'J, Duplatre and A» Haessler - Radiat* Phys. Chen. 20,. 275, 0982) 11. V.M. Byakov and V.I. 'Jrafutin - Radiat. Phys. Chem. 28, 1 12. $. Duplatre, A. Haessler and J.C. Abbe - J« Phys. Chem.89m 1956 (1985) 13. O.E. Mo IT - 29.10 FJ_S5I0N PRODUCED "Mo C.N.DESAI, Radiopharmaceutical Operations Board Of Radiation U Isotope Technology, V.N.P.Marg, Bombay-400 094, India. Now-a-days, pure radionuc1idic preparation plays an important role in diagnostic and therapeutic nuclear medicine. The field of nuclear medicine has becorr.e a complex, multi- disciplinary subject which embraces aspects of Physics, Chemistry, Biology and in the area of diagnostic imaging application of computer science. One of the requirements for the radioisotope preparation to be acceptable for use on patients is the freedom from unwanted radionuc1idic impurities. The radionuclidic impurities will result in undue radiation exposure to the patient and will not contribute significantly to the diagnostic information. The suitability of a radionuolide for a particular medical application will depend upon its availability in a radiochemicaIly pure form, formulated in a compatible medium, its nuclear properties and its chemical properties. In respect of the first of these considerations it is necessary to eliminate any extraneous radiation sources from a material destined for medical use. The radioisotope ""Tc meets with the requirements of an ideal radionuclide suitable for medical application. Despite the fact that other radionuc1ides have also been used in diagnostic Nuclear Medicine procedures '"Tc has good prospects to remain the radioisotope of choice for the same in the future also. Not less than 80% of the .'diagnostic Nuclear Medicine procedures are performsd using ""Tc labelled formulations. "*Tc in the form of sodium pertechnetate is a vitally important starting material and is prepared in Nuclear Medicine Centres using "Ho-"'Tc generator (COW) deploying Good Manufacturing Practices (GMP) coupled with Good (hospital) Radio Pharmacy Practices. (GRPI. The use of ''•Tc in nuclear medicine centres is mainly dependent on the availability of "Mo-""Tc generator or a suitable separation procedure. The radioisotope generator popularly known as "Cow" is based on the column chromatography separation procedure because of the fact that it is convenient to obtain ""To in a very short time in high quality and with less radiation exposure to the operator and the patient. It is estimated that at present about 37 million GBq of * *•Tc sre used in nuclear medicine centres world over. To produce this quantity of ""To, about 19 million GBq of "Mo is required. Despite the fact that "Mo should be produced by (n,f) irradiation of.stable molybdenum target, it has been realized that the fission produced "Mo because of its high specific activity will be of acceptable quality for the preparation of column chromatography type "*Tc generators. The latter have found wide acceptability because of their ease and speed of operation. Thus the importance of fission produced "Mo is due to extensive use of "Mo-"*Te generators for medical use. IT - 30.1 The object of this article is to review the current status and recent history of the technology (mostly chemical and ..hemical engineering) involved in recovering fission produced "Mo from a mixture of fission products and in purifying ''Mo to the exceedingly high degree required for medical applications. The ' ' Ho-' ' •Tc Generator. The generator contains fission produced mo 1ybdenum-99 adsorbed on alumina in a sterilized glass column surrounded by lead or depleted uranium shielding. The ''Mo dacays with a 2.78d half-life to ""Tc which in turn decays with a 6h half-life to "Tc. Elution of the column with saline displaces the TcO. ~ daughters by exchange with chloride but leaves the undecayed ''M0O41' on the column. This system provides the basis for a 2 *"Tc generator. As long as sufficient undecayed ''Mo04 " remains on the column, a sample of * *•TcO«" may be obtained as required • passing an aliquot of eluent through the column. The recent rchnological advances in the development and productionization ' generator technologies have provided the ""Tc sterile ierator which offers a convenient source of sodium l""Tc) rtechnetate in sterile, pyrogen free isotonic saline solution .ady for immediate intravenous administration or for aseptic separation of technetium-99m labelled radiopharmaceuticaIs. The •fnerator device, mainly the column is attached to a reservoir of terile pyrogen-free saline and the *'*Tc pertechnetate is silected in evacuated elution vials periodically. The eluete fulfills the specifications for Sodium fsrtechnetate (""Tel injection prescribed by the United States Pharmacopoeia, the European Pharmacopoeia, and' the BP monograph on sodium pertechnetate C**Tc) injection (fission). The end product namely the eluete of this radionuclida generator is used in the hospital, as the active ingredient for a variety of medicinal produc-is with a variety of diagnostic imaging indications. Such generators are convenient to use and readily transported to clinical laboratories remote from the radionuclide production facility. Production of ''Mo The specific activity of the ''Ho is the deciding factor for enabling the radiopharmaceutica1 manufacturer to prepare ""To generators of convenient size and also the generators of assured quality. The critical need to have high radiochemica1 purity and very high specific activity has a bearing on the means by which the radionuclide is produced. One potential method is by nuclear fission of a heavy element such as IleU. The fission yield of ''Mo is 6% and the process has the unique advantage that the fission process yields radioisotopes of very high specific activities. However, because the process produces a complex mixture of Fission Products (FPs), painstaking separation and purification of the desired radionuclide will be necessary. The problem is simplified somewhat by using a pure target isotope (enrichment greater than 90%). The use of enriched *38U enables the manufacturer to produce ''Mo of acceptable radionuc1 idic purity wherein the absence of transuranics more specifically 23*Pu is guaranteed. Moreover the process chemistry is also simplified because relatively lesser quantity of irradiated target material is required to be handled to produce a given IT - 30.2 quantity of "Ho. Fission in pure 3 3 9 I) is used to prepare "Mo in l03 3 carrier free form, although contamination by Ruf ' * 1 and 1!2Te was a problem in early preparations by this route. In addition to complexation, precipitation and other separation processes, particularly ion exchange and solvent extraction techniques, have now been developed for efficient separation and recovery of the fission products of interest such as "Ho, lsl I and '"Xe. The development of new highly selective processes complemented by remote handling and robotics system will enable the technologists to obtain potentially high recoveries and elevated standards of purities coupled with the economy of continuous operation. The final purification using radiation stable inorganic ion exchangers such as activated alumina, for an ion «*change chromatography is deployed to obtain a high purity "'MoO«J" product. As a consequence of the large amount of research and development work which has been done in recent years in th.- field of fission product recovery and purification, the designer of a fission product plant is faced with an almost bewildering plentitude of proven and workable processes, and it is probable that an advantageous combination of coordination cheristry, precipitation, ion exchange, and solvent extraction technologies will be employed in any integrated plant. The specific activity of the _" Mo is the deciding factor for preparation of convenient size of * * *Tc generator. The fission yield of "Mo is 6%. The fission process has the unique advantage that it yields radioisotopes of very high specific activities. In many instances specific activities obtained are very nearly the theoretical specific activity of carrier free radioisctopes. The specific activity obtained is of the order of 10000 Ci/gram of Molybdenum. Special problems in production of "Ho by Fission: 1. The production of "Mo by fission process requires detailed studies and considerations of the following aspects. 1.1 Target Technology Development 1.2 Current Process Technology 1.3 Problems Associated with Waste Disposal 1.4 Economic Factors 1.5 Proliferation Concerns 1.5. 1 Highly enriched uranium contained in fission "Mo production targets. 1.5.2 Plutonium produced in fission * * Mo production. 1.5.3 Uranium recycling. 1.6 Safeguards 1.7 Quality Assurance'and Quality Control 1.8 Possibilities for Technology Transfer 2. Due to the effect of intense radiation on reagents, column matrix (resins and sorbents) presence of unexpected impurities, existence of fission product species which behave differently from reagent chemicals, etc. the complexities and intricacies involved in the process chemistry tends to be far more complicated. IT -30.3 3. The failure at any stage of unit operation in process chemistry can lead to unwanted delay, radiatio exposures and poor quality product which may not be acceptable from the medical point of view. Additional work, both in the laboratory and hot cell, may be required to discover the cause of the discrepancy, followed by suitable modification of the flowsheet of unit operations. On pilot plant or plant-scale application of the process, further troubles are often encountered - due to process control difficulties or other factors - and additional hot-cell and laboratory work may be necessary before the process is fully perfected. 4. Besides the obvious ones of cost and radiation stability, reagents (such as halides) which would be excessively corrosive to stainless steel plant equipment are to be avoided, as are those which present severe fire or explosive hazards. In this category, oxidizing agents which might volatilize radio-ruthenium, and radio-iodine are generally not used. 5. The degrees of freedom available are severely restricted and much ingenuity is required to develop a simple, economical and practical plant process. Quality of "Ho: The requirements of "Mo for the production of *' * Tc generators are rather very stringent. The specific activity of the "Ho should be carrier free, higher than 185,000 GBq/g at calibration tima, chemical form Mo** in 0.2N NaOH, radiochemical purity greater than 99%. The radionuc1 idic purity is defined as the percentage of activity due to the stated radionuclide ("Mo) and also further .qualified by ensuring the absence of other unacceptable radionuc1 ides. The iadionuc1 idic purity' : I31 I/"Mo -lass than 5. 10-", «» 3 Ru/" Mo- 1 ess than 5.10"8, "Sr/"Mo- less than 6.10-", a'l other fi/r 9mitt8rs/"Mo- less than 1.10"* ('••Tc excluded), total «/"Ho-less than 1.10"'. "specifications for sodium pertec'nnetate " • Tc from fission "Mo. The radioactive concentration defined as radioactivity content per unit volume must be very high preferably of the order of 120-180GBq per milli litre- The chemical impurities which may originate from the process chemicals/reagents and solutions may be carried over to thra final product and they might have det r imsritd i effect on the end use of the final product. It is, therefore, exceedingly important to design the inprocess control during the separation of fission products so that the levels of chemical and radiochemical impurities are contained to be f^r below the acceptable levels. To give an example, trace? of aluminium if present in the final ''•Tc sodium pertechnetate solution may lead to incomplete labelling of biological molecules or abnormal behaviour of the labelled radiopharmaceuticaIs. Availability of T • Mo: At present 3 international suppliers have been producing "Mo in lots of 37000 to .85000 GBq per batch. For this purpose, the fission of uranium is deployed. Most of the present fission "Mo technologies make use of highly enriched 2S'U as Al-U alloys or Udj in a variety of target design and the corresponding chemical separation processes have been developed accordingly. IT - 30.4 References; 1. C.J.Jones: Applications in the Nuclear Fw=i Cycle and Radiopharmacy in comprehensive co-ordination chemistry volume 6 (Applications); Sir Geofry Wilkinson (Editor-in- Chief) Pergamon press, Oxford, pp681-1009; 1987. 2. S.J.Beard and R.L.Moore: Large scale recovery and purification of fission products in Progress in Nuclear Energy Series III, Process Chemistry, Vo1 -4. C.E.Stevenson, E.A.Mason and A.T.Gresky (Eds), Pergamon press, Oxford, pp645-667, 1970. 3. "Nuclear Medicine"; published by Amersham International pic 1990. 4. Fission Molybdenum for medical use; published by IAEA Vienna, 1989; IAEA-TECDOC-515; ISSN 1011-4289. 5. C.K.Sivaramakrishnan, et al; Preparation of High purity fission produced Mo 1ybdenum-99; Radiochemistry Division, BARC, BARC-847; 1976. 6. The United States Pharmacopoeia (USPXXI) and The National Fomulary (NF XVI); p. 1016, 1985. 7. Product Specification of Fission Mo Iybdenum-99; IRE, Be Ig ium. IT - 30.5 NC - Nuclear Chemistry Papers : NC - 01 to NC - 09 2 3 6 EFFECT OF MASS ASYMMETRY ON ANGULAR DISTRIBUTION IN THE U SYSTEM T. Datta, S.V. Dange, H. NaiK and Satya Prakash Radiochemistry Division, Ohabha Atonic Research Centre TROMBAY, BOMUAY-400 085. SOHMAM:- Angular distribution of fission products has been studied as a function of their mass in w^u4oa y'*' system using recoiJ catcher technique and gamma spectroaefry. The angular distribution is seen to be strongly correlated to mass asymmetry irrespective of the effect of multichance fission. [KEY WORDS:-Angular distribution, Mass asymmetry, Multichance fision] 1.INTRODUCTION:- Investigations on fission fragment angular distribution provide information about the fissioning nucleus at the second saddle point in the deformation energy surface. The distribution of the t&e ptojection(K) of the total spin(.T) on the fission axis, with variance K , is generally assumed to be frozen at the second saddle governing the eventual angular distribution. This aspect is useful to answer the question whether mass distribution is also frozen at the saddle since in that case strong dependence of K on mass asymmetry is expected. Limited data on mass resolved angular distribution of fragments seem to suggest such a strong correlation although the effect of multichance fission might affect such observations\1\. In view of these important questions, present work has been carried out on the angular distribution or 12 pass asymmetric fragments in tlie 40 MeV alpha particle induced fission of 0. 1J.EXPERIMENTAL: - Targets of II (-100-200 ( j i i ;i i 2 ;j i i ' * ;i and W{0)/W(90)]. -- 1+(b/a). - 1+ NC - 01.1 yields of the asymmetric and symmetric fragments will depend on it. As a consequence in the case of gross (Bass averaged) angular distribution there must be strong dependence of gross anisotropy on multichance fission. In the case of observations on the asymmetric and symmetric fragments however the effect of amltichance fission is not expected due to (a) 1 , values for symmetric and asymmetric shapes for subsequent nuclei 'are not significantly different (b)The ) . -values do not vary strongly with T or .T in the concerned region and (c) S and consequently K-distribution (K ) are decided in the compound nucleus (' Pu presently) and are expected to reaain nearly unchanged with pre-fission (low energy) neutron emission leading to chance-fission nuclei. The 1 .-value for symmetry was taken for •Pu from the calculation of Cohen ind Swiatecky\2\ while for the asymmetric shape 1^** value was evaluated from observed gross anisotropy in 17 MeV alpha-induced fission of " 0 (C.N. l'u) where multichance fission is. negligible pd symmetric fragment yield is much less. The K -values thus deduced (K..,o" syn>= 124.8 h and KQ asym-(.5.(t n ) are seen to be ia reasonable agreement with the experimentally pbser» -i values. Agreement was also seen for alpha-induced fission of 0 several energies\3\ confirming the considerations given above rath i the effect of muitichance fission. 1V.ACRW0WI.EDGKMENT:- Authors are thankful to Cr.i tf.Nataiajan, Bead, Kadiochemistry Oivi.soa. for his keen interest and encouragement. They acknowledge the assistance provided by Shri K. Guiin and the VKCC crew. V.RKFKRKNCES:- 1. H.Kudo,7..Nagame,H.Nakahara,K.Miyano and l.Kohno, Phys.UeV. C-25, 909 (1982); S.n.Manohar,T.I)atta,A.Goswami and SatyaPrakash, JAKA Syatp. on Phys. and Chem.of Fission, GKD. (1988). 2. S.Cohen and W.J.Swiatecky, Ann.Phys. 22, 465 (1963). 3. R.D.Leachman and L.C1umberg, Phys.MeV. B-137, 814 (1965). TABLK.1- EXPKKlMENtAL DATA ON ANISOTHOPY 2*2 Pu 2 2 2 2 Huclide Anisotropy K0 <» ) Nuclide Anisotropy K0 (li ) 95Zr 1 905+0 060 58.8+3.0 I12l>d 1.462(0 058 114 .0+10.0 S9Mo 1 789+0 087 67.5+6.0 1'J Cd 1.512+0 097 104 .0+14.0 10 I<1 W 1 V36+O 044 72.0+4.0 1.608+0 079 88.0+10.0 1 892+0 045 59.8+2.0 127 Sb 1.586+0 087 90.0+ 11.0 HI Te 1 88!»+0.040 60.012.0 13SCS 1 649+0 0!» 9 82.0+7.0 1 712+0 056 75.0+6.0 U1Ce 1 697+0 081 76.0+9.0 K 2-A- 68.9+8 2(ft2) K 2S: 99.0+ 1O.6(I12) K 0O2 7-A-C.a.l 65. 8 (Yi) K02 -S-Cal 124.8 (n2) o 0 NC - 01.2 HALF-LIVES OF PLUTONIUM ISOTOPES BY MASS SPECTR0M2TRY AND ALPHfc SPECTROMETKY S.K.Aggarwal, A.K.Parab, S.A.Chitambar and H.C.Jain fuel Chemistry Division, B.A.K.C., Bombay-400 085 94.U 2A2. SUMMARY: Half-life of Pu determined using Pu as a reference isotope by the relative activity method iSggiven.,.,Data obtained previously for the a decay half-lives of ' Pu, '"Pu and 3 decay half-life of Pu determined using thermal ionisation mass apectrometry and alpha spectrometry are also presented and compared with the values recommended by IUPAC in 1989. (KEY-WORDS: Pu isotopes, half-life, mass spectrometry,alpha spectrometry) I. INTRODUCTION: Precise and accurate values of the half-lives of Plutonium isotopes are of great importance in nuclear technology. During the last one decade, we have carried out accurate determination of the half-lives of different piutonium isotopes. Thermal ionisation mass spectrometry (.TIMS; and alpha spectrometry were employed for this purpose. Experiments were designed to eliminate the limitations of previously published data with a view to resolving the existing discrepancies. This paper summarises the results putjiished^ .earlier f,ro,m our laboratory for the half-lives of *" Pu, Pu and.,.*' Pu and presents data obtained recently on the half-life of Pu. UL EXPERIMENTAL: The half-live.', of 238Pu, 240Pu and 242Pu were determined by relative activity method. Synthetic mixtures using solutions of enriched isotopes were prepared using one of the Plutonium isotopes as a reference nuclide. The double dilution methodology was employed to maintain the atom ratios and the alpha.1activity ratios in the mixtures close to unity. Half-life of Pu was determined by t"he parent decay as well as the daughter growth method. In the parent decay method, a synthetic mixture of piutonium isotopes with different isotope ratios close to unity was prepared and used. In addition, an isotopic reference material SRM-947 was also used. The half-life was computed using the double ratio methodology to minimise the errors arising due to changes in the isotope fractionation pattern. In the daughter growth method, Am ingrowth was determined by isotope dilution mass spectrometry, isotope dilution alpha spectrometry and alpha proportional counting. Tail contribution at low energy peak due to energy degradation of high energy peak was accounted for by using a method based on the geometric progression decrease of the counts in the far tail of the alpha spectrum. RESULTS AMD. DISCUSSION: Table 1 presents the data obtained in our laboratory, the range of published values by different international laboratories and the values recommended by the International^ Union of Pure and Applied Chemistry iIUPAC; Commission on Radiochemistry and Nuclear Techniques in 1989 [lj. The data published previously from our laboratory have been NC - 02.1 included by IUPAC while recommending the half-life values.^.The discrepancy of aJ34Ut 6% existing in tne half-life values of Pu using Pu and Fu as reference isotopes was resolved 12 J. The potently' of relative activity method with double dilution methodology was demonstrated by determining the half - life..,, of "°Pu earlier L3] and ^*uPu now. The half-life of Pu determined is in good agreement with the IUPAC recommended value. But the helf-life oi Pu obtained in the present work is 0.8% lower than the IUPAC recommended value. This suggests further investigations and may be a re-determination of Pu half-life used as a reference nuclide need ba done in view of the high precision achievable with the present generation ma3s spectrometers and a better understanding of the tail contribution. The half-lives of Pu determined by the two independent approaches [4-9] are in good agreement with each other as well as *7ith the IUPAC recommended value. Relative activity method with double dilution methodology can24also be employed for determining accurately tho half-life of Pu and resolving a difference of 1.6S& in the two values published LI] in 1966 and 1969. IV. Afyqnwr.KOGEMEMTS: The authors express their sincere thanks to Dr D.D.Sood, Head of the Fuel Chemistry Division for hia keen interest in the work. 3L». REFERENCES: 1. N.E.Holden, Pure and Appl. Chem., £1, 1483 (1989). 2. S.K.Aggarwal, S.N.Acharya, A.R.Parab and H.C.Jain, Phys. Rev., £2£, 1135 (1979). 3. S.K.Aggarwal, A.V.Jadhav, S.A.Chitambar, K.Raghuraman, S.N.Acharya, A.R.Parab, C.K.Sivaramakrishnan and H.C.Jain, Radiochem. Radioanal. Lett., 4Jj, 69 (1981). S.K.Aggarwal and H.C.Jain, Phys. Rev.,CJLi> 2033 (1980). S.K.Aggarwal, S.N.Acharya, A.R.Parab and H.C.Jain, Phys. Rev., ££3, 1748 (1981). S.K. Aggarwal, S.N.Aeharya, A.K.t'arab and H.C.Jain, Radiochim. Acta, £R, 65 U981). S.K.Aggarwal, S.A.Chitambar, A.R.Parab and H.C.Jain, Radiochem. Radioanal. Lett., £4., 83 (1982). 8 S.K.Aggarwal, A.R.Parab, S.A.Chitambar and H.C.Jain, Phys. Rev., £3_i, 1885 (1985). S.K.Aggarwal and H.C.Jain, J. Radioanal. Nucl. Chem.-Articles, ILLS, 183 (1987). TABLE 1. Half-lives of Plutonium Isotopes Isotope Determined at Range of published Recommended FCD, BARC (yr) values (yr) by IUPAC (yr) 238Pu 87.98 ±0.25 8b to 89.59 87.7 + 0.1 240 6505 to 6610 6560 + 10 241Pu 6507 ± 10 „ tf Pu 14.36 ± 0.06 14.29 to 14.60 14.4 ± 0.1 14.43 ± 0.08** 242Pu (3.75 ± 0.03)xl0* (3.6b to 3.85)xlO{ (3.7b + 0.02>x Parent decay, Daughter growth, Values published after 1974 NC - 02.2 POSITROM ANNIHILATION STUDIES IN URANIUM OXIDES T.Datta, P.K.Pujari, A.K.Chaddha * and Satya Prakash. Radiochemistry Division, * Fuel chemistry Division. B.A.R.C, Bombay 400 085. Summary: Positron annihilation studies have been carried out in hyper-stoichiomet-ric uranium oxides using Doppler broadened lineshape analysis technique. It is seen that in these oxides formation probability of Ps states is strongly correlated to effective anion density over and above particle size effect rather than usual 0/M ratios. Results are. discussed. Introduction: Positron annihilation studies in chemical matrices provide information icjarding formation of various positron or positroniusn (Ps) states, their quenching modes and momentum distribution of annihilating electrons - all depending on the properties of the chemical matrix. The properties that can be investigated ara therefore anion density, electron polarisibility, aspects of defects and vacancy traps, particle size - free volume tifects, presence of redox, paramagnetic or adduct-forming species etc apart from electron momentum density distribution as required for fundamental structural investigation. In view of the rich potential of these techniques it is worth while to investigate positron/Ps annihilation in various types of compounds such as in binary oxides of non- stoichiometric nature. In this paper, work carried out on various uranium oxides UO2, U02.15, U4O9, U3O7 and U3O8 using Doppler broadened annihilation radiation lineshape analysis technique is described. Experimental: The oxides U02+x were prepared from starting material U02.15 by heating in air at various temperatures while UO2 was made heating UO3 in hydrogen atmosphere and was sintered subsequently. Structural confirmation was obtained from X-ray diffraction in each case; surface area was measured in each case by BET method. Annihilation spectra for each oxide in powder form were acquired using a 64cu positron source in the form of a strand with a high resolution {1.0 KeV at till KeV) HPGe detector system coupled to a 4K MCA at channel resolution 58 eV per channel. The annihilation spectrum for each oxide sample, acquired twice, was deconvoluted for detector responce deduced from pure gamma line of l-O&Ru. The deconvoluted intrinsic spectra wero resolved for various positrn or Ps states using standard PAACFIT code. Table-1 shows the intensities and widths (KeV) of the various states along with the variance of fit indicating goodness of the fit. Results and discussion: In all the oxide (UG2+x) three components were observed. The first component of wxdth ("l " 0.7 KeV has relatively low intensity " 2% compared to a second component of vidth f2 "2.2 KeV with intensity " 10% while the third component is widest, T3 " 2.7 KeV having highest intensity 13. It is so far established that in oxides M\ and particularly UO2 \2\ formation cf Ps species occur. It is also known that 1 low intensity, long lifetime component due to surface Ps state forms in UO2 \2\. The first and second component observed in the present case are therefore P-P,s and O-Ps states in all probability while the remainder of the intensity is accounted for the widest third NC - 03.1 component arising due to the free positron annihilation, r3 of "2.7 KeV closely corresponds to FWHM of angular correlation {r© ) of "10 mrad for free positron annihilation, iince the present set of compounds are powders with defects an'l vacancies delocalised or Bloch states Ps that has very narrow momentum width or TE <0.3 KeV, are not expected and not seen. Farther, the Ps species observed may not be pure but quasi-Ps (q-Ps) ( both para and ortho varieties) species due to finite overlap of electrons from surrounding matrix with positron or electron of Ps species. The narrow q-Ps (para) species corresponds to localised staes in traps of typical dimension 3-4 A. Table-1 also shows the d calculated effective anion density (n .1022/Cc) and particle size, n* was calculated as \3\ where n= ti$/h where N: Avogadro no., A= mol. wt. , -f- density ,->>+.: no. of cation or anion per molecule, R+is cation radius. It is seen that both II and 12 for q-Ps states increase with n* or anion density where lower particle size further favours the observed q- P3 fraction ( due to larger free volume). These correlations strongly indicate that it is itinerant than anion state-localised positron that give rise to q-Ps formation as seen by Bertolaccini et al \3\ while mere O/M ratio does not reflect the same trend as anion density. The momentum component for O-Ps in sintered UO2 shows width nearly same as that ror positron due to almost zero escape probability from bulk. For para q-Ps the momentum width shows increasing trend with effective an?on density perhaps due to decreasing free volume with increaseing anion density. TABLE 1 UO2.00 UO2.15 U4O9 030? U3O8 R(A) sintered 976 1108 1215 999 (1022 5.40 5.77 6..08 5.94 5.13 /cc) ri (KeV) .6691.019 7721.016 .797*..01 .776*.004 .7001.011 H % 1.83*.13 2 .311.23 2. ?.2±..18 2..101.14 1.661.04 r2 (KeV) 2.6031.01 2.34*.03 2. 231..04 2..201.005 2.1510.03 12 * 9.33*.06 10.831.38 10 .44d 10.301.57 8.341.04 T3 (KeV) 2.71.01 2.711.015 2. 77*. 01 2. 81.01 2.761.01 References: 1) P.Sen et al, Wuovo Cimento fi 64, 324 (1969). 2) D.D.Upadhyaya, Ph. D. thesis, 1982. 3) M.Bertolaccini et al, J of Phys.,C4, 734 (1971). NL - 03.2 INITIAL RETENTION, ISOCHRONAL ANNEALING AND THERMAL ANALYSTS - A CORRELATION Dedgaonkar, V.G. Departmwt of Chemistry, University of Pcona, Pune 411 007 Apte, R.G. and Bhagwat3 D.A- S.P.College, Pune 411 030 Keywords : Lanthanide bromates / retention / thermal analysis ABSTRACT : Cation effect on initial retention and recovery after heating in ten different lanthanide bromates is investi- gated. Difference in weight loss after decomposition seems to be more significant. INTRODUCTION : Rare earth cation bromates are thermally more stable than p-block cation bromate like AlCBrOo^'S^O and show much lower initial retention value and slower annealing than the d-block cation bromates (!)• Studies on f-block bromates (2) enabled to locate retention trends for various periodic groups. A correlation is sought in this paper between the retention and weight loss behaviour of several f-block cation bromates upon heat treatment EXPERIMENTAL : After freshly preparing (lag; the bromates 252 were subjected to^ neutronfintrftnn activatioHfitivatir.nn withh 252nCff fissiofiqainnr neutron source (10^ n cm"2 s~*-) • Separation of activities in the Br~ and BrO^ forms was carried out by solvent extraction and thermograms were recovered by MOM derivatograph. RESULTS AND DISCUSSION : Following fact may be noticed from the data presented in the table (i) the initial retention values in these f-block bromates are intermediate between those of s-block and d-block broraates. (ii) Nanohydrates show lower retention values with respect to ^OBr, OOIHQT an(j 82j3r than hexahydrates but more recovery on heating. This is perhaps due to greater ease of dehydration on heating as compare to the hexahydrates. (ill) In general, tetrahydrate3 and lower hydrates show lower retention and recovery than hexahydrates. (iv) The f-block systems decompose to form stable oxybromide; these show lower retention values than d-block bromates those form oxide. This is because percentage of oxygenated recoil fragments is maximum in the bromates forming oxide and is minimum for those form bromide, (v) In cases where total loss is more than that calculated for decomposition giving OBr, 3ome oxybromide decomposes into oxide; hence higher retention for higher weight loss in TG. (vi) Lanthanide contraction, lattice energy (3), variable oxidation states and third ionization potential of the M+3 lanthanide do not contribute significantly to the cation effect, (vii) Water molecules may have opposite effects on initial retention and recovery on heating. Perhaps dehydra- tion increases recoil separation in hot zone but decreases it during thermal annealing. NC - 04.1 REFERENCES : 1. Dedgaonkar, V.G. and Bhagwat, D.A. , 13th International Symp., Mount Fuzi, Japan, p. 53 (1987). 2. Dedgaonkar, V.G., Apte, R.G. and Bhagwat, D.A., Radiochem. and Radia. Chem. Syrap., Nagpur, RE 26-1 (1990;, 3. Kapastinskii, A.F-, Quart. Rev. (London), 10, 283 (1956) Systems Retention % Annea- Weight loss Lattice ling energy / 1 -0 kJ mol' u 1 u u cP-i 0 « 0 0 00 CO •-! CO Tota l tio n Dehydra - 1- 1 La(BrO3)3.2a2O 21 22 28 11 61 - Pr(BrOo),*6H«0 23 23 30 25 6 80 4205 23 23 30 39 10 71 4124 Sra(BrO3)3*9H2O 19 21 29 81 15.5 69 4151 Gd{Br03)3-9H20 19 23. 29 59 10.5 66 4142 Gd(Br03)3-4Hg0 22 23 31 14 1 72 - Dy(Br03)3-9Hg0 22 25 31 66 13 71 4137 Er(Br03)3'9H20 22 23 31 58 12 72 4123 Er(BrO3)3'4H2O 20 21 27 14 6 65 — NC - 04.2 PREPARATION OF A MULTITRACER SOLUTION FROM GOLD ! OIL BOMBARDED BY 95 H&VS NWCLEOJJ 4°Ar IONS IN RING CYCLOTRON F.Ambe. S.Ambe, M. Iwamoto. Y. Ohkubo, S.Y.Chen and A. N. Garg Nuclear Chemistry Laborarory, Institute of Physical and Chemical Research CRIKEN} , Watcoshi . Saltama 351-Ol . JAPAN. •» Chemistry Department. Nagpur University, NAGPUR-44O OlO. INDIA. SUMMARY: Simultaneous use of a number of radioactive tracers is useful for the precise determination of characteristic behaviour of different element under identical experimental conditions. 2OQand 1OO pM thick gold foils were irradiated with 93 MfV/nucleon Ar beam in Ring cyclotron employing falling bali irradiation system. A number of radionuciides are formed primarily by sp&llation raection of energetic heavy ion beam. A radiochemical method has been developed for the separation of radionuciides. CKeywords; Multitracer solution , Heavy ion reaction. Gold target, Argon~40 beam , Falling bali irradiation system D I.INTRODUCTION Radioactive tracer technique is widely used in various fields of science, technology and medicine. Multitracers have additional advantage that one can acquire data on several elements simultaneously and precisely determine their characteristic behaviour under similar experimental conditions. We have established a radiochemical procedure for the separation of carrier fr&& radioactive multltracer solution from gold target irradiated with 95 MeV/nucleon Argon-40 beam in Ring Cyclotron. II.EXPERIMENTAL Gold foils of 2O and 100 /uM thickness and 2.5 cm diameter were mounted on aluminium balls and bombarded with 95 MeV/nucleon Ar beam employing falling ball irradiation system /I/ in Ring Cyclotron of RIKEN.The beam intensity was 2OnA and th* irradiation time was 3Omin. After about 3 week delay the gold foils were heated to melt at 11OO°C under vacuum CO.O3-O. O4 Torr)and the radionuclide activities were collected by cold finger technique. Two traps with active charcoal, one bulb type and another U-tube type in liquid nitrogen, were used for further collection of gaseous; products. More than 7O>S of the activity is collected in this manner. Later the cold finger was washed for 3O minutes each with warm water C50 °O and aquaregia. The two washings contain a large variety of radioactive nuclides produced from the gold target. The gamma ray spectra of the gold foil before heating Cafter interposing 3mm thick acrylic sheet to cutoff ft- activity} and the two washing solutions were measured with a high purity Ge detector CEG & G ORTEO of 18?4 relative efficiency and 4 K MCA. The complex spectra were analysed by means of programme BOB on FACOM M78O computer. III. RESUl TS AND DISCUSSI ON A typical gamma ray spectrum of the irradiated gold foil after 2O days delay but prior; to heating is given in Fig.l. The dominant peaks are NC - 05.1 133, _ 183_ 16O 177 75 due to Ba. 146 Eu. Ce, Ta, Tb, Lu, Se, 46.'Sc and so on. In water fraction of the washing primary photopeaks 131 133. 16O 75 due to Ba. Ta, Tb, and Se have been observed leaving behind majority number of radionuclides in the aquaregia fraction. The kind of elements traceable by the nuclldes in the solution depends on time after the irradiation because radionuclides of different half lives are produced. The wi d<& range- of mass number of the radi onucl ides suggest fragmentation of the target by heavy ion bombardment. In this process several heavy CA = 124 - 183:' as well as lighter CA = 46 - 7SD fragments are formed . In earlier experiments carried out at RIKEN laboratory, target of Al , Cu and Ag foils were irradiated with 26 MeV/nucleon Ar beam. Adsorption isotherms of a number of elements were determined simultaneously under the same experimental conditions /2/. In the present experiment, we have been successful in obtaining a multitracer solution in carrier fr&& state by following the radiochemical procedure. This has particular advantage In solving a variety of chemical problems. Its application to a model experiment on marine chemistry is in progress. IV. REFERENCES: 1 . M. Yanokura. Y.Ohkubo, S. Ambe. M.Iwamoto and F.Ambe, RIKEN. Accel. Progr. Rep. 22, ISO Cl9885. 2. S. Ambe, Y.Ohkubo, M. Iwamoto, Y.Kobayashi. M. Yanokura and F. Ajnbe. RIKEN , Accel. Irogr. Rep. 23, 79 C1989D S00 1000 1500 2000 2 M0 3000 3S00 /.000 Channel No. ->• NC - 05.1 CHARGE DISTRIBUTION IN ALPHA PARTICLE INDUCED FISSION OF 209-Bi S.S.Rattan, A.Ramasvami, R.J.Singh and Satya Prakash Radiochemistry Division, Bhabha Atomic Research Centre, Bombay-400 085, India. SUMMARY The charge distribution in the alpha particle induced fission of 209-Bi has been studied at alpha particle enerOT of §5-7 MeV and 58.6 MeV. The fractional cumulative yield of -"Zr, Mo, Mo and P-1 have been determined using gamma ray spectrometry. Tne charge distribution have been found to be broad. KEY WORDS: 209-r III. RESULTS AND DISCUSSION: The peak area (A) of the gamma peak of the the daughter product in the required mass chain can be relaced to the cumulative number of atoms of parent (N^Q) formed ar>d independent number of atoms of daughter (^Q) formed by the following equation: Y(A,T,t) - £ N10 X«T,t) + e N20 Where, T and t are the irradiation time and decay time respectively, t is the efficiency of the detector for the required gamma ray. The details of the equation are given else where[3]. The calculated X and Y values were fitted t NC - 06.2 HELIUH-ION-INDUCED FISSION EXCITATION FUNCTION OF ERBIUft R.H. Iyer, R.C. Shama. P.C. Kalsi and A.K. Pandey Radiochemistry Division BARC, Trombay, Bombay-85. SUflliARY The fission excitation function of Natural Erbium (Z»68l induced by Helium ions in the energy ranje 30-65 BeV has been detenined experimentally. The data vere analysed in tens of the statistical todel expression for rf/rn, the ratio of the fission width to neutron Mission width to get the fission barrier snd the level density parameters. Key words: Fission excitation functions, Erbium, fission barriers, level density parameters, lighter eletents. i. INTRODUCTION In continuation of our systematic prograiee of work on the fission properties of lighter eleienti (2<80), we have measured the fission cos* sections of Erbiui (2=68) induced by Helium ions in the energy range 30-65 MeV from the Variable Energy Cyclotron, Calcutta. Low 2 fission systeis provide excellent opportunities, for determining important nuclear parameters such as fission barriers lEf), level density parameters for neutron emission (an) and fission la;), influence of shell effects etc. and for comparing these parameters with predictions of theoretical models ll!. Reliable and accurate measurements of fission barriers (which are equivalent to measurement of nuclear masses at the distorted'saddle point configurations) in the lighter elements not only help In the undemanding of the systematic: of fission process in general but also would enhance our understanding of the systematics of nuclear masses in general. II. EXPERIMENTAL 1 Experimental details are given in some of our publications'^ ancj ar9 gjV(n |,er9 briefly. Targets were prepared from highly purified Erbium oxide on high purity silver foils in the form of deposits 1-2 mg/cm^ thickness by an electrophoresis technique. Fission events recoiling in the backward direction were recorded using_ * cylindrical Lexan detector. Typical He-ion beam currents were 1-2 uAh per experiment. After irradiation, the lexan detector was etched in 6H NaOH. The absolute fission cross sections were calculated by comparing the number of fission tracks in a specified area (observation solid angle) in the detector from the sample (Erbiuml with the number of tracks in the sane area from a standard irradiated under the same experimental conditions. Be used Holmium and Cold as standards whose absolute cross sections as a function of energy are known '2.31, III. RESULTS AND DISCUSSION The experimental fission excitation function is shown in figure 1. The excitation energies were calculated by assuming full momentum transfer and a weighted average 9 value of -1.4 rleV(*', The excitation functions were analysed in terns of the statistical model expression l2'. a £2af'(E-E )*-l] r /r K n f •f'«R " f n '• ° 573 etp[2a{*IE-Ef)*-2an*(E-Bni*] where tf and OR represent fission and total reaction cross section! respectively. E is the excitation energy, Bn is the neutron binding energy. A Is tin mass number of the compounp d nucleus, Ko o is a constant I taken tat 010..7 K tKtV 1(1 1 IS)'* nt9i i Mdd Othth" r tert " " ddff*ffMdd «'rM(JJ)'. «g MS calculated using the optical model of Huizenga and l5) lgo . A least square fitting procedure was used to fit the experimental iyrn values to the theoretical equation to git the 'but fit' values for Efl a{, an and a{/an. The fitted curves corresponding to t{ • A/8 and A/20 (reasonable upper and lover limits) are shewn In figure 2. The 'beit fit' values were chosen on the basis of the «um, of squares ofvthe deviations. Thus, we assign a value of 27.813 fleV for the fiision barrier of the NC - 07.1 171 3 compound nucleus Yb - foraed by the interaction of Erbiua (Nat. J with He-ions. The corresponding values of af and a are 13 18 BeV"1 and 13.05 Me*"1 respectively. The uncertainty shown in the fission barrier muld allow for Inclusion of other values derived froa reasonable upper and lower Units of af and an values. The experimental value coipares very well with the theoretical value of 31.3 HeV (weighted average value! based on a siaple liquid drop «odel lass formation ul. The present work represents one of the lightest compound nuclear systems for which fission barrier has been deteriined experi»entally and extends the range of elenents studied and brings out ssmal systeaatic trends in low Z-syste«s(°\ IV. REFERENCES 1. R. Vandenbosch and J.R. Huizenga, Nuclear Fjssion. Acadeiic Press, H.I. 119731. 2. P.C. Kalsi, fi.C. Shar»a, A.K. Pandey and R.H. Iyer, Paper R-l, Radioche»istry and Radiation Cheiistry Syip. IGCAR. Kalpakkai, Jan. 4-7, 1989. 3. R.H. Iyer, P.C. Kalsi, A.K. Pandey and R.C. Shana, Paper 0-38, Nucl. Phys. Sjrip. AH«, Aligarh, Dec. 26-30, 1989. 4. U.D. Myers and D.J. Swiatecki, UCRL - Report No. 11960 (1965). 5. J.R. Huizenga and GJ. Igo, Hue I. Phys. 29, 462 (1962). 6. R.H. Iyer, A.K. Pandey, P.C. Kalsi and R.C. Shana (to be published). Figure 1 Figure 2 11 I I I | I I I I | I I I I | I I I I | I: fcl I I I | T M I | I I i I ] I I T I | 1=1 (Nat ) • Ho4 1.0E-31 N U •- I.OE-32. =1 I.OE-33- ) QE-3-1' i i i i I i i i i I i i i i I i i i i I i I.OE-IO i i i i 1 i i i i 1 i i i i I i i i i I i 30 40 50 BO 79 30 40 50 60* 70 OClTftTION ENERGY (feV) EXCffftriON ENERGY NC - 07.2 JSOMERIC CROSS SECTION RATIOS IN 12c INDUCED REACTIONS ON B.S.Tomar, A. Goswami, S.K, Das, A.V.R. Reddy, S.B. llano^ar and Satya Prakash, Radiochemistry Division, B.A.R.C, BOMBAY - 400085, INDIA. SUMMARY: Isomeric cross-section ratios(ICR) have been determined for 99,97R.n and 9.'5,94TC produced in the 12c induced reaction on 89'/. The ICR values increase with the increase in projectile energy'from 43 to 73.b^eV. Further the a emission products show lower ICR values than neutron emission products indicating larger angular momentum carried by a particles. (Key words :- Heavy ion reactions, 12c beam, 89Y target, isomeric cross section ratios ) Heavy ion reactions provide an ideal tool for studying various nuclear reaction mechanisms owing to the wide range of excitation energy, angular momentum and entrance channel mass asymmetry. In the light heavy ion induced reactions the most predominant mechanism at projectile energy <10 MeV/amu is the compound nucleus mechanism and is called complete fusion (CF) . However as the bombarding energy increases fusion is preceded by with the emission of few particles from projectile or target/1/. The evidence for such incomplete fusion (ICF) has been obtained from the experiments on -- (i) velocity distributions of residual nuclei/2/, (ii) folding angle of the two fission fregments/3/ and (iii) particle gamma coincidence studies/4/. It ha;s been proposeci/1/ that the ICF occurs due to the disapearance of the pocket in the one dimensional potential energy as the angular momentum is increased. In order to reduce the effective angular momentum of the composite system, an increasing fraction of the projectiles escapes and carries away some angular momentum, thus fecilitating fusion. This should reflect in the isomeric cross section ratios of the evaporation residues (ERi. Isomeric cross section ratios are found to increase with the increasing angular momentum of the compound nucleus(CH). When ICF reactions start dominating the isomeric cross section ratios are expected to either saturate or decrease with increasing projectile energy. However no such attempt has been made so .far in this direction. With this in view we have measured the isomeric cross sectiori ratios of some ERs namely 99Rh, 97Rh, 95Tc and 94TC produced in 12c induced reactions on 89y at six projectile energies varying from 4 3 to 7 3.5 MeV. The experiments were carried out at the BARC - TIFR pellecron facility. Self supporting yttrium metal foils of thickness 2.2 mg/cm2 were bombarded with 12c beams having beam intensity of loll particles/a. Two yttrium metal foils separated by Al degraders of 2 mg/cji»2 were kept in a stack. The incident beam energy was varied by choosing a particular charge state and/or terminal voltage. Three irradiations of 45 minutes were carried out. The target later was counted for the gamma activities of ERs on a precalibrated HPGe detector coupled to 4k MCA. The activities were followed for a suitable length of time and were used to obtain the activities of the individual nuclei at the end of irradiation [A(Ti)J. In the case of common gamma lines the individual components were resolved by least square fitting of decay data. The relative cross sections for the high spin and low spin isomers of the ERs were calculated from A(Ti) NC - 08.1 using A(Ti) = Noi* [ 1- exp(-\Ti)] (1) where i refers to h or 1 for the high spin and low spin isomers respectively. The nuclidts studied in this work are shown in Table-1 along with the channels by which they are produced and the relevant nuclear spectroscopic data. The isomeric cross section ratios (oh/oi) are shown in table-2 for various projectile energies. The data show lower values of isoir.eric cross section ratios for the ERs formed by alpha emission compared to those formed by neutron emission alone. This is expected as the alpha particles can carry away large angular momentum from the CN. The isomeri'* cross section ratios are found to increase with projectile energy as expected from the complete fusion hypothesis in which angular momentum of the CN increases with projectile energy. This indicates that the ICF does not become a significant fraction of the fusion reaction upto a projectile energy of 6 MeV/amu. Acknowledgements The authors are grateful to Dr. P.R. Natarajan Head, Radiochemistry Division for his keen interest in this work. The authors are thankful to the pelletron operating staff for their cooperation during these experiments. References 1. J.R. Birkelund and J.R. Huizenga, Ann. Rev. Nucl. and Part. Sci. 33, 265 (1983). 2. D.J. Parker, Phye. Rev. C39, 2256 (1989). 3. V.E. Viola et al., Phys. Rev. C26, 178 (1982). 4. T. Inamura et al., Phys. Lett., 84B, 71 (1979). Table-1. Nuclear spectroscopic data Reaction Nuclide Half Spin Gamma Abundance channel life state Energy % 2n 99m-Rh 4 .5 h 9/2 + 340.8 keV 69.1 2n 99g~Rh 16 d 1/2- 527.7 keV 40.7 4n 97m-Rh - 44.3 m 1/2- 421.5 keV 13.3 4n 97g-Rh 31 .1 m 9/2+ 421.5 keV 75.0 a2n 95m-Tc 61 d 1/2- 204.1 keV 66.5 a2n 95g-Tc 20 h 9/2 + 765.8 keV 93.9 a3n 94m-Tc 52 m 2+ 871.0 keV 94.1 a3n 94g-Tc 4 .88h 7 + 702.7 keV 99.8 871.0 keV 100.0 Table-2. Isomeric cross-section ratios (ICR) in 12c + 89y Lab. ICR ( oh/ Energy(MaV) 99-Rh 97-Rh 95-Tc 94-Tc 43.2 b. 66 - 2.04 — 52.0 11. 82 8.7 6 4.31 5.75 5S.4 17. 67 10.07 4.72 8.76 63.0 — 16.04 8.57 16. 45 66.8 — 17.33 73.5 — 22.06 NC - 08.2 THE LOW-LYING EXCITED STATES OF 197Hg (64-2 H) AND THE ISOMERIC TRANSITION AND ELECTRON CAPTURE DECAY OF Tig (23.8 H). N.Chakravarty, S.S.Rattan, R.J.Singh, A.Raroaswami and Satya Prakash Radio-chemistry Division, B.A.R.C, Bombay— EXPERIMENTAL - The 19jT,}, 197mHg & 197Hg activities were produced by bombarding a thin Au foil with 50 MeV a particles from the VECC cyclotron at Calcutta. The stacked foil technique in concert with an 8% HpGe detector (FWHM 2 keV at 1332 keV) a 16K Canberra MP sy^teai & standard ORTEC electronics were used. The details of the experiment are to be found in [5]. The gamma ray spectra were analysed using the programme SAMP0[6]. The peak area was corrected for the dead time losses arising from the spectroscpy amplifier[7]. RESULTS & DISCUSSIONS - Using the principle of level intensity balance for the 3/2- & 5/2— levels (cf. decay scheme in ref.8) & the standard expression relating absolute intensity to dps for Tl, we obtained the absolute intensities of the 152, 134 & the weak 18.18 keV photons [Table 1]. In this way we also obtained the relative intensity of the 165 keV gamma ray arising from the IT of Hg & it's CEC, a165 [Table 2]. In the determination of these quantities, the values of 0134 4 01152 were taken to be the NC - 09.1 experimental CECs[8]. The IT & EC branching fractions were obtained from a linear least square fit_of the growth decay equation pertaining to the activity of Hg. The log(ft)g was determined using the branching fractions and the log(f) values[9] [Table 2]. The 18.18 keV peak was confirmed to be an Ml transition by compai Ing the experimental intensity branching ratio to that obtained for an Ml & E2 transition using the single particle model of Moskowski—Weisskopf[10]. Using the same model, the ^ transition probability for the 165 keV photon was compared to the "4 transition probability obtained experimentally whereby the hindrance factor was found to be only about 4, strongly suggesting that the isomeric state is a single particle 13/2+ state. This confirms the neutron ii3/2 assignment given by other workers[l] & does not indicate a significant anomaly of this level's charecterstics. REFERENCES 1. R.Vianden & K.Krien, Nucl.Phys., A277, 442 (1977). 2. B.Jung & H.Svedberg, Arkiv Fysik, 19, 429 (1959). 3. A.J.Ha*-erfield et al, Nucl.Phys., A64, 379 (1965). 4. S.Dhulok.a Reddy et al , Pramana, 8, 81 (1977). 5. N.Chakravariy et alr Radiochimica Acta, 48, 1 (1989). 6. J.T.Routi & S.G.Prussin, Nucl.Instr. & Meth., 72, 125 (1969). 7. S.S.Rattan, REPORT BARC-1527 (1990), B.A.R.C. Bombay, India. 8. C.M.Lederer & V.S.Shirley, Table of Isotopes, 7th. Edn., Wiley, N.Y. (1978) 9. N.B.Gove & M.J.Martin, Nucl. Data Tables, 10, 205 (1971). 10.S.A.Moskowski, a B & % Ray Spectroscopy, ed. K.Seigbahn, North-Holland, Amsterdam (1965), Chapter 15. TABLE 1: Absolute gamma ray intensities and total CEC 165.0 keV 133.9 keV 152.2 keV 18.18 keV X Inten- (8.51 ± 1.80 ± 0.13 8.24 ± 0.56 (4.29±1.07)•10-3 -3 sity. i.7O)*io Table 3 197mHg Decay IT Branching Fraction EC Branching Fraction 1og(ft)EC 91. 4 ± 0.7 8.6 ± 0.7 6 .21 ACKNOWLEDGEMENTS - The authors thank the Cyclotron Operations & Computer staff of the VECC Calcutta & the Instrumentation Section Staff of R^.Ch.D., B.A.R.C, for their help & cooperation. They express their thanks to Dr. P.R.Natarajan, Head R.Ch.D., B.A.R.C, for his encouragement for this work. NC - 09.2 RC - Radiation Chemistry Papers : RC - 01 to RC - 46 SOLVATION OF EXCESS ELECTRON IN POLAR LIQUID :AN OVERVIEW T.Bandyopadhyay Chemistry Division,Bhabha Atomic R«search Centre, Bombay - 400 085, India. SUMMARY^ Most recent advances in time-resolved studies of aolva- tion of excess electron and its geminate recombination in polar liquid in the picosecond to subplcoaecond time domain has been reviewed. Emphasis is given on the applicability of twc-state model of solvation of this speciea.Potentlality of F«ynman path integral technique and SCF-X^-SU technique to describe absorption spectrum of this species has been indicated. IMTRODUCTION: Absorption spectrum measurement of excess electron after its injection into the medium (by means of pulse radiolysis, laser photolonization or phot©detachment)has now become a power- ful probe for investigating the dynamics of excess electron localization and solvation in liquids. In this article a set of observation is chosen upon excess electron solvation and geminate pair recombination of aolvated electron. EXPERIMENTAL; Recent advances in the technology of creation and detection of ultrashort optical signal generates an observable population of the transient on a shorter time scale as compared to the natural life time and also ksap track of the population dynamics in a well synchronized manner.In particular, in subplcosecond studies using a mode-locked dye laser the ionizing pump beam gen- erates the excess electron subsequently a second beam (prob* beam) after probing the excited region of the sample is directed toward a moncchromator and a silicon or germanium diode.Precise dynamical measurements are obtained by this technique. RC - 01.1 RESULTS AND DISCUSSIONS: The experimental data from photoionization of pure liq- uid or mixture of liquid and photodetachinent experiments of simple h&lide ions suggests that the solvation process can be described by a two-state model.An absorption band in the IR is observed at the early stage which then decays and an analogous band grows in the visible.The IR band is characterised duo to localization of electron in preexisting solvent trap whereas the visible band ia due to the electron in equilibrium solvation. Intereat 1ngly,an absence of tine-dependent shift in the absorp- tion apectra suggests that the relaxation process of the sur- rounding solvent molecule is not continuous and perhaps the sol- vation process occurs through an electronic transition induced by the dielectric relaxation of the medium. The inference become straight forward when one considers the energetics of quasifree bound electron transition and refers the original notion of "dry","wet"and "solvated" electron to the quasi-free state,quasi localized state and localized state, respectively. Electron photodetachinent studies from a simpla halide ion in aqueous solution Identifies a trapped state of *:he electron, a precursor of wet electron which absorbs in the visible. The wet to solvated electron transition once again conforms to the two- state model. Further investigation on the decay of hydrated electron visible absorption band in the first 100 pa. reveals that the tine dependence of this band can be described by a geminate recombination process occurring between es~u j and Ha0 or OH. On th<- Theoretical aide Feynman path integral simulation technique is the first to provide a detailed picture of the mi ture of electronic transition monitored by spectroscopic method. To the authors knowledge this powerful technique is yet to be ap- plied for studying geminate recombination process. Electronic structure calculation of the equilibrium solvation cluster by SCF-X<-SU method could be a potential technique for a better un- derstanding of the nature of absorption spectrum of this species. ,?C - 01 .2 THE EPR PARADOX, THE HYDRATED(SOLVATED) ELECTRON, AND THE REACTIONS OF THE HYDRATED ELECTRON C. GOPINATHAN CHEMISTRY DIVISION, BARC, BOMBAY 400 085, INDIA SUMMARY- Einstein, Podolsky and Rosen pointed out in 1935 a serious flaw in the reasoning behind quantum mechanical modelling. It is shown in this work that this flaw applies to the interpretation of the hydrated electron and its reac- tions as well. ( Key words : hydrated electron , reactions, EPR paradox) INTRODUCTION In their classic paper of 1935/1/ Einstein, Pouolsky and Rosen argued that quantum mechanics does not provide a genuine picture of reality since concrete values can not be given for all the parameters considered. Since then inspit© of lot of work(e.g. Bell's theorem) satisfactory answers have not been provided for the questions raised by Einstein. The Schrodinger's cat is a crude example of this problem. In the present woik it is suggested that the hydrated electron in its potential well ^s a better model for this paradox than the famous cat. THE HYDRATED ELECTRON The discovery of hydrated electron by E.J.Hart in 1962 created tho problem of explaining away the electrostatic at- traction of ths parent positive centre espescially since the low fast dielectric constant applies in short time scales. However, the concept of the effective dielectric constant helped to some extent. However, the quantum mechanical aspect of the stability of the hydrated electron has not been looked into seriously. The absorption maximum of the hydrated electron corresponds to an energy of about 1.7eV. Since, A E * A T = h where h is the Planck's constant andAE and AT are the uncertainties in energy and time respectively. Therefore from the uncertainty principle the hydrated electron 3hould be identifiable only after a few picoseconds. However, we know now that the hydrated electron is fully formed in 100 femtoseconds. The second contradiction- comes from the Bell's theorem. Since it is formed by the electron ejection from an atom, quantum mechanically the fata of the hydrated eletron and NC - 02.1 its parent ion should be linked. Tha radius of the spur can be as low as 0.5nm/2/. The stability of the hydrated electron therefore is surprising. We have shown earlier/3/ that in frozen systems the trapped electron reacts predominantly by tunnelling. The same is true of super fast reactions in liquid water at room temperature/2/. These tunnelling processes take place over a distance(3 nm or so at 77K). With the possibility of such reactions how is the hydrated electron is formed in the first place? Once we know the answers to these questions we will knew the nature of quantum mechanical reality. REFERENCES 1 A. Einstein, B. Podolsky and N. Rosen, Phys. Rev., 41, 77 (1935) 2. C. Gopinathan and G.Girija, Radiat. Phys. chem. , 2JL, 209, (1983) 3. S. Kapoor, C. Gopinathan and R.M. Iyer, J.Radioanal. and Nucl. Chem., Articles, 130, 219, (1989) RC - 02.2 POLARITY EFFECT ON SOLVATKD ELECTRON REACTIONS Sudhir Kumar Kapoor and C. Gopinathan Chemistry Division, Bhabha Atomic Research Centre Bombay 85, India. SUMMARY - The effect of the addition of isopropanol on the reactivity of solvated electron towards uranyl nitrate and urea was studied using the pulse radiolysis technique. The results show that with the increase of isopropanol concentration in water the reactivity of solvated electron towards uranyl nitrate does not change. The results have been interpreted on the basis of the trapping of the maxtrix. In case of urea the behaviour was different. (Key Words: e^ctron accelerator, solvated electron, pulse radiolysis, polarity effect) I INTRODUCTION Mixtures of alcohol and water show many interesting properties/1/. it is known that optical absorption spectra of solvated electron depend on the polarity of the matrix. To examine the effect of the polarity on the reactivity of solvated electron, we have studied in detail the reactivity of solvated electron in water: isopropanol mixtures. II EXPERIMENTAL Isopropanol (G.R.) was purified by refluxing with sodium borohydride for 7 hrs. It was then fractionally distilled. The pulse radiolysis set up employed has been given elsewhere/2/. All solutions were prepared in nano pure water and were purged with high purity Iolar grade nitrogen. III RESULTS AND DISCUSSION The reactivity of solvated electron towards scavengers were measured at 650 nm in various compositions of water and isopropanol. First order decay rate constants, k-^, of solvated electron were measured at different concentrations of each scavenger. The bimolecular rate constant, k2, was obtained from the slope of the plot of k-, against the scavenger concentration. The results are given in Taole 1. t The addition of isopropanol in water decreases the polarity of the matrix. As can be seen from the table 1, the bimolecular rate constant for uranyl nitrate does not change with the composition of the matrix. This shows that the scavenging efficiency of the efficient scavenger does not change with the composition of the matrix. However, in the case of urea the reactivity increases at about 10% (v/v) of water in isopropanol. RC - 03.1 Thus, it can be said that the scavenging efficiency of inefficient scavenger increases as the polarity of the medium decreases. In other words, the tunnelling of the solvated electron towards inefficient scavenger may take place only when the polarity of the medium is low. The results are strikingly similar to the results obtained by us in frozen systems under steady state irradiation/3/. IV REFERENCES 1. J. Cygler and G.R. Freeman, Can, J. Chem. £2, 1265 (1984J 2. S.N. Guha, P.N. Moorthy, K. Kishore, D.B. Naik and K.N. Rao, Proc. Indian. Acad. Sci (Chem.Sci) ^9, 261, (1987) 3. S. Kapoor, C Gopinathan and R. M. Iyer, J. of Radioanal. and Nucl. Chem,, Articles, 130, 409, (1989) Table \ Bimolecular rate constant of solvated electron Composition k2(dm mol •"• s •*") (v/v) Water: Isopropanol Uranyl nitrate Urea 100 : 0 6.6 x 1010 2.3 x 10; 75 25 6.1 x 1010 2.2 x 10: 50 50 2.1 x 10; 10 25 75 5.3 x 10 2.1 x 10; 20 80 2.2 x 10; 10 90 5.7 x 1010 5.1 x 10- 0 :100 4.1 x 1010 RC - 03.2 DIMER AN ION FORMATION IN THE PULSE RADIOLYSIS OF 4-PYRIDINOL IN AQUEOUS SOLUTIONS D.B.Naik and P.N.Moorthy Applied Chemistry Division B.A.R.C, Trombay Bombay 400 085 SUMMARY - Absorption spectra of semireduced 4-pyridinol species formed by reaction of e".q with 4-pyridinol were investigated at different concentrations of the s.olute. From the difference in the nature of the transient absorption spectra at low and high concentrations, formation of dimer anion has been inferred. Dimer anion absorbs in the 620-740 nm region as compared to the nionomeric species which has absorption maximum at 320 nm. By plotting l/0Di7o n. versus 1/[4-pyridino 1 ] equilibrium constant for the dimer formation has been determined to be 97. ( Key words: Pulse radiolysis, 4-pyridinol, Dimer anion, Equilibrium constant) I. INTRODUCTION The one electron reduction of 3- and 2-pyridinols have been reported earlier/1,2/. In this paper, the results of our investigations on the one electron reduction of 4-pyridinoi are presented. I I. EXPERIMENTAL 4-pyridino 1 (F!uka, Tech grade) was purified by passing its aqueous solution first through an anion exchange column! Dowex-i X8 50-100 mesh) and, then through a cation exchanger! Tulsion T-42 (GEDH + 2Q-50 mesh) to remove ionic impurities and the purified solution was evapourated to dryness. The resulting material was stored in a desiccator as it is hygroscopic. Pulse radiolysis set up used has been described elsewhere/3/. Solutions were prepared in nanopure water! conductivity<0.l &S ). III. RESULTS AND DISCUSSIONS 4-pyridinol has pK. at 3.27 and 10.09/4 /. In the pH region 5-9 it exists in the neutral form. At pH 6.8 (adjusted with phosphate buffer) the rate constant for the reaction of e" «„ with 4- pyridinol was determined by following the pseudo first order decay of e".„ at 720 nm and was found to be 8.4x10* dm3mo 1"' s"' . In deaersted neutral solution! pH 6,8) i containing 2xlO~4 moldrrr3 4- pyridinol (with 0.1 moldnr3 t-butanol as OH-radical scavenger), the transient absorption spectrum was monitored in the 250-750 nm region and was found to exhibit ^.». at 320 nm and a shoulder in the region 360-400 nm attributable to the semireduced species. The latter decays by good second order kinetics in the 3,60-420 nm region with 2k/£l value of 1.5x10' s"• (at 380 nm). The decay In this band is' accompanied bv build up of absorbance at 320 rim. RC - 04.1 At higher concentrations of 4-pyridinol, the nature of the absorption spectrum of the transient species was found to be different with a new absorption band appearing in the region 820- 740 nm. In this region transient absorbance was found to increase with increasing concentration of 4-pyridino 1 (Tab 1e l) suggesting that this long wavelength absorption band is dijs to a product formed by reaction of 4-pyridinol with tlie initially formed semireduced species. It is reasonable to infer this to be a dimer anoin. Equilibrium constant for dimer anion formation: 4-pyl + 4-pyl" > ( 4-pyl' )2~ is given by the relation C (4-py 1*>2 " ] 1/ ______ [4-py 1 ] C4-py 1*" ] From this we can derive: 1/ODo,,. - 1/OD... + 1/0D... .K. [4-pyl 3 . where 0Do „ „ is the 0D of the dimer anion observed at 670 nm and OD. . „ corresponds to the OD of dimer anion under conditions such that al1 the monomeric species are converted to the dimer anions. The experimental data obtained at different concentrations of 4- pyridinol were found to obey the above re 1 ationshipw The value of the equilibrium constant as calculated from intercept/slope ratio is 97. IV REFERENCES 1. D.B.Naik and P. N. Moor thy, J.Chem.Soc. , Perkin Trains. II. 705 (1990) 2. D.B.Naik &nd P.N.Moorthy, Proc.Indian Acad.Sci.(Chsm.Sci.) (commun ic^ted) 3. S.N.Guha, P.N.Moorthy, K.Kishore, D.B.Naik and K.N.Rao Proc.Indiain Acad.Sci. (Chem.Sci), 99., 271 (1987) 4. S.F.Mason, 1253 (1959) Table 1 Concentration of 4-pyridinol Absorbance at 670 nm mo 1 dm"3 8.0 x 10"4 0.0065 1.6 x 10"3 0.0106 2.4 x 10"s 0.0145 3.0 x 10'3 0.0167 4.0 x 10"3 0.0217 RC - 04.2 REACTION RATE CONSTANTS OF e^ AND OH RADICALS WITH ALKYLBROMIDES (AB) IN AQUEOUS SOLUTIONS H.S. Mahal and Manohar Lai Chemistry Division Bhabha Atomic Research Centre Trombay, Bombay 400 085, India. Halogenated molecules even though toxic in nature are increasingly being used by humans for a variety of purposes.* ' The toxicity appears to be due to the free radicals formed when the halogenated molecules are exposed to light and/or by the one electron reduction of these molecules during metabolism. Radiolytic degradation of certain simple model halogen compounds have earlier been reported . We have in the past reported the K-radi olysis of bromoform'J' and 1,2 dibromoethane' ' . The pulse radiolysis of 1,2 dibromoethane has earlier been reported' . We report, here t he reac* ion rate constant of e~ and OH radicals with ethlybrornide, l,S,t butylbromide, bromochforoethane, ProPylbromide, bromopeniane, tetrabromoothane and 1,2 dibromoethane in aqueous solutions by pulse radiolysi3 technique. Experimental: All alkylbromides were of G.R. quality. All other reagents used are of purest grade. Nanopure water with conductivity ~0.06 us has been used in this study. Fresh samples were prepared prior to each experiment. For e~ reactions the purified water containing 1M t-butanol in glass cells with rubber septum was saturated with iolar N2. Similarly for OH radical reactions, aqueous solution was saturated with t^O. To these ^/or N^O saturated solutions is added known amounts of volatile alkylbromides via the rubber septum and stirred with a magnetic starrer to dissolve it.*4'. This solution is transferred by a syringe to a cell. Dose per pulse ~12 Gy. Results and Discussion N2 saturated aqueous solutions containing IN t-butano] and 0.05 mM to 0.005 mM alkylbromide is exposed to pulse of electrons. The decay of the optical absorption of e~ at 700 nm is analysed for different alkylbromide concentrations. The decay was found to be exponential with time. The 1st order rate constant when plotted against [AB] were straight lines. Biomolecular rate constants determined from the slopes of such plots are presented in table 1. The rate constant for the reaction of OH radical with alkylbromide could not be evaluated by the simple competition technique (reaction 1 & 2) because the alkyl OH + C2H5Br > C2H4Br + H?O (1) OH + SCN~ > SCN + OH~ (2) bromide radical decaying to C2H4 and Br and Br also oxidiing thiociate 4 SCN/Br + SCH/Bc^zz^ Br/Br~2 + SCN/(SCN)~2 (3) RC - 05.1 This problem was sorted out hy taking low Br^dxlO M) and high [AB] >> 5x10 M and varying t-butanol concentration e.g. 10 J - 10 **M and exposing it to electrons. The now between AB and t-butanol for OH radicals and iE _ for,maMon is the indicator for reaction 1. Taking kpn+t-But =5x10 M the kOH+AB«s are evaluated and presented in table 1. Table 1 Rate constants for the reactions of e~ and OH with Alkylbromides Compounds pH jZlO 9 MS 8 MS *eaq+AI *OH+AB Et hy.l bromide 7.0 8.0 (12) 1.2 1 Butylbromide 7.0 ' 9.0 (10) 1.4 t-Butyl bromide 7.0 7.2 2.0 S- Butyl bromide 7.0 7.2 1.4 Bromochloroethane 7.0 8.0 - 1-2 dibromoethane 7.0 14.0 (12.0) 2. A (2.0) propylbromide 7.0 10.0 1.5 bromopentane 7.0 8.0 1.1 tetrabromoethane 7-0 17.0 0.8 The results demonstrate high fcpaq + AB -l-2xlO10 M 1S~ . These rate constants agree well with those reported earlier (see values in parenthsis in table 1). ^eaq + AB are h^9ner f°r tne alkylbromides having higher number of bromides attached to it. OH rate constants are = l-2xlO8 M~ S~ . The OH abstracts one hydrogen atom to form a alkylbromide radical and water whereas e~ cleaves the brcaiide atom to form Br and a radical. kOH+AB are higher for compounds having higher number of C-H groups in it. References T.F. Slater, Free Radical mechanisms in tissue Injury, Pion London 1972 E.S. Reynolds and M.T. Moslen, in Free Radicals in Biology Vol IV ed. W.A. Pryor, Academic Press, New York 1980.. J.R. Trudell,- B. Bosterling and A.J. Tervor mol. Pharmacol, 2i, 710, (1982) Manohar Lai, J.Monig and K.D. Asmus J. Chem. Soc. Perkin Trans II 1639, (1987) 5 Manohar Ldl & H.S. Mahal Radiat. Phys. Chem. 32, 599 (1938) 6 Manohar Lai, Radiat. Phys. Chem. 32, 741 (1988) 7 Manohar Lai, J. Monig and K.D. Asmus, Free Rad. Res. Commn. 1, 235 (1986) Manohar Lai, C.Schoneich, Jorg Monig and K.D. Asmus. Int. J. Radiat Biol 54, 773 (1988) ' RC - 05.2 CHARACTERISATION OF THE TRANSIENTS FORKED FROM ACRYLIC ACID AND MSTHACRYI.IC ACID Hanmohan Kumar and M. H. Rao •Chemistry Division, Bhabha Atomic Research Centre, Trornbay, Bombay 400 085, India. SUMMARY - The transient species formed from the reaction of hydrated electron and hydroxyl radical with acrylic acid and methacrylic acid monomers have been characterised by their ab- sorption spectra and redox behaviour.' Based on these it is :on eluded that the site of attack by hydrated electron is the car - buny1 group, whereas OH radical adds onto the >C=C< bond uf the monomers and the most probable structures of the resultant tran- sient.'; are assigned. (Key Words: Electron Accelerator, Pulse Radiolysis, Polymerisation, Acrylic Acid, Methacrylic Ac.id) I. INTRODUCTION In the radiolys.is of dilute aqueous solution of vinyl monomer, the primary species from water (i.e. hydrated electron, hydroxyl radical and hydrogon atom) can react with monomer to form different transient species which can subsequently propagate polymerisation under suitable experimental conditions/] 3/. A time resolved technique such a"s pulse radiolysis is a powerful tool to infer about- the structure of these different resultant transient species. II. EXPERIMENTAL Acrylic acid and methacrylic acid were vacuum distilled before use and G.R. grade acetic acid was used. Full details of the pulse radiolysis set up are given elsewhere/4/. Tertiary butanol was used for scavenging OH radicals in N2 saturated matrix, whereas N20 gas was used to scavenge hydrated electrons. III. RESULTS AND DISCUSSION Reactivities of acrylic acid and methacrylic acid with hydrated electron and OH radical were found in the range of 5xl09 to 12xl09 dm3 mol * s ' , whereas the reactivity of acetic acid was found about three orders of magnitude lower (3xl(36 to 8xl0R dm3 mol i s ' )- These observations conclude that presence of >C = C' group in i •<~'ii juge.ti on with^ >C~0 group enhances the reac- tivity of monomer .wards hydrated electron and OH radical. The resultant. transient species were showin • absorption in the lange of 250nm to 380nm with extinction coefficient at respective >max in the range 400 to 5000 dm3 mol-i cr' It was found that the transients formed by the reaction of hydrat.ed electron with the monomers were capable of reducing the thionine dye molecule. This shows the reducing nature of these RC - 06.1 On the other hand the transients resulting from OH radical reactions with these monomers were not able to reduce thionine dye. In the case of acetic acid both the transients formed by hydrated electron and hydroxyl radical reactions were found reducing in nature These observations suggest that the most probable site of attack by hydrated electron is the carbonyl group of the monomer resulting in a reducing carbon centered radical, whereas OH radical adds onto the >C=C< group, resulting in a non-reducing tertiary radical. IV. REFERENCES 1. Manmohan Kumar, M.H. Rao, P.N. Moorthy and K.N. Rao, Radiat. Phys. C'hem. M, 219 (1989). P.. Manmohan Kumar, M.H. Rao and P.N. Mooi-thy, J. Hacromol . Sci.-Chem. A2K3). 299 (1990). 3. Manmohan Kumar, M.H. Rao, and P.N. Moorthy, Radiat. Phys. Chem. 36. 811 (1990). 4. S.N. Guha, P.N. Moorthy, K. Kishore, D.B. Naik and K.N. Rao, Proc. Indian. Acad. Sci. (Chem. Sci.) £9, 261 (1987). RC - 06.2 PULSE RABIOLYSIS INVESTIGATIONS 018 THE REACTION OF OH RADICALS WITH ALKY It. SULPHIDES Hari Mohan Chemistry Division, Bhabha Atomic Research Centre, Bombay 400085, B* Anklara an<3 K.-D. Bereich Strahlenchemie, Hahn Meitner Institute, Berlin,F.R.G. SUMMARY The yield and life time of dimer radical cations, formed on reaction of OH radicals with alkyl sulphides, \s observed to decrease with increasing chain length of alkyl group and Amax shows a red shift with increasing electron inductive power of the substituents. (Key words: Alkyl sulphides, pulse radiolysis, aqueous solutions, OH radical reactions) I. INTRODUCTION The OH radicals are known to undergo one electron oxidation of organic sulphur compounds and form sulphur centred radical cations. The sulphur centered radical cations show strong ten- dency to stablize by co-ordination with a free electron pair from another sulphur or hetro atom,, The effect of alky], group on the stability and /\max of dimer radical cations of alkyl sulphides is reported in this papetr. II. EXPERIMENTAL The pulse iradiolysis investigations on alkyl sulphides are carried out in aqueous solutions on irradiation wi':h high energy electron pulses (1.55 MeV, 1 fis) from the Vande Graaff ac- celerator of HMI, Berlin. The dose per pulse is ~2 Gy. III. RESULTS AND DISCUSSION Pulse radiolysis of N?O saturated aqueous solution of propyl-S-methyl (0.5 mmol dm , pH=3.5) produced transient opti- cal absorption band with Amax=480 nm. The decrease in the RC - 07.1 conductivity of the solution on pulse radiolysis suggests the formation of a positively charged species. The intensity of the transient band increased with solute concentration, indicating it to be due to dimei: radical cations.1 Another small absorption band observed in the region of 280-300 nm is assigned to <^-thio radicals formed on H atom abstraction. These studies have been extended to various alkyl sulphides (Table). The yield and life time of the transient band of dimer radical cation decreases with increasing chain length, indicating th« decreased stability of of tne the dimer radical cations. The Amax transient band shows red shift with increasing electron inductive power ( o~ ) of the substituents and a linear correlation is observed between them. TABLE Physical properties of dimer radical cations of various alkyl sulphides (0.5 mmol dm ) Solute (nm) (us) (dm3 mol"1 cm 1) Ethyl-S-methyl 475 160 3.8 4490 Propyl-s-methyl 480 150 3.7 5020 Propyl-S-ethyl 490 55 3.2 5640 Propyl-S-i*propyl 520 20 2.4 4300 But.yl-S-methyl 485 14H 2.6 5520 Sec-butyl-S-ethyl 515 14 1.6 5535 2,methylbutyl-s-methyl 490 110 2.1 5710 Hexyl-S-methyl 495 75 1.7 5470 Hept.yl-S-methyl 500 5 5 1.5 5890 Octy1-S-methyl 515 40 0.4 5240 REFERENCES 1. M. Gobi, M. Bonifacic and K.-D. Asmus, J. Am.ChenuSoc. 106, 5984 (1984) RC - 07.2 PULSE RADIOLYSIS STUDIES OH THE REACTION OF e~q WITH SUBSTITUTED ALKYL SULPHIDES Hari Mohan Chemistry Division Bhabha Atomic Research Cent re Bombay 400 085 SUMMARY Due to lowering of electron density at: sulphur in substituted alkyl sulphides, the reactivity of e~ is observed tc increase when electron withdrawing groups are present in the parent compound. (Key words: Pulse radiolysis, reaction of e~ , alkyl sulphides) I. INTRODUCTION The knowledge of oxidation/reduction reactions of organic sulphur compounds is important in order to understand the physico chemical processes taking place with sulphur drugs, amino acids and other biological systems containing sulphur. With this objective, the reaction of e~Q with substituted alkyl sulphides have been studied by the pulse radiolysis technique. II. EXPERIMENTAL The pulse radiolysis experiments are carried out with high energy electron pulses (7 Mev, 50 ns) from a linear accelerator. The dose per pulse is y-> 1.0x10 ev cm. III. RESULTS AWD DISCUSSION The reaction of e~ wwith different alkyl sulphides is aq studied by following the decay of t he transient band of eaq' formed on pulse radiolysis of N2 saturated aqueous solution (pH = 7.0, t-butanol = 2.0 mol dm"-, )\ = 700 nm), for various concentrations of substituted alkyl sulphides. The decay of eT" becomes faster and of first, order, suggesting the reaction of eT_ 1 aq with substituted alkyl sulphides. The pseudo first: order rate RC - 08.1 constant is observed to increase linearly with the solute concentration. The bimolecular rate constant is determined from the slope of linear plot of pseudo first: order rate with solute concentration and the values are shown in the table. The reactivity of e~ with alkyl sulphides is quite low due to high electron density at sulphur. The presence of electron withdrawing groups are expected to lower the electron density at sulphur and thereby increase the reactivity of e~ with substituted alkyl sulphides. The high electron withdrawing power of COOCH3 (+2.0), compared to those of CH3 (0.0), CH2OH (+0.56), and CH2COO~ (-0.06) would lower the electron density at sulphur and increase the reactivity of e~ This explains the higher rate constant value observed for (CH2CH2COOCH3)2S. IV. REFERENCES 1. G.Vt Buxton, C.L.Greenstock, W.P. Helman and A.B. Ross, J.Phys.Chem.Ref.Data, 17, 513, 1988. 2. (a) R.W. Taft, J.Chem.Phys., 26, 93, 1957. (b) G.B. Barlin and D.D. Perrin, Quart.Rev. 20, 75, 1966. TABLE Rate constant values for the reaction of e~ aq with substituted alkyl sulphides Solute KxlO 8 (dm3 rnol"1 s"1) (CH3)2S 0.20' (CH2CH2OH)2S 0.36 tCH2CH2CH2OH)2S 0.40 (CH2COO")2S 0.89 (CH2CH2COO~)2S 0.70 (CH2CH2COOCH3)2S 5.20 •^reference 1. RC - 08.2 PULSE RAPIOIYSIS OF THIOHREA IN AQUEOUS SOLUTION S.P.Ramnani,S.Dhanya and P.K.Bhattacharyya Chemistry D i v i a i on , Rhabha Atomic Res«dn:h Cen t. re , Bombay 400095 SUMMARY Reactions: of OH radical and e with thiourea were investigated i ri aqueous solut iona.Thje rate constants for these reactions were determined as 5.3x109M~JS"* and 4.5x109M"*S~* respectively.These reactions lead to different transient radical species.Decay kinetics of these transients were also inves- t i gated. CKey words:Pulse rad i o I y s i s , th i our ea , 11 ans i ent opedra.rate cons tan t n] I. INTRODUCTION Pulse radiolytic investigations on the aqueous solutions of thiourea have been carried out.It was reported that /I/ on pulse radiolysis of thiourea transient species were generated resulting in a complex absorption spectra with maxima at 29*3 and 400 nm.These were tentatively assigned to reaction products of e~* and OH radicals respectively with thiourea.The present investiga- tions are aimed at determining the rate constants of these reac- tions and the subsequent decay kinetics at different conditions. I I.EXPERIMENTAL Pulse radiolysis was carried out using linear accelerator capable of generating 25ns to 2us pulses of 7 Mev eJectrona.Detai1s of pulse radiolysis set up are described else where/2/.The transients were monitored by following their optical absorptlon.A11 the aqueous solutions were made in nanopure water.For purging the solutions IOLAR grade N^ and N^O were used.Thiourea^frum SI SCO Research Laboratories.Bombay was recrys- tallised from water before use. I I I.RESULTS AND DISCUSSION a)Reactiun of Oil radical ^0 saturated solution of thiourea (pH=7) was pulse radio- lysed using 50 ns electron pulse.Under these conditions e is scavenged by N£0 to increase the lyield of CH radicals.A transient growth was observed which was monitored in the wavelength range of 280 to 500 nm.Frum the final absorption of this growth a tran- sient absorption spectrum with maximum at 400 ntn was con- st rue ted. Growth rate at 400 nm followed pseudof i t-st order kinetics.Thus rate constant for the reaction of OH radical with thiourea was obtained as 5.5x1G^M^S"1.Transient absorption at 400 nm f6llowed second order kinetics with k/,c RC - 09.1 cal.However at pH above 9.5 the decay kinetics changed to pseudo- first order depending on the OH concentrations,indicating a reaction between OH and the transient.which has a bimolecular rate constant of 5x108M']S"J. b)Reaction of e~ with thiourea For siudying this, N£ saturated solution of thiourea con taining 0.1M tert-butanol was pulse radio1ysed.Under thib condi- tion OH radical is scavenged by tert-butanol and only hydrat td electron reacts wtth i h i our ea . Decay of hydrated ulttclron absolu- tion at 700nnn followed pseudof irst order kinetics depending ori the thiourea concent rat i on . Thus the b irno 1 ecu 1 ar rate constant is determined as 4. }x)0°M ^ S ' .During the decay of e ~ abr. cirption a transient qrowtli was observed which was monitored in the wavelength range of 270 rim to 500nrn . From the f i na 1 absorption of '-his growth s i gna 1 transient absurpt.iun spectrum having maximum at 290nni was obtained. The reaction of thiourea with Ohi which lead to the forma t i on of a neutral transient species can be either hydrogen abstraction from NHp group or addition to >C=S group.The visible absorption peat: at 400nm •. tronyly i.uyyutits that it is a sulfur centered radical . Simi lar absorpt ton wa<> also found in the case of ethylene tr i Ihioi.arbonale which also contain a >C = S group/?/. On the other hand 290 rim absorption peak in the reaction of thiuurea with e~ suggests that the transient produced is carbon centered radi- ca 1 . IV.REr ERENCES 1.A.Chai1esby,P.J.Fydelur,P.M.Kopp,J.P.Keene and A.J.Swa11ow. "Pulse radiolysis' Ed.by J . H . Baxenda 1 e and J . P . K'-ene . Acadenii c Press,New York<1965). 2 . S . P . Ramnan i.S.Ohanya a.iM.1 P.K . Bhdttacliaryya , Raiiidt . f-'hy a . Chew 36 , 409 < 1 'J9Q . > RC - 09.2 A PULSE RADIOLYSJS STUDY OF ONE ELECTRON OXIDATION OF SULPHACETAMIDE S.Sabharwa!, Kama! Kishoie and F.N.Moorthy Applied Chemistry Division Bhabha Atomic Research Centre Trombay, Bombay - 400 085 SUMMARY - PI.I 1 s9 radiolysis technique has been used to produce and characterize t'he transients formed by the one- electron oxidation of s IJ 1 phace tarn i de in aqueous solutions using specii io ons electron oxidants I ike Cl2 • ,Br,' • ,N3 • ,S0. • . The spectral and kinetic characteristics of the rasulting species, its acid dis- sociation constant t r.K. ) and subsequent reaction with ascorbic acid have been investigated. (Key Words: Su1phacetarnide, Cation radical, Pulse radinlysisl 1. INTRODUCTION Suiphonamide drugs, widely prescribed for the treatment of many infections, are known to undergo photodegradation and cause photoa11ergic and phototoxic reactions/1/. This has stimulated interest in the photochemical redox behaviour ot these drugs/2/. The present paper describes our investigations of one-electron oxidation of su1 pacetamide using pul^" iadiolysis to characetrize spectral and kinetic properties of the resulting species and its reaction with ascorbic acid, which in biological system can limit the damage caused by the radicals. I I . EXPERIMENTAL Su1phacetamide sample was from Sigma chemicals and used as received. All other reagents were of AnalaR grade. Detaiis of pulse radiolysis set up are discussed elsewhere/3/. Single pulses of 50 ns were used and typical doses were about 15 Gy as measured by thiocynate dosimetry. The pHs of solutions were adjusted using appropriate buffers. III. RESULTS AND DISCUSSION Radiolytic one electron oxidation of su1phacetarnide was brought about using specfic one electron oxidants viz. N3 • ,B r7 " • , C 1 2 ~ ' and S0»"• and these were generated by reaction of OH radical with 2 Ns~,Br ,C1" and e «„ with S=0e" respectively under suitable conditions of ambient and pH. We chose these radical species as the primary oxidizing agents rather than OH radicals to ensure that electron transfer is the major pathway for reaction with the drug, free of addition and/or abstraction reactions which are common witr OH radicals. The absorption spectra of the transient species in al I the cases showed one band with \. at 420 nm (G=1300 dm3 mo I ' ' cm"1 ) at pH 9.2 and 440 nm (£=1800 dm- mol"1 cm"' ) at pH 3.8. From a study of the pH dependence of the tran- sient absorbance at 44O nm, over the pH range 3.5 to 9.2 using Err• as the oxidizing radical,the pK. of the cation radical RC - 10.1 SA* SA(-H) H + ( 1 i was determined to be 5.3 which is within limits of experimental error the same as that of the parent compound pK. 5 5.4. The rate constants , determined by following either the build up of absor- bance at A. or from the decay of the oxidant radical absorbance, are presented in Table 1. The species decayed by second order kinetics with 2k/£l = 1.3 x 10* s" ' at pH 9.2 and 8.2 x 10" s'l at pH 3.8. Reaction of ascorbic acid with the cation radical : The electron transfer from ascorbate/ascorbic acid to a number of drugs has been studied by pulse radiolysis because of the unusual biologi- .cal protective properties of ascorbate against free radical damage. It was therefore of interest to see wheather sul- phacetamide cation radical can bring about the oxidation of as- corbic acid. It was found that the cation radical formed at pH 9.2 can bring about the oxidation of ascorbic acid leading to the formation of ascorbate radical via reaction (2) as characterized by its formation at 36O nm. The rate constant for the reaction was determined to be 1.9 x 10* dm3 mol'i1 s~ ' . SA' + AH" SA + A" (2) It has been suggested that the OH radical adduct of sui- phacetamide decays to yieid its cation radical/2/. This has been confirmed by following the reaction of ascorbate ion with the OH adduct of sul phacetamid© . The spectrum obtained after 300/<-s was found to be the same as that of ascorbate radical indicating that the OH adduct decays to the cation radical which subsequently reacts with ascorbate to give the ascorbate radical. IV. REFERENCES 1. G.L.Mandeil and M.A.Sande, in The Pharmacologicai Basis of Therapeutics Edited by A.G.Gilman, L.S.Goodman, T.W.Rall and F.Murad, Macmillan Publishing Company (1985). 2. E.J.Land, S.Navaratnam,B.J.Parson and G.O.Phillips, Photo.Chem and Photobio1..35.637(1982) 3. S.N. Guha, P.N. Moorthy, K. Kishore, D.B. Naik and K.N. Rao Proc. Indian Acad. Sci.( Chem. Sci.j 9£, 351, (1987) TABLE 1 OXIDANT PH '•max (nm I Rate C(Dnfstar OH 3.8 420 1. 1 X 10' 9.2 400 9.3 X 10' Br7"- 3.8 440 8.0 X 10' 9.2 420 1. 1 X 10« N3 " 9.2 420 3.2 X 10* so«- • 9.2 420 6.0 X 10* Tla* • 3.3 440 7.0 X 10* cu-- 2.0 440 8.0 a 10» RC - 10.2 One electron reduction of substituted perflaoroarcmatic compounds. A pulse radiolysis study. Lian C. T. Shoute Chemistry Division Bhabha Atomic Research Center Bombay 400 085 SUMMARY One electron reduction of CgF5X (X= - NO2, -COOH, etc.) ieao to the formation of molecular anions. The rate constants for, their reaction with hydrated electrons vary from 1x10 to 2x10^" 1 1 M" s~ . Hammett plot of log k vs. tf"p yields /> of +0.3 indicating that perfluorocompounds have similar substituent effect as that of the corresponding hydrocarbons. INTRODUCTION Although redox properties of aromatic hydrocarbons have been extensively investigated ' , there .is no report for the perfluorocarbons. in this short communication we report the rate constants of hydrated electrons with perfluocoaromatic compounds. EXPERIMENTAL Perfluoroaromatic compounds are purchased from P. C. R. Re- search Inc. U.S.A. are used as received. Nanopure water was used in these studies. Sample flow system was fabricated in the Chemistry Division Work Shop. Pulse radiolysis system and other experimental details are as discussed earlier « RESULTS AND DISCUSSION Solution for pulse radiolysis typically consisted of 8-lGi' /u M CgF5X, 0.2 M t-butylalcohol and 2 mM phosphate buffer and bubbled with N2. £-%£*fi reacts with hydrated electrons to yiel.i anion as indicated by the decay of the 720 nm peak. The aniom- formed for example in pentafluoronitrobenzene and tetrafluoroben zoquonor.e anions have bands at 440 and 290 nm and at 435 and 304 nm respectively. The rate constants were determined from th RC - 11 .1 REFERENCES 1. G. V. Buxton, C. L. Greenstock, W. P. Helman and A. B. Ross, J. Phys. Chem. Ref. Data, J/7, 513(1988). 2. Lian C. T. Shoute and Jai P. Mitral, Radiat. Phys. Chem. 26, 739(1985). 3 D. B. Naik and P. N. Moorthy, J.. Chem. Soc. Parkin Trans. 2, 705(1990). Tablel. Rate constants of hydra ted electron with sustituted perfluoroaromatic compounds (pH 7). compound kxlO* compound kxlO / M -L c: ^- \ \ I 1 O / C6F5COOH 6.8 C6F4(COOH)2 2.8 C6F5NH2 8.8 CfiFsNHNH9 15.4 C6F5CN 9.4 C6F5SH 6.6 C6F5NO2 22.5 C6F4°2 24.9 C6F5CHO 12.6 C6F5OH 6.3 CgF5COCH3 11.6 ACKNOWLEDGEMENT: The author gratefully acknowledged the excellent job of Mr A. Chakraborty,, Chemistry Division Work Shop in fabricating the adapter for the sample flow system of pulse radiolysis. RC - 11.2 RADIATION IHDUCED REACTIONS OF OH/O" WITH 3,4,DICHL0R0T0LUENE Hari Mohan and J.P.Mittal Chemistry Division, BARC, Bombay 400 085 - INDIA Mohan Mudaliar and B.S.M.Rao University of Poona, Poona 411 007 - INDIA Summary Reactions of 'OH and O" with 3,4,dichlorotoluene are studied using pulse radiolysis technique. The treansient absorption spectra obtained with 'OH and 0~ are different in their characteristics. A constitent reaction mechanism is proposed on the basis of experimental findings Key words : Pulse radiolysis, OH adduct, abstraction Introduction: The reactions of the OH radicalCOH) with benzene and its substituted derivatives have been studuied extensively (Ref. 1,2) by pulse as well as steady state radiolysis whereas very less number of reactions have been investigated with the 0~ radical ion. The studies have shown that the'OH behaves as an electrophile and undergoes addition leading to hydroxycyclohexa- dienyl radical (HCHD). O^prefers abstraction of H instead of addition. The present work deals with the reactions of "OH and 0r with 3,4,dichlorotoJuene. Experiaeotai 3,4,dichlorotoluene( Aldrich Co.) used was of purum grade, The solutions prepared in triply distilled water were T saturated with nitrous oxidle(NaO). The reactions of 0 was studied at pH greater than 12.5. LINAC (8 MeV) pulse radiolysis facility available at BARC, Bombay, was used. The details if the set-up are published elsewhere (Ref.3). Results: In NaQ saturated aqueous solutions the e£ are converted to "OH by reaction (1). s OH 4. 0H The transient absorption was measured from 260 nm onwards. This spectrum is presented in Fig. 1 along with that obtained at high pH of 12.5. At this pH the reacting species is 0T (reaction 2). OH + OH" The "OH undergoes addition reaction to give HCHD of 3,4.dichlorotolune having strong absorption in ultraviolet region. The spectra shows maximum absorption at 325 nm in well agreement with the previously reported values. The formation of the adduct was followed at 325 nm with varying ooncertrations of the substrate. The bimolecular rate constant for the OH adduct formation was found to be 1.2 x 10 dm*mol s"! RC - 12.1 OH H, OH Ci CI As shown earlier, at pH greater than 12.5 the only reacting species is the 0~ radical. The reaction of the 0T with the substrate is different from that observed with of 'OH. In this the abstraction reaction seems to be considerable along with the addition reaction giving absorption peaks at 265, 325 and at 410 nm. The peak at 265 nm is attributed to 3,4.dichlorobenzy1 radical obtained by abstarction of H from methyl group. 'CH-, 4- The overall rate constant for the reaction measured at 330 nm was 8a 8.2 x ID8 dmJ mol"' The comparatively lower rate constant is quite obvious due to the slower abstraction process. J4.0C! +DH : 1 003 34DCt+-CT : 2 300 Fig. 1 : The transient absorption spectra obtained for the reactions of 'OH and 0" with 3,4,dichlorotoluene. REFERENCES 1. S.Steenken, J Chetn. Soc, Faraday Trans.1, 83, 113 (1987). 2. G.V.Buxton, J.R.Langan, and J.R.L.Smith, J.Phys.Chem., 80, 6309 (1986) 3. S.N.Guha, P N.Moorthy, R.Kishore, D.B.Naik. and K.N.Rao, Proc. Indian Acad. Sci. (Chem. Sci.) 99, 261 (1987) 4. Hari Mohan, J.P.Mittal, Mohan Mudaliar, C.T.Aravindakunar, and B.S.M.i'ao. (subniitted to J. Chem. Soc. , Perkin Trans. 2). 5 J.P.Mittal and E.Hayon, Nature, Pnysical Science, 240, 20 (1972 ) . RC - 12.2 PULSE RADIOLYSIS OF BENZAMIDE IN AQUEOUS SOLUTIONS Arpita K., L.V. Shastri and J.P. Mittal Chemistry Division Bhabha Atomic Research Centre, Trombay, Bombay-400085 SUMMARY: Spectral characteristics and self-decay kinetics of transient species from aqueous benzamide have been determined. Rate parameters for electron-transfer from radical anion and ketyl species of benzamide to different substrates and from other transient anions to benzamide have been evaluated. (Key words: Pulse radiolysis, benzamide, transient species, absorption spectra, electron-transfer, reaction kinetics). I. INTRODUCTION: Radiation Chemistry of amides in aqueous solutions is interesting, owing to the presence of the biologically important -CONH- group. Electron capture rates of amides vary by orders of magnitude, depending upon the molecular structure, reaching diffusion-control limit in the case of benzamide. However, very little work has been done on the chemical properties of the transient derivatives of amides. II. EXPERIMENTAL: 25-50 ns pulse widths of 7 MeV electrons was employed for the pulse radiolysis. Details of kinetic spectrophotometry were as described elsewhere (1). Solutions were prepared with A.R. grade Chemicals and triple distilled water and deoxygenated by bubbling oxyger.-free N2/N2O. pH was adjusted with NaOH or phosphate buffer. Extinction coefficients (€) were estimated by reference to the absorbance of (CNS)'2~. III. RESULTS AND DISCUSSION: The spectral characteristics and selfdecay kinetics of the varirous transient species derived trom benzamide, as also the experimental conditions are summarised in TABLE I. It was noted that values of £ in the transient spectrum due to the reacticn of e" with benzamide were very different from those of benzoyl radical, C6H5CO. This was contrary to an earlier suggestion (2) that e" attachment to benzamide might cleave the C-N bond, to give benzoyl radical and ammonia. We also verified, by gamma radiolysis, that the yields of ammonia were very low, G ^ 0.4. The observed spectrum was therefore assigned to the radical anion, C6H5C(o~)NH2. Our values of £ at 315 nm for the anion, values of £ and 2k for the corresponding ketyl radical, are substantially different from those reported in the literature, good agreement being found in the case of the OH adduct (3,4). and 2K values for the 6~ adduct were not available in the literature. Results on the kinetics of some of the electron-transfer reactions, obtained in our work, are summarised in TABLE II. These were obtained by following the decay of the benzamtde transient or the growth of the product species. From our results, a redox potential of d -2V was estimated for the CgH5CONH2/C6H5C(O~)NH2 couple. RC - 13.1 TABLE I Spectral data and self-decay kinetics of transients from aqueous benzamide (0.001-0.002 mol.dm"-') 3 l 3 Species Exptl. Conditions /^ax(nm)£(dm mol 2k(dm 7 cm"1) mol~1s"1 l.Radical_anion pH>ll,0.2-1 mol.dm-3 315 1.85+0.5xlQ47.2xl0? :i C6H5C(O~)NH2 t-butanol, N2 450 4.5 + 0.1xl0 2.Ketyl radical pH=5.8,0.2-lmol dm -3J 430 1.6x10" 2.2x10- Cf-ficC t OH ) Nri7 t-butanol, No 3.OH adduct pH = neutral-11 , N2O 345 3.3x10- 1.0x10- 4.0" adduct pH=13, NoO 335 1.7x10- 5x108 TABLE II Kinetics of elect ron-»ransfec . (^Bzam~ and BzamH ot.and for the radical anion and Ketyl radical from benzamide, respectively}. Reactants pH Rate constant K(dm rnol" Bzam~ + nitromethane 11 SxlOj " + nitrobenzene 4.5x10; " + benzophenone 2.7x10 BzamH + nitromethane 5.8 6.5x108 " + nitrobenzene 3x10; It + benzophenone 1.6x10 " neutral 2.5x10 C(O")CH3 + benzamide 13-13.7 2.8x10 c63cCN + 11 6.5x10? iU Acetamide " + 11 ~10 REFERENCES 1. S.N. Guha, P.N. Moorthy, K. Kishore, D.B. Naik and K.N. Rao, Proc. Indian Acad. Sci . 99^ 261 (1987) 2. E.J. Hart and M. Anbar, 'The hydrated electron1, Wiley Int. Science, M.Y., (1970) page 137 3. E. Hayon, T. Ibata, N.N. Lichtin and M. Simic, J. Phys. Chem. 15^, 2072 (1972) 4. E. Hayon, T. Ibata, N.N. Lichtin and M. Simic, J. Am. Chem. Soc. 93 5388 (1971) KINETIC PARAMETERS OF ACIi) HYDROLYSIS OF HYDROFEROX1DKS OK DNA-MODEL SYSTEMS : A RADIATIOM CHEMICAL STUDY Shailesh Phuikar1, B.S.M Rao and Clemens von Sonntag 1: Department of Chemistry, University of Poona, Pune - 111007. Ind ia 2: Max Planck Institut f. Strahlenchemie, Stiftstrasse 34-36, D-4330, Muelheim a.d. Ruhr 1, FRG SUMMARY: The rsjt.t- of hydrolysis of hydroperox ides formed in the radioly.-.is of .JI loopropyl ether (DIPE) and 2 , 5-d imethy i tetrahy iruf unu. (DMTHt'j using the oxygenated millimolar aqueous solutions it, found tu be low av near-neutral pH. These hydroperoxides wert found to undergo sn acid-catalysed (pH - 4-6) hydrolysis with stcund order rate constant of 1.9 and 1.4 M"* s"1 respectively. In case of DIPE, the protonation occurs approximately in J I ratio at both of the peroxyl oxygen atoms. < Keywords : Hydropero.'.ides , Di-isopropyl cth^r, 2 , 5-d lmei.liy 1 tetrahydrofuran, Acid Hydrolysis) I. INTKODUCTION : Hydroperoxides are normally encountered in the steady - state radiolyyis of uxygeri saturated aqueous solutions of DNA-model systems (Kef. 1.2). The reaction steps leading to the formation of such ooaipounds are well-established (Ref . 3). As often is the case, little is known concerning' the identity of these hydroper- oxides However, they have been suggested to be formed (Ref. 4,b < and many investigators: have reported the formation of more than one hydroproxide and their yields are denoted as total peroxide yield (Kef. 4). Hydroperoxides are known to undergo acid catalysed rearrangement into a carbonyl conpound and, usually, an alcohol (Ref. 6). Here, we report our preliminary data on the identification and yields of hydroperoxides formed, by radiation chemical methods, from DIPE and DMTHF. II. EXPERIMENTAL Di-isopropyl ether <. Loba Chemie, Bombay) was AR Grade and used as received. 2,5 di-methyl tetrahydrofuran was AR Grade and used after distillation. Catalase from beef liver ( Boehringer Mannheim i was used to differentiate the yields of H^O^and ROOH 'a hydrope.oxide ) . The irradiations were done using a 60Co-f- source with a d se rate of 0.44 Gy s1 . The substrates being low-boiling and to reduce their oxidation, water was pre-saturated with 0a (Induyinsl Oxygen Ltd.) and the concentration of the substrate made to ,;«iM . pH were adjusted by HftS0*/Na0H. ROOH and H^Oa yields were measured by the method of Allen et. al. (Retr. 7). III. RESULTS AMD DISCUSSIOH : In V- radiolysis of water e^, H-atom and OH-radical are formed as primary reactive radicals. In the presence of 0a , e^ and H-atom RC - 14.1 are sr enged by O^to form Oj, while OH does not react with Oabut prefers to abstract a substrate-bound H-atom. The substrate radical thus formed, is converted, at diffusion controlled rate, to the corresponding peroxyl radical (Ref. 8). ROOH is produced by an electron transfer reaction between Oj, and the peroxyl radical. In case of DIPE, three hydroperoxides are reported with a foial S-value of 1.8. When 0^saturated 2mM DIPE and 2mM DMTHF (pH - 6.5), 9 -irradiated solutions are allowed to hydrolyse in pH--range 4-6, the respective total ROOH undergo pseudo-first order rearrangement. The second order rate constant for the decay of ROOH. calculated from the plots of kObsd vs proton concentration, are 1.9 and 1.4 Ms respectively. Schuchmann et. al. have suggested 2-isopropoxypropyl hydroper- oxide as the principal ROOH in oxygenated DIPE solutions. Here, if acid-hydrolysed, proton has two sites of attack; leading to the formation of methanol, isoprcpanol and acetic acid, if the attack is at the 0-atom adjacent, to hydrogen ( reaction 1) or acetone and isopropanol, if the attack occurs at 0-atom attached to isopropoxypropyl group ( reaction 2). It leads to the formatiop of H^O^, which comes as the part of hydrolysis. Me + t-p*0-C-0OH H -A Me + We have measured the excess yield of H^Oi , arising from reaction 2, which amounts to G-value of 0.5, while the consumption of ROOH is 1.2. It follows from-this, that the reactions 1 and 2 contribute approximately 50% each to the acid-catalysed rearrangement of this hydroproxide. Acknowledgement: S.P. is grateful to DAAD, Bonn for financial support. IV. REFERENCES : 1. H. Schultze, D. Schulte - Frohlinde Z. Natur,29b,91 (1974). 2. M. N. Schuchmann, C, von Sonntag, Z Natur, 33b, 329 (1978). 3. C. von Sonntag, The Chemical Basis of Radiation Biology, Taylr and Francis, London (1987). 4. M. N. Schuchmann, C. von Sonntag, Z Natur, 44b, 495 (1987). 5. M. N. Schuehmarin , C. von Sonntag, . Phys. Chem., 83, 780 (1979 > . 6 D Swern, Organic Peroxides, Vol. I Wi ley-Inteiscience, New York f 197U ; 7. A.O.All^n. i: .J . Hochanadel . J . A . Ghorm leyan , d T.W.Davis, J. Phys Cheur., 56, 575 (1952). 8. G.E.Adaini;, R.L.Willson, Trans. Faraday Boc. 65, 29bl (1969). KC - 14.2 PULSf-: RADIOLYT1C REDUCTION OF AMINO AND HYDROXY ID 1 SUBS r ! TUTED ANTl 1RAQUI NONES H.Pal, D.K.Palit, T. Mukherjee and J.P. Mittal Chemistry Division, ^habha Atomic Research Centre, Tromb&y, Bombay 400 085 SUMMARY On e-electrun reduction of 1 - am i n o - 4 - h y d r o x y - 9 , 1 0 -• a n t h r a - quinone (AHAQ) and 1 , 4-d i ami no-9, 1 0-an t Iiraqi: i norm (UAAQ) ir various matrices has been investigated by >J lee iron pulse radiolysis. Spectroscopic and kinetic parameters, acid dissociation constants (pK. } of the reduced stni i q u 1 nors« radicals and one-e I uc i r on r clue t i un 'po ten t i a I fur Alt,-V> have l>e<;n nriasured. I. INTRODUCTION The substituted an t hr a.ju i nones arc. .:i*:d".l compounds for a number of antltumo'w agents, vat dyes ynd some quinones In living systems. Thus their free radical chemistry is of vast practical importance. Th e substitution u( the hy d rox y group • > f 1,4-dihydroxy- Q^lO-ant hraqu i none ( qu i n i i a r i n , QZ) by ammo group, considerably affect the radiation chemical behaviour. In this communication, the spec t roscop i c, kinetic, and redox characteristics of the semiquinones of AHAQ and DAAQ produced by one electron reduction, have been reported. I I . EXPERIMENTAL Quinones (Tokyo , III. RESULTS AND DISCUSSIONS Spectroscoplc studies hi pure isopropanol and in mixed solvents at different pM indicate t tie existence of at least two forms ot radicals foi AliAQ and atleast three forms of radical a for DAAQ in the pH range oi 1 to 11. The semiquiinone absorption spectra In mixed solvent, of AIIAQ at pH 1 and of DAAQ ai fuH b, compare very closely to the respective spectra RC - 15.1 In pure lsopropanol and are attributed to the neutral form. Following the variation of absorbance with pH, the pK values for the semiqu1 nones have been estimated. The equilibria involved are - (a) AHAQ; 2.3 7.9 (b) DAAQ; QH, QH The spectroscopIc characteristics of the radicals obtained in alkaline aqueous t-butanol and formate are similar to those in alkaline mixed solvent and the same radical forms tn all the cases. The spectroscopic and kinetic characteristics of the radical 3 are listed in Table 1 and 2 respectively. The one-electron reduction potential values (EE )) in aqueous formate solution was estimated to be -405 mV for AHAQ at pH »j 11.0, using methyl bipyrldyl system as the reference. The E value could not be estimated for DAAQ due to its insolubility in aqueous media. Table 1. Spectroscoplc, Acid-base and Redux characteristics of the setnlquinone radicals in mixed solvent 1 <^ii(lone Amax'nm( mW'cnf ) PKa(l) pKfl(2) E|J ,mV + QH2 QH • _ Q 388(5.8) QZ 410(11.6) 3.3 -395 475(13.7) 390(6.3) AHAQ 410(6.8) ~ 6.3 -405 500(13.1) DAAQ 365(7.4) 390(6.5) 385(6.0) 2.3 7.9 — 465(6.9) 490(8.0) 490(13.5) 580(6.1) Table 2. Kinetic parameters for the semiquinone radical.. 9 3 1 Reaction Rate constant, 10 dm mol"V (pH) AHAQ + (CH)3)2COH 2.06(1.5) 1.63(8.4) i.70(lL0) — DAAQ + (CH3)2d0H 1.06(11.3) 0.92(5.5) 0.90(8.0) 0.93(11.0) — AHAQ • e~q — — — 50.0(11.0) 53.0(14) AHAQ + COjT — — 1.13(11.0) 1.27(14) 2(AHAQ-Semiqulnone) 1.4(1.5) 6.8(8.4) 6.7(11.0) — 2(DAAQ-Semiqulnone) 1.1(1.,3) 1.6(5.5) 2.5(8.0) 2.4(11.0) — RC - 15.2 ELECTRON TRANSFER REACTIONS IN AQUEOUS SOLUTION OF DOPAMINE? A PULSE RADIOLYSIS STUDY D. K.Mai ty and Hari Mohan Chemistry Division Bhabha Atomic Research Centre Trombay, Bombay 4OO O85, India SUMMARY: Pulse radiolysis technique has been employed to detect and characterize the species formed by one electron oxidation and reduction of dopamine. The one electron oxidants used are Cl2, Br- and f~ and the reductants are e and H atom, Spectral and 2 aq kinetic parameters of the transient species have been de'-.ermined. (Key words: Dopamine, 4-[2-Aminoethyl]-l,2-benzenediol, Tyramine, Pulse radiolysis, aqueous solution) I INTRODUCTION: Dopamine is an important biological molecule. The pathological significance of dopamine and its functioning as a neuro-transmitter have stimulated interest in its chemistry. The present 3tudy focuses on the one electron oxidation and reduction reactions of the molecule. II EXPERIMENTAL: Dopamine used was obtained from Sigma chemicals (H 8502). All solutions were prepared in nanopure' water. Kinetic spectrophotometric experiments were carried out using high energy electron pulses (7Mev, 50 ns) from a linear accelerator (dose~lxl0 eV cm per pulse). All the reactions were carried out at room temperature (24-25°c). III RESULTS AND DISCUSSIONS The reaction of e with dopamine (D) is studied by aq monitoring the decay of e , formed on pulse radiolysi3 of N aq _ Z saturated aqueous solution (pH=6.0, t-butanol=2.0 mol.dm , X=700 nm) containing different concentrations of dopamine. The decay becomes faster and of first order. RC - 16.1 The bimolecular rate constant is determined from the slope of the linear plot of pseudo first order rate with dopamine concentrations and the value is 2.5 x 10 dm . raol .s The time resolved studies shewed the formation of a new band with X. =355 max nm. This band is assigned to radical anion of dopamine, D as it appeared on decay of e .It decays with first order kinetics with aq t =144 fjs. The H atoms, formed on pulse radiolysis of N saturated aqueous solution (pH=l.ft, t-butanol=0.5 mol dm ), are also observed to reduce dopamine to its radical anion (A =355 max nm). The bimolecular rate constant, as determined from the growth of 355 nm band is 4.2x190 d3m mo-i-l si . The oxidation of dopamine is studied from the decay of the transient bands of specific one electron oxidants such as Cl , Br and I for various dopamine concentrations. In presence of dopamine, the dacey was faster and of first order. The time resolved studies showed the formation of a new band with X =300 nm, which is assigned to ita radical max cation (D ). The radical cation decays with first order kinetics with t =116 fs. The physical properties of radical cation and anion of dopamine are shown in the table. From these studies, it is concluded that dopamine is able to undergo one electron oxide, ion and reduction reactions. Table Physical properties of radical anion and cation of dopamine. A (nm) Forma$ion_ rate Reaction 3 ™i* -i cons(dm rool s ) max (dm mol cm ) D + e" —> t~ 355 3650 2.5xl08 oq + D + H —> D" + H 355 - 4.2xl09 D + Ci; ->D% 2C1~ 300 2398 1.6xl09 8 D + Br2 —>D% 2Br- 300 2238 1.0X10 D + 21" 300 - 7 2 1.3xl0 RC - 16.2 PULSE RAPIOLYSIS OF HYPOXANTHINE IN AQUEOUS SOLUTION Hari Mohan and J.F.Mittal Chemistry Division, BARC, Bombay 400 085 - INDIA C.T.Arav indakumar and B. S. M.Rao Department of Chemistry, University of Pootm, Poona 411 007 INDIA SUMMARY i The kinetic parameters for the reactions of the hydrated electron and the reducing r adicals derived from n&thanol, ethanol and isopr opanol ttith hypoxanthine at different pH are evaluated. The tr ans ient absor ption spectr a of the electron adducts formed under different pH conditions are interpreted in torws- of th& protonation eqtii 11 ibr t'». INTRODUCTION: The radiation cheroistry of nucleobastiB is very important as they form ideal DNA-mr.id.eJ systems to understand the underlying reaction mechanism. Among the primary radicals of water, OH and the hydrabed electron are very reactive towards purines{Ref. 1,2). The reactions of hydrated electron are mainly studied with adenosine(Ref, 3,4). The information on the reactions of OH as well as the hydrated electron with the purines has been obtained from pulse radiolysis with optical and conductance detection and the redox titration. The present study deals with the reactions of the hydrated electron and some reducing alcohol radicals with hypoxanthine(Hx)• EXPERIMENTAL: Hypoxanthine {Merck) of high purity was used and the alcohols ( roethanol, ethanoi, and i-propanol ) were of A.8. grade. No and N2O gases of IOLAR grade { Indian Oxygen) were used. Solutions were prepared in triply distilled water. For obtaining the kinetic parameters the substrate concentration was varied between 0.4 5 mM. The pH of the solution was adjusted using various buffers. Pulse radiolysis experiments were carried out using high energy electron pulse(7 MeV) from a linear accelerator. The details of the facility are described ©lsewhere(Ref. 5). The absorption of the transients, monitored at appropriate wavelengths, was recorded with time. RESULTS AND DISCUSSION : The kinetics of the reactions of the hydrated electron with Hx at different pH were determined by monitoring the decay of the hydrate*! eJectron at 720 nro as e function of the subfitrate concentration. The build-up ofthe electron adduct was also monitored {t;ee Table 1) at appropriate wavelength. At neutral pll, the three major observations are following: 1) There is an immediate build-up of ttbsorbanoe at 310 nm from the end of the pulse, 2) The rates of the formatior« of the electron adduct at 310 nm and of the decay of the hydrated electron at 720 nm are similar,' RC - 17.1 3) Linear dependence* of k(ob*:d) on suhst.rnt.e eonoentration is observed. A decrease by an order of magnitude in the rate constants has b*?n obsnrv«i for the deeay rate of the hydrate! electron with Hx at pH- 10 and at i»U 3 3.5 ooraparf-d to that observed at pH 7. This de<-re«K** in the rotfi constant with increasing pH is attributed t<, the electrostatic repulsion resulting from the deprotonation at the H i tew M i irxi N 9. Reaotions wi U« < Hydrwty Alkyl KndicaJu : K'>r Uin ;;tu k{i propano] ) > k{e(hanoJ ) > .thwriol ) Table 1 "- "foax(nn»), ^max(M <:m ) «nd rate <;orii>tantK(M s ) observed in the pulse rad iolytn s: <>f Hypoxanthine. Keaetant .v max )0 6.5 305 3300 J. 4 X io 10.0 310 6190 1. 5 X 8 0.5 300 HfiHO 1. 3 X 10 470 1 5 10 CH3CHOH 0.5 300 5600 1 X 6 0.5 300 2460 2. 9 X 10 REFERENCES. 1. von Sonntag.C. , The eheanoa! ba.s-i of Kadintion Biology, Francis and Taylor (1987) 2. Ste*mken,S., Chem. Hev., 89, 503 (19H9). 3. Hissung.A., von Sonntag, RC - 17.2 Effect of Nickel(II) and Nickel(II) Complexes on the y-Radiolysis of Thymine S.Chakrabarti, P.C. Mandal and S.N. Bhattacharyya Nuclear Chemistry Division Sana Institute of Nuclear Physics 1/AF,Bidhamnagar,Calcutta-700064. Summary y-Radiolysis of thymine(T) in presence of Ni(II) ions and its complexes were carried out in dilute aqueous solutions at neutral pH. Unlike Cu(II) or Fe(III) ions ,Ni(II) ions have very little effect on the radiosensitivity of thymine .Formation of glyoxalic acid and formaldehyde in the presence of Ni(II(complexes, indicated the oxidation of the metal complexes to Ni(III) species by C--TOo H radical species, (KeywordsiT'iyminejRadiolysis,Nickel ions.TLC ) • • I.INTRODUCTION With a view to understand the mechanism of radiosensitization of hypoxic cells by transition metal ions,model radiation chemical studies were undertaken using thymine as the target molecule. Here effects of Ni(II) and its complexes with aminopolycarboxylic acids on the radiosensitivity of thymine were investigated. II.EXPERIMENTAL Samples of 2mM thymine containing 2- C-thymine were irradiated in the presence of different nickel(II) compounds asing Co-yrays. The products of radiolysis arising from the degradation of thymine and the unchanged thymine were neasured after separating them by bidimensional thin layer chromatography using the solvent systems :chloroform:methanol:water (4:2:1) and ethylacetate:isopropanol:water(75:16:9). Formaldehyde was measured by chromotropic acid reagent whereas glyoxalic acid was determined with 2,4-dinitrophenylhydrszine. III.DISCUSSION In y-irradiated dilute aqueous solution of thymine different radical adducts,viz,TH,TOH,T~are formed .Radiolytic study in presence of t-butanol,in OH radical scavenger, shows that TOH radical plays a predominant role in the degradation of thymine in the absence of metal ions. When the radiolyais was carried out in the presence of Ni ions, the degradation of thymine was found to be similar to that observed in the absence of metal ions (Table-I). However, formation of formaldehyde and glyoxalic «eid in the radiolysis of thyaine in the presence, of Ni1'-aminopolycarboxylic acids suggests that a part of TOH (e.g,C,-TOH) which is oxidising4 in nature oxidises Ni11 complexes to Ni"1 RC - 18.1 species which gives rise to formaldehyde and glyoxalic acid. Thymine de-radation products , however are formed in the sama way as that formed in the absence of metal ions. :!~>—>6-TOH" + Ni1 "(~ -CO Ni H 0 HCHO + H n Nin(R,NCH,COO~> > Ni l * (R,NCH,COO""> ' ^") + H+ 2Nin(R,NCHCOO"! CHOCOOH COO~) HCHO Table I Yields of different products in the radiolysis of thymine (2 x 10~ K) in the presence of different Mi(II) ions T T+ T+ T+ T+ T+ Saturation only HI SO NiEDTA H J I DA H i(I DA)^ NiHTA Argon 1 -9 1 .6 1 . 6 1 . 5 1 . 5 1 .6 G<-T\ nno 3.9 3. 2 2 ~t 3. 2 3. 2 3.3 ** "7 2. 6 2 .7 2. 8 2.6 2.8 °2 Argon 0.5 0. 1 0. 1 0.0 G(HCHO) HOO 0. 6 0.2 0.0 0. 1 0.0 0.0 0.0 0.0 o2 .\ r g o n 0. 2 0. 0 0. 2 0. 1 6(CHOCOOH) HOO 0 .5 0. 2 1 . 2 0.6 0.0 0.0 0.0 0.0 °2 IV.REFERENCES 1. S. N. Bhattacharyya, P. C. Manotal *nd S. Chakrabarti, Anticancer Res., 94121.(1989). 2. S. N. Bhattacharyya, P. C. Handal and S. Chakrabarti, Bull. Chea. Soc. Jpn., (in Press). 3. S. N. Bhattacharyya, P. C. Handal and S. Chakrabarti. Radiat. Phys. Cit«a. (in Press). 4. S. Fujita and S. Steenken, J. Am. Chem. Soc, 103, 2540,(1981). RC - 18.2 GAMMA FUDIOLYSIS OF Ni(II} COMPLEX OF METRONIDAZOLS P. C. Mandal, M. Basu Roy and S. N. Bhattacharyya Nuclear Chemistry Division Saha Instituta of Nuclear Physics l/AF, Bidbannagar, Calcutta-700 064 SUMMARY s v -radiolysis of aqueous solution of Ni (II) metronidazoia (Ni M) was carried out under different conditions. The complex undergoes OH radical induced oxid=»tive denitration resulting in its decomposition. No denitration was observed due to the reac- tion of e '* with the metal complex. INTRODUCTIO'N Although nitrohetrocyclic drugs viz., metronidazoia/ misonidazole, etc, are good radiosensitizers, the idea of using them as metal complexes is gaining importance. This is because the metal ions could preferentially bind at the target and incre- ase the local drug concentration at the reaction sited, 2). However, before studying their effects in the cells, it is nece- ssary to study ch','ir radiation chemical behavior-?. In thi3 re- port, studies on some aspect of its radiation chemistry in dilute aqueous solution have been undertaken. EXPERIMENTAL 1:1 Ni(ll) complex of metronldazole {Hi1 M) was prepa- red by mixing equimolar amounts: of aqueous solutions of metroni- dazole and Ni(II) sulphate. The loss of chromophore, G(-NiI:IM), was determined spectrophotometrically at: 320 nrn. HNO2 was deter- mined spectrophotometrically using o RC - 19.1 The relative yields of HNO2 is much higher in the radiolysis of N±**M as compared to the free aietronidazole [jGfHNC^}1^ 1.6J under similar conditions. This may be explained by assuming that at least a part of the OH radicals attacks tho nvatal centre to give Ni(IIl) species. Further studies in this regard and on the for- mation of ligand degradation products are in progress. TABLE-1 G~values of the products formed in the radiolysis of Ni11 (M) (5 x 10"** moi dm~J) under different conditions. Condition GC-Nl11!*) G(H1"0O) Argon saturated 3.4 2.3 N20 saturated 5.5 3.3 3 J>.. 4 moll dm"" t—but-buOCH Q « + Argon saturated REFERENCES 1. Farrell, N., Gomes Carneiro, T. H., Einstein, F» N.B., Jone S, T., skov, K. A. Inorganics Chim. Acta., 92, 61 (1984) 2. Farreil, N. P., Skov, K.A. Radiat. Res. 91^, 378 TT982). 3. A. I. Vogel, A text book of quantfetfatoto inorganic analysis (ELBS, London), 1962, p-784. RC - 19.2 Cu(II) Induced Radiosensitization of Cytosine P. C. Mandal, K. Chabita and S. N. Bhattacharyya Nuclear Chemistry Division Saha Institute of Nuclear Physics l/AF, Bidhannagar, Calcutta-700 064 Summary Radiosensitization of cytosine by Cu(II) ions has been investigated in dilute aqueous solution using Co-60 f-rays. In N2O saturated solution the base degradation yield, G(-cyt-) uas found to be *• 3.1 and ^4.3 in the absence and in the presence of Cu(ll) ions respectively. However, in argon saturated condit' i Cu(ll) ions induced radiosensitization of cytosine was r.t obser- ved. The role of Cu(II) ions in increasing the radiosensitivity of cytosine has bean described as due to electron transfer betw- een Cu(II) ions and cytoaine radicals. Introduction It is well established(1) that ionizing radiation induced damage to the cell is principally due to the alteration and/or damage of the constituent bases of the DNA. Recent studies have shown that various metal ions enhance the base degradation. It has been reporteo previously that Cu(II) compounds induce sensitized degradation of uracil(2) and thymineO)., The present report describes the effect of Cu ;i) ions on the radio- sensitivity of cytosinejan important DNA component. Experimental 5 ml of 2 x 10" M cytosine solutions were irradiated with Co-60 *• rays in the presence and absence of CuSO^. However, before irradiation the solutions were saturated with argon or N2O gas by bubbling the respective gases through the experimental solutions for 30 minutes. The decomposition of the base was ascertained after separating the products of radiolysis from cytosine by HPLC using ODS. C-18 chromatography column. The flowrate was 1 ml/min. The absorbances of the column eluents wvre monitored at 260 nm. The Cu(I) ions were complexed with neocuproin and absorbances recorded at 457 nm. Results and discussion In argon saturated solution Cu(II) ions have no influence on the radiosensitivity of cytosine. However, when the radiolysis was carried out in N20 saturated solution the base degradation yield was found to increase from 3.1 to 4.3 due to the presence Of Cu(II) ions. The results are shown in table 1. It;is also observed that in the presence of Cuu-) ions in N2O saturated solution G(-cyt) value is nearly double that RC - 20.1 observed in argon saturated solution. From the results it may be assumed that OH radicals are mainly responsible for the base de- gradation in the presence of Cu(II) ions. This has been further supported by the fact that no base degradation was observed when 3 a mixture of 2 x 10- M cytosine and 5 x 10-^M CUSO4 was radio- lysed in the presence of 0.4M t-butyl alcohol in argon saturated solution. In dilute aqueous solutions "*-irradiation of cytosine produces CytH, CytOH and Cyt." (2). From the measurements of Cu(I) yields JUnder- different conditions it may be suggested that the electron transfer takes place from the CytOH, CytH and Cyf radi- cals to th3 Cu(ll) ions. CytH + Cu(II) > Cu(I) + CytH+ CytOH + Cu(II) •—» Cu(I) + CytOH+ Cyt" + Cu(II) .—» Cu(l) + Cyt CytOH+ —* products. In the absence of Cu(II) ions highly fluorescent product in formed in the radiolysis of cytosine. This fluorescence inten- sity incraases with increase in dose rate. From the earlier studies(4) it appears that the product might be dimetic one. However in the presence of Cu(ll) ions the dimer formation is presented. Hence the increase in the Oi (-Cyt) value in the presence of Cu(II) ions might be due to the increase in the yields of monomeric products. Further work on the product ana- lysis in under progress. Table 1 G-values of degradation of cytosine and formation of Cu(I) ions in the radiolysis of cytosine. Condition G(-cyt) G(Cu(D) A B Argon saturated 2.5 2.31 5. 18 N20 saturated 3.1 4.3 5. 10 Argon saturated 2. 9 + 0.4M t-bctanol —4 A, in absence of Cu504 ; B, in presence of 5 x 10 M CuS04 Reference 1. Hutterraann J, Kohnlein w, Teoule R and Bertinchamps A.J, Effects of ionizing radiation on DNA, Berlin, Hidelberg,, New York, springer (1978). 2. Bhattacharyya S. N.and Mandal P. C, J.C.S. Faraday Trans-1, 79, 2613 (1983). 3. Bhattacharyya S. N, Mandal P. C. and Chakraborty S., Anticancer Research, £, 1181 (1989). 4. Mandal P. C. and Yamamoto O., Biocherr.. Int. 1_1, 197 (1985). RC - 20.2 RADIATION EFFECTS ON DIHYDROOROTATE DEHYDROGSNASE IN AQUEOUS SOLUTION A.Saha. P.C.Mandal and S.N.Bhattacharyya Nuclear Chemistry Division Saha Institute of Nuclear Physics 1/AF Bidhannagar. Calcutta-700064. Summary Tae effects of the selective free radicals. (SCN) ' Bro and 'I GH the ity of dihydroorotate dehydrogenase in aqueous solution at pH 6.5 were studied. The results implicate the possible involvement of cysteine and Cyrosine residues in the activity of the er.'.yme. Further, the effect of y-radiation on the Kinetic parameters V and K of the enzyme were max m investigated under difrerent gaseous environments. This study suggests that OH rad.ral possibly causes a general denaturation of the enzyme whereas H atom r^ctn with iis substrate binding site. words: dihydroorotate dehydrogenase.?-radiolysis) I. INTRODUCTION In continuation to our previous study on the radiation induced of dihydroorotate dehydrugenase the role of the selective :<:aJM. "KSCN) 'Br, and "I, were studied. Further, the changes in the • -*iamoters(K and V ) due to 7-radiation were investigated in dilute m ttsax sol ill" ion. H.KXPEK111F.NTAL hi h\"'r.<>orotate dehydrogenase purchased froe Sigma Chemical Co. was 11 radi it^i] with y-rays (do.^e rate » 3.3 Gy/»in) in aerated, argon and N,0 : III.RESULTS AND DISCUSSION The species responsible for inactivation of the enzyme on ^-irradiation in annaous solution at pH 6.5 were previously determined to be OH and H. To ob.a;it more information on the mechanism of inactivation selective radical anions, "(SCN)~. *Br" and "I. were used as reacting species. '(SCN), had little effect on the activity of the enzyme whereas both "Br. and "I, had crr.iuuinc«»d effect. Considering the loss of enzymatic activity and the rate data for the reactions of these radicals with different RC - 21.1 (2) aminoacids . it waa suggested that tyrosine arnd cysteine residues are the crucial aminoacid residues for the activity of the enzyme. The radiation inactivation might '.iia.nge the kinetic^ parameters of the enzyme.The values of iiichaelis-Menten constant K and maximum velocity V m max for unirradiated and irradiated dihydroorotate denydrogenase were determined under different conditions (TABLE I). No chancre in K was observed when the m enzyme solutions were irradiated under aerated and N O-saturated conditions. it but V decreased in both the cases. This suggests that the inactivation proceir!* under these two conditions is not due to chemical modification of the substrate binding site, but rather is due to reaction of those residues which are responsible for maintenance of the conformation of the enzyme. When the enzvme solution was irradiated under deaerated condition, both K and V m max were form! to decrease. Unlike that in aerated and N O-saturaf:ed conditions the concentration of H atoms in argon saturated condition was higher because e was converted to H atoms in presence of 0.2M DhosDhate buffer.' aq Therefore, lowerina in K value under araon saturated condition miaht be due m to modification of substrate binding sites of the enzyme by the reactions with H atoms. TABLE I K and V values for dihydroorotate dehydrogenase using orotate as substrate m max after irradiation at doses for 50% inactivation. Irradiation conditions K (mM) V (;.M/min) m max Unirradiated 1.4 13.33 Air-saturated 1.1 8.33 Argon saturated 0.3 3.45 N,O-saturated 1.2 6.25 REFERENCES 1.Methods in Enzymatic Analysis. ed. H.U.Bergmeyer, Verlag Chemie International. Florida. 2nd Ed., vol.4, p.1963 (1981). 2.G.E.Adams.J.E.Aldrich.R.H.BisbyrR.B.Cundal.J.L.Redpath and R.L.Wilson. Itadiat. Res. . 49. 278 (1972). 3.S.Na RC - 21.2 RADiOt.Yf;i<; OF ACIUEOUS AZ1DE SOLUTIONS G.R. Dey, Kanial Kishuie, S.B. Si ivastav3 ami p.N. Hoorthv A j i p i i -' d Chemistry D i v i P i c i n Phahhji Atomic Research Centre irdiiibav, Bombay' 400085 SUMMARY In the rpHiolvis o,f aqueous slide snl'itions NH3 is produced by the igarijon of e".fl with hN3 and H-atoms with N3 and HN,. This is liuppoi ted by the 0> values of NH3 messured at various pH s in Ns ::•& I u ra • e>d &?. i dp solutions,. In air or oxygen saturated azide solution? Mil. is not lonimu, in these sohjti ons at high do^as ( :• SO K tad). NC)2 was rJ, Amiiion i <=> . •in I ues ) I. INTRODUCTION Azidp i on(N, ) in a wil Lnnwn OH i .-nil a I scavenger. f? . „ can leact with only MNj anil not wi Mi N, On the other hand H-al.oras react with both UN, ;.i..d N3 , r ••»»<•'• ii-n being f-^'er with azide as comparer1 to hyJra/nic acTid/l'. [JI Iiec/2/ it was; proposed that tho it'Mct. iuii nf H atnnis with N.5 |Hfiris '.o the foim?lion ot I-J.. Bnd H2 . Recpnl ly/.'i/ it has heen sh.iwn • n;- ( n i j.ij /i . 9 both e . <, and H atoms react wilh HN,/N, to give .:• in in cm i a as a i i na I product. Experiments havi-? been c;.i Ii f>d out a( various pH s to ni«asiir« 'he i sdinlytic yu'liJs or Nil,f II. bSPhJRIMHN I Al. A I 1 <;fit'in i ;.JJ| 1 .: rnjjloyed in the .t'idi III. RESlM.Tii Ann DISCUSSIONS In y i r i ..- ii i .» .) nitrogen :<•' t • >r. • t. e main 1 radiol vi. i- •• •- '- s expH.-ted j . « I;.. H.l j JIUI Hll3 N 3 ' • Oil N3 • OH ( 1 ) N3 t H (2 ) According to Alfassi ft a!/2/ reaction 1Z) ads to the formation of Hi and Hi i iui i c :. ' in the presence of H' . However, r even 11 y / 3/ it has been uiiierved that the reaction leads to the formation of Nflj and Hj is not. produced in the process. The proposed reaction scheme ,. s ; RC - 22.1 N 3 + H N2 (3) HN3 + e" H* (4 ) &H2 + N3 - + -* NHj + (5) 2N, • -* 3N, (6) We have measured G(H202) and G(N.H3) in /-irradiated nitrogen saturated 10" z mol dm"3 azide solutions at pHs 7, 4.9 and 2 and the results are given in Table I. It can be seen that G(NH3) is very close to the value of GH at pHs 7 and 2 viz. 0.55 p.nd 3.6 respectively, whereas at pH 4.9 it is closer to the value of GIe"«„ + H) viz. 3.2. The lower value at pH 2 can be attributed 3 1 to the slower reaction of H-atoms with KN3( k = 7.2*10* dm mol" s"'). From these results it can also be inferred that this reac- tion also leads to the formation of NH-. H N (7 ) NHa (8) Simalarly at pH 6.2 by converting e".„ to H-atoms in presence of high concentration of H2 P04 "" , it was possible to show that the yield of ammonia is equal to that of total GH. In air or oxygen saturated azide solutions at pH 7 no ammonia was formed again indicating the involvement of e".q and H-atoms in this process. N02" formation was observed in these solutions at higher doses(>50 K rads) which was found to increase with increasing dose. IV. REFERENCES 1. G.V. Buxton, C.L. Greenstock, W.P.Helman and A.B. Ross, J.Phys. Chem. Ref. Data, 17_, 513(1988). 2. 2.B. Alfassi, W.A. Prutz and R.H. Schuler, J.Phys. Chem. 90, 1198< 1986) . 3. D.T. Deeble, B.J. Parsons and G.R.A. Johnson, Radiat. Phys. Chem. , 36_, 487 ( 1990) . 4. J.A. Ghormley and A.C. Stewart, J.Amer.Chem.Soc., 78, 2934(1956) . 5. M.B.Shinn, Ind. Eng. Chem. analyt.Edn. , 1J3, 33(1941). TABLE I 2 3 Product yields in the radiolysis of 10" mol dm" N3 " solutions. System pH G(NH3 ) satd. 7 ti 11 0.53 4. 9 0 . 4 3.0 2 0 .6 2.8 N2 satd., containing 6. 2 0 .5 2.7 3 i mol dm" KH2PO, Air satd. 7 0 .5 ni 1 Air satd., containing 7 1 . 9 1 »!ol dm'3 t-butanol RC - 22.2 EFFECT OF GAM.A IRRADLATIC*: 01, THE NUCLEAR GHADI:. IOW EXCHAri G RESIN R.S. Lokhande and M.K- Ingle Department of Chemistry, University of Bombay, Vidyanagari, Bombey-400 098 S'JMMARY A gel type ^polystyrene sulphonic acid resin of nuclear grade in Li"*% Cr^* and Fe^"*" cationic form -jere irradiated ip to dose 12MGy using 60co source and their chemical yields as the result of radiolytic degradation are calculated. Kay Words : Radiation Effect/Nu clear Grade Ttesin I . INTRODUCTION Now a days ion exchange is an integral part of nuclear power plants and nuclear industry. Radioactive charged species transfer energy to an ion exchanger making the system more succeptible to rapid deterioration that results loss of exchan- gable sizes and decrossiinking of resin matrix. Hence quantita- tive knowledge of degr ^ed product in ion exchange syste'. ia of prime importance. II . EXPERIMENTAL 0.5 g gel type sulphonic acid resin of nuclear grade (Indion-225) in Li+» Cr^* and Fe3+ catioric foruis,was irradiated under aqueous condition upto dose of 12M Cr?+ > Li+. This is mainly due to diffe- rent nature of hydration shells. Fe'+ ion has leaser hydration shell hence more is the compact pair formation between Fe^+ ion and functional site. The same is totally opposite in the case of lithium. pH values and hence the HP^4 yie^d as roilliequivalents and cocentration of cations ( TABLE Iljincreasee as the gamma dose. Here Fe^+ ionic form gives lees chemical yields and the trend is Li+> Cr5+> Fe3+. This is because of compactness of metal ion to the exchangable site and large ionic radius of RC - 23.1 cation. Here Li+ farm of resin is found less stable towards irradiation ae compared to Cp+ and Fe^+ cationic forms. IV, REFERENCES 1. 'Ion Exchange Resin* by Kunni 2. Technical BuJLetin of Ion Exchange (India) Ltd., Ambemath. 3« V. G. Dedgaonkar and CM. Bhavsar, Int. J. Appl. Radiat. Isot., 32, 895 (1981). TABLE I Percent decrease in exchange capacity and pH values of aqueous effluents. DOSE Exchange Capacity pfi of aqueous effluent M$y # deerease + 5+ L Cr? Pe Li+ Cr5+ Fe3+ 0 0 .0 0.0 0.0 6.54 6.70 6.80 1 2 .5 2.0 1.7 3.60 3.80 3.90 2 11 .5 8.U 6.5 3.10 3.55 3.67 3 14 .6 11.7 y. 9 2.70 3.10 3.15 5 18• 3 14.0 12.5 2.46 2.70 2.81 7 21 .6 17.0 16.0 2.38 2.65 2.69 9 24 .8 19.5 17.0 2.31 2.50 2.58 11 28 .2 22.5 19.0 — — _ 12 29 .0 24.0 20.0 2.28 2.46 2.55 TABLE II Dose dependent concentration of cations in aqueous effluent as ppm DOSE Concentrat ion/ppm 3+ Li'h Fe 1 138 96 84 2 638 380 326 3 818 562 504 5 1026 666 634 7 1208 812 718 9 1388 934 872 11 1548 1072 980 12 1610 1124 10 26 RC - 23.2 IRRADIATION STUDIES ON DODECANE AND DIOCTYL SULPHIDE IN DODECANE USING ABSORPTION SPECTROSCOPY G. H.RIZVI AND P. R, NATARAJAN Rad ioohmniF; t cy Division* Bhabha Atomic Research Centre Trombay, Bombay-400 085, India SUMMARY— Irradiation effects on diuctyl suljV-ide in dodecane were studied as a function of radiation dose. The IR spectrum of the irradiated dioctyl sulphide sample showed two new peaks at 77O an 1 790 cm"^, which werr not present in the pure uni rrad i at eel sample. About 90% of the dioctyi sulphide was found to decompose at: an absorbed dose of 17 MR. (KEYWORDS: Irradiation, dodecane, riioctyl svwjhide, infra-red, absorption spectroscopy> I. INTRODUCTION In ord^r to -stimate sulphide content in dioctyl sulphide TABLE-l: Effect of irradiation on dioctyl sulphide S. No. Dose Opt ical dens i t y Degradat ion in MR at 307 nm of DOS 1. _ 1. 579 Nil 2. 1 1. .""76 Nil 3. 2 1. 409 10.8 4. 4 1.065 32.6 5. a 0.501 68.3 6. 17 0. 134 9 1.5 7. 20. 5 0.099 93. 7 RC - 24.2 RADIOLYTIC REDUCTION OF U(VI) TO U(IV) IN NITRIC ACID MEDIUM F<. San kar Fuel Rpprocessinq Division F'.k . that tacharyva Chemistry Division Bhabha Atomic ( SUMMARY Gamma radiolysis o-f uran/1 rtitraLe aoluhoji in nitric acid medium was carried out in presience of sul phani 1 I. INTRODUCTION s The radiolytic reduction of U(VI> to Udv1) iB known in Sulphuric acid and perchloric acid media. Whereas no such radiolytic: reduction ie reported in Nitric acid "tedium. Perhaps tho estimation of U(IV) in Nitric: acid and the role o-f radiolytic product HNOo make the investigation difficult with rasped to its reaction with U(IV). Thus a method of determining the radiolytic yield, SCU II. EXPERIMENTAL : (i> Irradiation of UO2(NO3>2 solution in Nitric acid medium : A 5 ml. aqueous solution containing l.O mM UD-:.> (NO-:;) 2 and 10 mM Sulphanilami.de in 0.4 M HN0--., was irradiated at different time intervals. Sulphamlami.de fVH1?* reported to be a good scavenqer far e aqt^H, and HNOj The solution was purged with Argon gas and was irradiated using Co " source (Dose r ate : 1. 182 * 10 ev\yin\irit) . (ii) Estimation o-f U(1V) in presence o^ U(VI) : The BLU(IV)3 was calculated by es^niating the concentration o-f U by, adding eiiCBss of Fa J then ectimating the generated Fe '" epectr ophotonietr i cal 1 y at S10 nrn by compl e>: i nu it with 1 , 1 0 Phendnthrol ine. The molar absor bane e of Ft1 *~—Phencjnthr oi i ne complex waa determined ae 10,62S M cm III. RESULTS AND DISCUSSION * On rcitJiDlysit ot aq.HNOj following radical and molecular products ware repot ted. -, e \a, H, OH, ^O^, NO.-5, H2O2, HN02 .--d> RC - 25.1 In dearaated acid medium jL,. | raC| converted to H atoms. Sines 01-1, HN(J-> and H2fJ2 ** * ® known to oxidize U(IV) it is necessary to scavenge these radiolytic products. Since Suiphani1 amide was reported to sci vetige OH radical and HNCIT by forming SAQH and diazonium ion which do not participate in the rerios reaction of U H + SA > BAH (2) OH' SA > SAQH (3) + + 3AH + UQ - -:••• U02 SA + H (4) +4 U +• 2i-;2C) (5) u H->Q 0*H t- 3H+ and GCU(IV) ] = 1 /-? USH , aq eH202:l The experimental value of 1. is in (jood anreement with that o-f calculated value. Reaction <2> to be very fast in comparison witith that of (H UO^+ ) and it i r, assumed that SAQH is not rreacting either with U UV) or U(VI). Recently Guha et al . ''"' reporeporter d G("U IV. REFERENCES t 1. P.K.Bhattacharyya and R. I). £>«i r.i , hit. J . Radi at.. Phya. Chem. ,3, 91-99,(1973) 2. P.K. Bhaltcicharyya, R.D.Sain:, and P.B.Ruikar, Int.v7.Chem. Kinetica, 13, 385-401 ,< 19S1.) 3. S.M.Buha, P.N.Moorthy and K.M.Rao, Radi .at. Phy». Chem, , 29, Ho. it, 425-428 , < J 987 > RC - 25.2 RADIATION CHEMTSTRY OF THE AQUEOUS ALUMINIUM NITRATE SOLUTION CD. Kalkar, R.B. Date Department of Chemistry, University of Ponna, Pune 411 007. SUMMARY : Radiolysis of aqueous aluminium nitrate solution is studied as a function of concentration in the range 10 M to ]0 M« The stable radiolytic product of nitrate radiolysis is nitrite. The yield of nitrite linearly increases with absorbed dose. The G (NO,,) values are determined at various concentrations of aJuminium nitrate. A suitable mechanism is proposed to explain the observed G-value for the reduction of nitrate to nitrite. KEY WORDS : Radiolysis of aluminium nitrate. I. INTRODUCTION : The radiolysis of aqueous nitrate solution has been extensively studied in the past . The various factors which control the radiolytic yield of nitrite are concentration of nitrate, pH, temperature, dose, presence of dissolved oxygen. The present study deals with the yield of nitr-te formed as a function of concentration of aluminium nitrate. IT. EXPERIMENTAL : All chemicals used are of AnalaR grade. Aluminium nitrate solutions of various concentrations are prepared in triply distilled water. The-samples are prepared by taking 10 ml aqueous aluniniun nitrate solution in a stoppered glass tubes. Those samples tubes are irradiated f> gamma radiation using a Co gamma source; whose dose rate is 2.8 kGy hr as determined by Fricke dosimetry. The-radiolytic yield of nitrite is determined :3pect rophotomet r ical 1 y by modified sh inn's method . III. RESULTS AND DISCUSSION : A linear relationship is observed between the yield of nitrite and dose absorbed by aqueous aluminium nitrate solution. The G (NO) ) values are evaluafed bv varying the concentration of aluminium nitrate (Table 1). The yield of nitrite increases with increasing concentration of aluminium nitrate. Radiolysis of water produces e , OH, II and UOO^ as the primary radiolytic products. The reduction of nitrate to nitrite occurs due to capture of e and H by NO-. The OH radica] acts as a scavenger for NO9 in aqueous solution. The following mechanism is suggested for the reduction of nitrate. NO~ + e" *. NO, .- (l) •i 3 CJ .. (2) -- (3) .. (4) •• (5) .- (6) Two nolecuien of NO former) by reaction with H and e~ rHsprnport ionatos to proHuce one NO,, ion (reaction-sff whil^ OH radical renoves the nitrite thus for'rnon. = 1/2 (Gp- + nH - nOR) .. (7) Using the reported values of G Table 1 : The yield of G(NO_) at various concentrations of a ] tin i n i un nitrate 1 Concentration of A)NO, . 9H?-0 itiH lit" G 0. 1 0-J 0 0. 5 0.J 6 1.0 0.22 5-0 0.26 10.0 0.3 0 50. 0 0. 39 1 00. 0 0.39 IV- REFERENCES 1. fl.A. Bakh, V-I- fledvedovsky, A.A. Revjna, R.D- Bitaikov, Proclst All union Oonf. on Rad.Chem., Moscow, 1957/45. 2- M. Danieln, E,E. Wigg, J-Phys.Chem., 73/1969/17 03. 1. M.A. Proskurnin, Y.M- Kototyrkjn, Proc. 2nd Int.Conf. Peaceful fines of Atomic Energy, Geneva, 29/1958/52. 4- M.r,. Hyder, .1 . Phys-Cliem . , 69/1065/1858. 5- M. Daniels, E.E. Wigg, .7 . Phys . Chen . , 7 17 1 967/1 024 . 6. M.B. Shinn, Ind.Eng.Chem,Ana].Ed., 1941/1333- RC - 26.2 GAMMA RAD10LYSXS OF BINARY MIXTURES : NITRATK-ISO-BUTANOL AT 12 pH S.F.Patil and R.M.Pattl Department of Chemistry, University of Poona, Pune-411 007. (1NOJA) SUMMARY The products nitrite,aldehyde and hydrogen peroxide formed in the V radiolysis of binary agueous solutions of nitrate-iso-butanol at 12 pH have been estimated in aerated and de-oxygenated solutions. The G-values of the products are found to vary with the concentration of the reactantB. The mechanism consistent with the observed results is proposed. (Key words : Radiolysis of aqueous binary solutions, pH effect) 1. INTRODUCTION : Several reports are available en the radiolyeit, of neutral solutions of different organic and inorganic solutes However, limited data exits on the radiolyeis of binary solutions containing nitrate and alcohol particularly in alkaline medium. Present work includes a systematic study of radiolysis of solutions containing nitrate and iso-butanol at 12 pH in aerated and de-oxygenated solutions. Further, the effect of concentrations of each species on the product yield is examined. II. !£XPKUJMENTAL -^The aerated and de-oxygenated solutions were irradiated with Co-gamma source(dose rate 1.7KGy h ). After irradiation, solutions were neutralised with HC1 and then analysed for the products formed. Nitrite was estimated, by Shinn's method' aldehyde by Johnson and Scholes methoa ' and H2O2 was by Ghormley's method0'. III. RKSUF.TS AND DTHOUSSTON : The. formation of NO2, (CH3)2CHCHO an i H2O2 is found to be linear with absorbed dose. G-valufta of product., computed from yield-dose curves are presented in Table-1. The G values of nitrite are found to be higher in de-oxygenated solution^ than that observed in the aerated solutions. No definite trend wa-i observed for thj aldehyde formation. Though W-^p'i wae formed in th- aerated solutione it was not detected in the de-oxygenate.i solutions. The following mechanism ie proposed to explain fcht: observed results. The nitrate ions are known to scavenge preferentially egq giving nitrite as follows, 2 N03^ e^g > NO3 (1) NOg + «H30 )H > HNO3 + H9O (2) HNO3 > NO9 + OH (3) 2NO2 + H2O > N02 +"N03 + 2H (4) The oxygen also acts an a scavenger for e'dg as, eig + °2 > °2 _ <5> 202(+2H2O) > H2O2 + O2 + 20H _ <6) In basic medium OH radicals get ionized into O'which in turn reacts with nitrate leading to tha formation of NOjj. r MO3 + 0 + H20 > MO3 + 2OH (7» 2NO3 > 2NO2 + O2 (8) RC - 27.1 the MO2 formed gives NO^ via reaction(4). fornted in the radiolysis in basic m.jdium reacts with OH H2O2 + OH ' > H20 + H02 (9) OH and H radicals initiate H-abstraction reaction with alcohol. OH > (CH3)2CHCHOH + H20 (10) H > (CH3)2CHCHOH + H2 (11) the radicals thuf formed either react with NO3 + N0§" > Table-1 Dependence of ( (NOo), G(ittobutyraldehyde) and G(H202> en nitrate and iso-butanol concentrations. Aerated De-oxygenated rN03] tAlcohol! T. ) G(-CHO) G(H2O2) G(N02) G(-CHO) 1 1 1. 33 0.60 2.23 1.27 1 100 1.00 2.73 0.65 2.41 2.21 100 1 1.48 1.20 0.67 3.18 1.20 100 100 1.67 1.73 0.75 3.56 1.68 L.M.Daniels, E.Eric, and E.Wlgg J . Phys . Chern. ,71( 4L, 1024 (1967). 2. M. B. Hyder, J - Phys. Chem. , 6SI.6J , 1858 ( 1965) . 3. J.T.A.! Ian, J . Phys. Chem. , 66(9.)., 2697 (1964). 4.M.B.?hinn, I nd . EngfZ. Chem. (Anal .Ed. ), 13, 33 (;941). 5.G.R.Johnson and G.Scholes, Analyst, 72, 217 (1954). 6.A.O.Allen, C.J.Hcchanads1, J.A.Ghormley, and T.W.Pevis. J.Phvs.CheM. 56", 575 (J952). 7.H.A.Mahlman, J . Ghem. Phys. . 35, 9-36 (1961). RC - 27.2 EFFECT OF HETEROPHASE AiDDITJVKS ON THE GAMMA HAJHULYSJS OF SOME NITRATES S-F.Pat.il and S.S.Pawar Department of Chemistry, University of Poona, Pune-41) 007.(INDIA! SUMMARY: hrudies of radiation decomposition of eutectic of j KNC^and Cu(NO^. 3H2O in presence of oxides or metal powders with ti o absorbed dose reveal that the oxides or metaJ powderf; acceleratr- the rate of radiolvsis. These results are explained on the basis o. energy transfer processes occurring at. the surfaces c-f t.h«« constituents and aJso in terms of electron donor - System ti(N0 Pure eut • ectic of NaN03-KNO- 3 0. 58 >10 mo 3 e% Co304 0. 63 t 10 mo 3 e% As2O3 0. 66 i 10 mo 3 e% Mp.O 0. 69 i 10 mo! e% Bi2<->3 0. 76 i 10 trio 3 e% T1 2 03 0. 88 f 10 mole:', PbO 0. 61 -f 10 mo 1 e.% PbO 2 0. 63 i 1 (;i mole% p|)3O4 0. 70 » 10 mole.% Zn-powder 0. 66 i 10 mo) e% Cd-powder 0. 70 1'ure Cu( NO 3 >2 . 31 i2 <"' 0. 15 10 mo J e% i JO mole.% Zn powder 0. 36 l 10 111'"'] €:% (.Id- powder 0. 19 RC - 28.2 ENHANCEMENT IN GAMMA RAY IMDUCED DECOMPOSITION OF BAPIUM AMD STRONTIUM NITRATES BY SULPHATE AND CORBONATE ADDITIVES Miss N. G. JOSH1 and A. H. GARG Department of Chemistryr Nagpur Uni varsity, NAGPbf"4.4OOlO. SUMMARY - Decomposition of barium and stronti •-!• nitrates by ^-irradiation is enhanced ay sulphate and corbonate addttives. Thermo— luminescence studies suggest the formation of radical species SO«,SOs, Oa etc. which may interact with the transient species of nitrate causing energy transfer. ESR studies also support this suggestion. A comparison of the two additives shows that corbonate matrix induces more decomposition compared with sulphate matrix. Decomposition also varies with the absorbed dose. CKeywords-Radiolytic decomposition.ESR.Thermoluminescence.GCNO2}, Effect of additives. 3 I INTRODUCTION- Earlier we hav* reported energy transfer in thorium nitrate by sulphate additive's wi^h different cations Nix. SVudies on \,ne alkali and alkaline earth metal nitrates and various additives have been report-ad from this laboratory \2,3x. Present studies were undertaken to get further evidence on radiolytic and energy transfer n»echanism\4\ in y-rariiolysis of these nitrates at various compositions ar>d absorbed doses. II.EXPERIMENTAL The binary mixtures were prepared by grinding together AR grade salts C50 mesh!) in an agate mortar and irradiated in » Co-60 Gamma Chamber-9OO at a dose rate of O. 8O k<3y h * Experimental details of [NO2] determination and further calculations of GCNO2? are same as described earlier \3\. The ESR spectra of the ^-irradiated samples were recorded through the kind courtesy of Dr.M. D. Sastry of BARC Bombay. III. RESULTS AND 01SCUSSION y-radiolytic decomposition of barium and strontium nitrates in binary mixtures with th^ir respective sulphates and carbonates was. studied in the composition range O. S-1OO mol% of nitrate upto 'v 25O kGy. [NO2J and GCNO23 for various binary mixtures of barium nitrate with sulphate and carbonate matrices at 5O kGy are given in Table 1. It is pbserved that GCNOO values decrease with_the increasing amount of nitrate in the binary mixtures. Further ,GCNO23> deer eases significantly upto 5 mol% of nitrate iji binary mixtures after vhich decreasing trend slows down. Higher GCNO23 values in th» lower composition range C <5 mol '/i of nitrate} could possibly be attributed to the interaction of radical species and defect centres formed in sulphate matrix with that of nitrate. TL-glow curves_ at various doses CFig.15 have shown the presence of Oa *nd SO* species in y-irradiated binary mixture* of BaCNO9?2 + BaSO*. At 2SO JcGy TL peak for SOi disappears and intensity of Oi is reduced.lt explains low GCMO23 at higher dose* for all th« compositions. Th^se radical species interact with NO» thus enhancing the decomposition- A similar study of the binary mixtures of these nitrates with respective carbonate additives shows enhanced decomposition at all RC - 29.1 compositions and doses. Carbonate additive seems to be more efficiant medium for energy transfer compared to sulphate additive CTable ID. ESR studies have shown the presence of NOa and CO2 species at an absorbed dose of SO kGy and TL studies support this observation. A possible mechanism for the interaction of various species has been suggested. I V.REFERENCES: l.N.G. Jashi and A. N.Garg.Ccommunicated3. 2. S. P. Kulkarni and A. N. Garg.Radiat. Phys. Chem. 32, 609 C19883. 3. S. P. Kulkarni and A. N. Garg. Radi at. Effects & Defects in solids, 11.3, 315 Cl9903. 4.M. Khare and E.R.Johnson, J. Phys. Chem. 74_, 4O85 C19705. Table. 1. t NOz 3 and GCUO2O values of binary mixtures of barium nitrate with barium sulphate and barium carbonate at SO kGy. Mol% of BaSO* BaCOs BaCNO3~ •» GCNO2D [NOa] GCNO2D 0. 5 336 61.57 6931 1 .. O 352 227 3O71 1727 2. O 34O 11C 1O1 28. 4 5. O 317 4O 669 74. 4 10. O 1O2 6. 7 418 23. O 20. 0 313 9. 4 S39 14. 4 4O. O 354 4. 9 426 5. 4 6O. O 34O a. 8 424 3. 4 8O. O 374 2.1. 4OO 2. 2 100. o 365 1.5 409 1.7 f ^.50 K °3 i 1 1 1 1 1 \ I 1 s>r ! IfVi DOSE= '70 UGy DOSE - 2S0UGy a: DOSE =100 kGy / I V °3 7\ Z<50K J \ y h .'/ , t 300 500 A00 500 400 soo . (K) -•• HG.5 TL GLOW CURVES FOR BaS04l---) AND BaSO/. + lMOL'/. Bo(N03)2 RC - 29.2 LYOLOMTNESCENCE OF LUMTNOL INDUCED BY y-IRRADIATED INORGANIC PHOSPHORS CD. Kalkar, V.M. Raut, V.J. Pat H and Neeta Lala Department of Chemistry, University of Poona, Pune 411 007. SUMMARY : Emission spectra of lyol uminescence induced by •y-f riradiated inorganic phosphors in aqueous luminol solutions are recorded on a Fuess spectrograpV). Mqht emission occurs due to interactions of paramagnetic centres with Tuminol producing excitation of luminol molecules during released of trapped centres in aqueous solution. KEY WORDS : Lyoluminescer.ee, luminol . T» INTRODUCTION : Inorganic phosphors when exposed to V-radiafcion store energy in the form of colour or paramagnetic centres. This energy is released during dissolution of the phosphor in water in the form of light known as lyoluminepcence (LL). The LL glow is of short duration and very weak in intensity. However, the LL intensity enhances in the presence of an activator. *5-Amino-2,3- -dihydro-1,4-phthalhydrazinedione is a chemi1uminescent substance known as luminol. It is interesting to study the LL emission of luminol in aqueous alkaline solutions induced by y-irradiated phosphors such as NaCl, Na_CO_. and Na_SO.. II. EXPERIMENTAL : The optimum concentration of luminol in aqueous calcium hydroxide, borax, ammonia and ethyl amine is 3.5 x 10 M. The LL emission spectra are recorded by maintaining the LL glow in front of the slit of a Fuess spectrograph during exposure by continuous addition of irradiated salt in aqueous luminol solution. A Kodak film of 400 ASA is used for recording the LL spectrum and low pressure mercury lines are used for the calibration purpose. III. RESULTS AND DISCUSSION : The Intensity distribution of LL spectrum is obtained with the help of a microdensitometer. The LL intensity is replotted on a linear wavelength scale. The resolution and fitting of LL emission is done with the help of a microcomputer . The emission range, band maxima and half-band width for LL emission spectra are summarized in Table 1. A shift in LL emission peak to the longer wavelength is observed with the nature of the irradiated phosphor. Alkali halides produce electron :*nd hole centres while sodium carbonate and sulphate contain various paramagnetic centres on exposure to V-radiation. These centres interact with luminol producing excited states which emit light on returning to the ground states. RC - 30.1 Table 1 Analysis of LT. emission spectra of luminol Phosphor System Emi ss Ion Emission Half-band range (nm) peak (nm) width (nn) NaCS L + Ca (OH) 390-530 437 1 05 L + Borax 380-520 438 90 L + Ammonia 395-590 440 75 L + Ethyl amine 391-508 436 75 Na CO, L + Ca (OH) 400-495 442 61 L + Borax 400-495 443 62 L + Ammonia 4 00-497 442 61 L +• Ethyl amine 400-495 442 60 Na SO + Ca(OH) 3B5-530 449 ] 03 + Borax 385 - 525 446 1 00 + Ammonia 390-530 448 98 I, + Ethyl amine 395-500 445 95 IV. REFERF.NCRS 1. E. Wiedemann and G.C. Schmidt; Ann.Phys., 56, 210 (1895), 2. CD. Kalkar, H.J. Arnikar, S.v. Doshi and R.S. varkhede; Int.J.Appl.Radiat., Isot., 36, 51 (1985). 3. CD. Kalkar; Radlochem.Radioanal .Lett. , 58, 317 (1983). 4. CD. Kalkar and Neeta Lala; Appl .Radiat. Tsot. , 41, 635 (19«>0) 5. R. Ottmerj Z.Physik., 46, 798 (1928).. 6. N. Hariharan and J, Sobhanandrij Tndian j.Pure and Phys. , 8, 252 (1970) . RC - 30.2 Gamma Irradiation of Bi-Suptsrconductors Amitava Der N.R.Das and S .N.B'iattacharyya Nuclear Chemistry Division Saha Institute of Nuclear Physics 1/AF, Bidhannagar, Calcutta - 700064 SUMMARY Preliminary studies on the effect of /-irradiation on high T superconductors, Bi9Sr0Ca.Cu,0 and its Pb-substituted analogues, indicated , in general, a decrease in T values for both the varieties. However, the effect of degradation in the superconductivity is more pronounced in Pb-dopped specimens. Thur, the presence of lead, inspite of its stabilizing i INTRODUCTION Studies on the effect of radiation on high T^ superconductors which may have some possible applications in different radiation enviorments such as in outer space, are of crucial importance. Several workers investigated the effects of different radiations on the characteristic properties of YBa-Cu.O superconductors and it was observed that there is, in general, a negative effect on the superconducting behaviour of YBa.Cu,0 . However, irradiation study on Bi-superconductors, although it has got comparatively higher T s, is very much lacking. The present paper deals with a preliminary study on the effect of gamma radiation on Bi-superconductors as well as on its Pb-substituted variety with a Co source. EXPERIMENTAL Superconductor samples with nominal compositions of Bi-Sr-CajCu-O and Bi- Pb Sr,CaoCu,0 were prepared with appropriate amounts of Bi,O,, PbO_, RC - 31 .1 samples, the mixture was first fired for 24 hours at 840 C, the calcined mixture was then powdered, pressed into pellets and again sintered at 860 C for about 70 hours. It was then cooled slowly to room temperature (1 C/min.). The resistivity measurements were carried out by standard four probe method. Electrical contacts were made with silver paints. The pellets were irradiated with Co-60 y-source at different doses. X-ray powder diffraction (Cu K ,) studies of the samples were made with Philips automated Diffractometer„ DISCUSSION Studies on the resistivity of the synthesised superconductors against temperature revealed that the samples with nominal composition of Bi,Sr,.Ca.Cu,0 contained only the low T (221?) phase having the T (R = 0) £t £t it J y C C arround 78K whereas the Pb-substituted samples with nominal composition of Bi.. gPbg -Sr.Ca-Cu^O was found to contain a high T (2223) phase as indicated by the drop in resistance measured at 94K. On irradiation of the samples under identical condition there was a decrease in the T values for both the varieties. But the effect of radiation was more pronounced in the case of Pb-doped samples. For the pure variety, after irradiation there is a fall of T value by about 4K from 78K to 74K, whereas in the substituted variety, T value decreses by about 10K from 94K to 84K. X-ray studies also corroborate our experimental findings regarding the presence of different phases in both non-irradiated and irradiated samples. The pronounced degrading effect of radiation on the T value of the Pb-dopped superconductors may be attributed to the presence of lead which leads to disordering the weak Cu-0 bond resulting in the decrease in the fraction of the high T phase. However it is yet to be ascertained. REFERENCES 1. J.Bohandy, J.Suter, B.F.Kim, K.Hoorjani and F.J.Adrian, Appl. Phys. Lett., 51, (1987), 2161 2. J.R.Cost, J.O.Wills, J.D.Thompson and D.E.Peterson, Phys. Rev.B 37, (1987), 1563 3. K. Shiraishi, H.Ito and O.Yoda, J. Appl. Phys., 2J7, (1988), L2339 RC - 31.2 THE EFFECT OF GAMMA RADIATION ON THE PHYSICO-CHEMICAL AND CATALYTIC PROPERTIES OF LaQ,6Sr0.4C0O3 Prince C.Koran.V.R.S Rao. V.R&makriahnan and J.C.Kuriacos* Department of Chemistry, Indian Institute of Technology, Madras-600 036. India. SUMMARY The effect of Co gamma radiation on the catalytic activity of lan $SCQ 4C0O3 for the decomposition of hydrogen peroxide has been studied using various physico-chemical techniques. Irradiation of the catalyst with gannt rays alters the bulk concentrations of Co , Co and Co + irons, the average oxidation number of Co and oxygen non-stoichiometry of the sample. The electrical conductivity of the sample decreases and the intensity of ESR signal increases with increasing radiation doee. An explanation of the catalytic activity of irradiated catalysts is proposed. [ Key words : Gamma ray irradiation, Hydrogen peroxide. Catalytic decomposition, Redox system, Radiation dose. ] I. INTRODUCTION PeroV8kite type oxides of the type LnM03 and Ln1_xBxM03 [Ln= rare earth metal, M= transition metal, B= Ca.Ba, and Sr] have been proved to be potential oxidation-roduction catalysts tor many reactions /X/. The effect of ionising radiation on the catalytic activity of certain perovsklte type oxides for the the decomposition of hydrogen peroxide has been reported /2,3/. In the present investigation, the effect of Co gamma radiation on the catalytic activity of LaQ ^SrQ 4C0O3, a typical strontium substituted lanthanum cobalt oxide of perovskite structure, for the decomposition of hydrogen peroxide has been studied. II. EXPERIMENTAL *-a0.6Sr0 4C0O3 nas been prepared by ceramic technique and characterised by XRD using CuR,*, radiation. The catalytic activity of oxide samples has been determined using a gasometfic apparatus described elsewhere /3/. The samples have been irradiated in open petridlahea for differet periodf of time in a 5000 Cl gamma chamber at a dose rate of 0.15 Mrad /hr. The total cobalt content has been determined by EDTA tltration method. The average oxidation number of cobalt and the extent of oxygen non-stoichiometry exibited by unirradiated and irradiated samples have been determined by estimating the amount of Co and Co iodometrically, The samples were leached with water and the amount of Co in the solution estimated by atomic absorption spectroraetry. The determination of activation energy for the reaction, measurement of electrical conductivity ot the pelletised samples using four probe technique, identification of the chemisorbed oxygen species and the investigation of topography of the surface of the samples by means of SEM have been carried out. RC - 32.1 III. RESULTS AND DISCUSSIONS The XRD patterns of unirradiated and irradiated samples show only provskite phase. The rate of catalytic decotnpoai t ioti of hydrogen peroxide varied with variation of irradiation time of the samples. It shows a minimum with the catalyst irradiated for 54 Hrada and a maximum with the catalyst irradiated for 86.4 Hrads (Tabl«-I). The activation energy for the reaction increases on increasing the radiation dose and reaches a maximum with the catalyst irradiated for 54 Hrada. There seems to be an abrupt change in the rate and the energy of activation for the reaction with the catalysts, irradiated for 86.4 Mrads, indicating the possibility of two different surface processes for the catalytic decomposition of hydrogen peroxide. The results of chemical analyses show a decrease in the concentrations of Co , average oxidatipn number of cobalt and an increase of concentrations of Co and Co and oxygen ncn-stoichiotnetry in the oxide samples, on increasing the radiation dose (Table-II). The ESR spectra of irradiated samples show a signal (JCJ-2 . 0028 ) , charact er eat i c of the auperoxide ion, 0~^ . This value of g-tenaor is very close to that of free electron. The absence of hyperfine structure in the ESR signals suggests that the chemisorption of oxygen takasplace on Co , by transfer of one electron from Co to the O2. to forrs an adsorbed complex j There is a gradual increase in the intensity of ESR signals, with samples irradiated for longer time. The electrical conductivity of the irradiated samples decreases on increasing radiation dose. The SEM photographs of samples irradiated for 122.4 fti/ada, show surface cracks and surface corrosion which may be due to surface segregation of Co * ion and the formation of superoxide ion. The variation in the catalytic activity of irradiated samples can be explained on the basis of valence states of cobalt ion in the oxide samples. According to theory of " compensating reactions ", an oxide of the element which may form a redox system involving two different valence states can act as a catalyst for the decomposition of hydrogen peroxide /4/.The presence of various valence states of cobalt makes La^ ^STQ ^COO^ a catalyst for the decomposition of hydrogen peroxide .The catalyst subjected to a small radiation dose has still a relatively larger proportion of Co in the higher oxidation state while large radiation dose leads to a higher proportion of the lower oxidation states. The cobalt in the higher oxidation state can function aa an electron acceptor and that in the lower oxidation state aa an electron donor. The ability of hydrogen peroxide to function aa both oxidising and reducing agent makes it possible for both the higher and lower oxidation states of cobalt to be active for the decomposition reaction. However the rates of decomposition of hydrogen peroxide on the two species may be different. ACKNOULEDGEMENT One of the authors ( PCK ) expresses hia sincere thanks to CSIR, New Delhi for the financial assistance. R C - ~i~i .2 IV. REFERENCES 1. Y.F.Yu Yao, J. Catal. 36. 266 (1975). 2. S.Balasubramanian, V.R.S. Rao, B.Vi3wanathan and J.C.Kuriakoae, J. Radloanal. Nucl. Chem. Letters 96, 301 (1985). 3. B.Srinivaa, Ph.D. Theeia (1989), Indian Institute of Technology. Madras. «. G.H.Schwab. Z. Anorft. Alleg. Chew. 295. 36 (1958). TABLE-I Rate, Activation energy. Electrical conductivity and ESR signal Data Radiation Rate Activation Intensity of Electrical dose energy ESR signal conductivity Mrada raol/llt/min/s KJ/tnol a.u. mho/cm 0.0 2.181xlO~2 38.104 0.0 2208.19 25.2 1.682xlO~2 46.677 316.8 1878.95 54.0 8.985xlO~3 75.830 431.2 1860.59 86.4 3.623>rlO~2 45.679 1734.0 908.60 122.4 2.141xl0~2 52.450 3520.0 861.42 TABLE-11 Bulk composition, Average oxidation number of Co and Oxy(t«n non-atoichiometry (S) in LaQ $Sr0 ^Q.oQ^_£ Radiation Co2+ Co** Co4+ Average doae oxidation o fir ad a g/g of catalyst number 0.0 0.0 0.1838 0.0780 3.2984 0.1020 25.2 2.1xl0~5 0.2124 0.0495 3.1888 0.1056 54.0 8.808xl0~4 0.2210 0.0401 3.1501 0.1249 86.4 2.422xl0~3 0.2289 0.0306 3.1079 0.1463 122.4 8.593xlO~3 0.2316 0.0251 2.9749 0.2109 RC - 32.3 INFLUENCE OF PRE-ANNEALED DAMAGE FRAGMENTS IN THE DECOMPOSITION OF T-IRRADIATED CAESIUM BROMATE D. Bhatta, M.K. Sahoo and H3. S. Mishra Department of Chemistry, Utkal University, Bhubaneswar-75100-: SUMMARY The present study throws light nn the influence of pro- annealed radiolytic damage entities on ihe isotheimai decornpoti t ion of V-irradiated caesium brornate. It is indicated that damage fragments accelerate the process in the initial stage but retard the same in the later stage. Pre-annealed samples decompose at a lower rate than that of the untreated one. The process follows initial gas evolution, acceleratory and decay stages. Key words : Radio] ytic damage entities, isothermal decomposition, caesium bromate, Y-irradiation, preanneaied samples. I. INTRODUCTION The effect of irradiation and doping on the isothermal decomposition cf inorganic molecular ions has been well studied but the crystals which undergo decomposition in a molterybemi- molten state forming an eutectic has not received much attention. It is of interest therefore to undertake this present piece of work. II. EXPERIMENTAL Caesium bromate was exposed to a dose of J.O MGy of CO V-rays and the total oxidizing fragments as well as bromide were estimated . Recovery study was carried out at 433.0 K at different time intervals. Isothermal decomposition study on unirradiated, irradiated and pre-annealed samples was carried out at 673.OK gasometrically. The fractional decomposition, ot, was calculated from pressure values. RC - 33.1 III. RESULTS AND DISCUSSION The total oxidizing fragments generated in the crystals of caesium bromate upon irradiation is 43.70 J^eq/g which gradually disappears with increasing time of heating (Fig.l) suggesting that annealing of the species is taking place. These damage entities play an important role both in nucleation as well as in nucleus growth, occuring j, decomposition. The process follows (Fig.2) initial gas evolution, acceleratory and decay stages. Though irradiation facilitates the decomposition in the initial period, it retards the process subsequently (Fig.2). As the decomposition of caesium bromate is taking place in the molten/ semimolten state, in the acceleratory stage when radiation-induced lattice defects and trapped charges no longer exist, the observed effects are due only to the chemical damage fragments generated upon irradiation which constitute decomposition nuclei themselves and may be termed as irradiation nuclei. But as the decomposition proceeds, a part of the damage entities recombine to give bromate ion resulting lower decomposition. In case of pre-annealed crystals, major part of the oxidizing fragments , BrO , BrO., , O , O, and O? undergo recombination resulting BrO ion whereas the remainder decomposes to yield bromide. Thus the decompo- sition is controlled by two simultaneous opposing effects, i.e., acceleration by the catalytic effect of damaged bromide, Br and retardation due to back reaction to generate bromate ion. In the pre-annealed samples, the influence of bromide ion is overpowered by the annealing effect causing lower decomposition. IV. REFERENCES 1. D. Bhatta and M.K. Sahoo, Radiat. Effect. Defect. Solids (1990) (In press). 2. D. Bhatta, S.R. Mohanty and K.C. Samantaray, Radiochim. Acta, 2^, 13 (1981). 3. N. Bohidar and S.R. Mohanty, Radiochim. Acta 27, 19(1980). RC - 33.2 -i-' "a '2 C o i ' o" o Of o 14 o cf o in o O o LTI OJ O O o o o d o* CD o CO o 2 £ CD O LU 6/nbaiV'xo-ja UON RC - 33. RECOVER/ OF RAOIOr.rTTC UrMAd ENTITIES AtfO ROLE PT.AYED BY DOPANT IN Y-LRU -iDf ATED SODIUM BWOMATE D. Bhatta and K.K. Sahu Department of Chemistry, Utkil University, Rhnb;.neswar-7510O4 SUMMARY Recovery of radiolytic damare entities in /-irradiated sodium broraate and the effect of Ra doping on the prrs<-ess has boon studied. The dam?-ge hypobromite, BrO and bromite, BrO_ generated by 0.5 MGy of Co Y-rays are respectively 7.(">0 and 10.10 JA mol/g. The data show that cation vacancy (doping) hns marginal effect on radiolysis as well as on recombination of radio - lytic damage fragments. The process follows diffusion controlled mechanisms. Key words : Rorovery , radio! ytic damage entities, }' iti doping, diffusion controlled mechanisfp-3. 1. INTRODUCTION Although radiolysis arid recovery of damage species in •V-irradiated lialates and perhalates has befii reported by many workers , effect of vacancy, cation/anion on the annealing process ).as not been well explored and the present study has been carried out. II. EXPERIMENTAL Crystals of sodium bromate were doped with BJ" ions (1.0 mol%) by coprecipitation method. Pure as well as doped materials were exposed to 0.5 ta'Cv of CO Y-rays and the hypo- ' 2 bromite and bromite were estimated'. Recovery study was carried out between 413.0 - 443.0 K us-ing a thermostatically controlled air bath. RC - 34.1 III. RESULTS AND DISCUSSION The major radiolytie products, BrO and BrO., (p* mol/g) induced in the irradiated crystals are respectively 7.60 and 10.10 (pure), 7.75 and 10.25 (doped) indicating that cation vacancy has marginal effect on the initial damage which may arise due to smaller size of the host material. i. ANALYSES OF THE RECOVERY ISOTHERMS The recovery isotherms (Figs- 1&2) show an initial fast reaction fallowed by slow change. Hypcbromite, BrO being an unstable species anneals to a greater extent than that of bromite, BrO } though the radiolysis is in the reverse order, the fraction. IV. REFERENCES 1. D. Bhatta, S.R. Mohanty and K.C. Samantaray, Radiochim. Acta 2ft_, \i (1961) and references therein. 2. T. Andersen and H.E.L. Madsen, Anal. Chem. 3J7, 49 (1965). 3. D. Bhatta and Late S.R. Mohanty, Radiocbim. Acta 49_, 57 (1990). RC - 34.2 0.8 f —I (S0.6 T / S < U xT z (I // Ct < 0.4 z w SODIUM bROMATF o tUlf (B1-O2 ),0.SMGy t- o O 413.0 K < r A 423.0 0-2. IX i O 433.C v 423.OK (8a2 + + Doped, ) 0 5 10 15 20 TIME OF HEATING,h Fis-1 Annealing of bromfie in r- irradiated Sodium SODIUM MRC.VATE CJrO") L.5 MGy o 413.UK A 423.0 • 433.0 o 443.0 2+ 0 V 423. CK(Ba Doped ~ 5 10 IS 20 TIME OF HEATING h . 16.2 Annealing of liy^jo'bromife in r-irradioted RC - 3A.3 SOLUBILITY OF KUTCH LIGNITE IN DIFFERENT SOLVENTS MODIFIED BY Y -IR R A DIA T1O N I..C. Ram, P.S.M. Tripathi, S.K. J'ia, G.S. Murty and S.P. Mishra* Central Fuel Research Institute 1>.'J. FRI - 828108, DHANBAD IBiha;, SUMMARY Solubility of Kutch lignite in different solvents (ben^em-, elhdm.,1, 1.-1 ethanol- benzene mixture and pynLdine) by Y^-irradiatiori at vai yimj doses (50, i.;0 ana i 5 0 Mrad) in different media (aerial, aqueous, and C el ) ;s inoriifiea by th» :n . jocular 4 rearrangements due tc aither degradation or cross-jinking <-A poly-s remade groups in lignite during irradiation. (Key words : lignite, solubility, y-irradiation, roiyaioma- tic groups, degradation/cross-linking) £, INTRODUCTION : As a sequel to our previous communication , the i esuJts of extended investigations on the extra stability of Kutch iigrd'e in different sol- vents modified by Y-irradiation in different media are reported in this paper. IL EXPERIMENTAL : The Co-60 Gamma Irradiator unit is described elsewhere . The experimental procedure for solubility studies was exactly the same as men- tioned in our earlier paper . Aqueous irradiated lignite prior to solutilisataon was filtered and washed with hot water till free from acid and dned at 1O5''C for 1 and 1/2 hours. The solubility was determined by weight loss on dry ash free basis. HI. RESULTS AND DISCUSSION : The results on the solubility of raw Kutch lignite in the above solvent visra-\as that of irradiated lignite in aerial, aqueous and C Cl media are shown in Figs. 1-3. The important findings are capsuled below: 1. It is observed (Fig. 1) that during aenal irradiation of lignite, both degrada- tion and polymerisation or cross-linking take place simultaneouslyJ, the former at all doses upto 150 Mrad (change in solubility in benzene from 10.72 to 5.14 %) and latter after 50 Mrad (solubility increasing from 4.56 to 10.87 % in ethanol). In case of azeotropic etwnol-benzene mixture, primarily the solubility of both raw and irradiated lignite in the mixture solvent seems to be much more effective as compared to individual solvents (e.g. solubility on aerial irradiation 14.95 (raw). 16.13 (50 Mrad), 18.19 (100 Mrad) and 12.34% (150 Mrad) instead of 10.72 and 10.70 % of raw; 9.98 and 4.86% of irradiated upto 50 Mrad; 8.3 and 9.38% of irradiated upto 100 Mrad; 5.14 and 10.87% of irradiated upto 150 Mrad in benzene and ethanol solvents respectively). The same observation holds true for other media of irradiation (Fig. 2 and 3). This might be due to the selecti- vity of the solvent mixture. Further, the solubility in pyridine solvent does not seem to be atfected considerably, thereby showing that pyridine- soluble polyai omatic groups are less prone to be affected during aerial irradiation. * Department of Chemistry, Banara.'i Hindu University, VARANASI 221005 (II.P.) RC - 35.1 In case of aqueous irradiation (Fig. 2), the solubility decreases almost at all ooses, however, it is much more pronounced at 150 Mrad (i.e. f ro m 10.72 to 2.69* in benzene; and 10.7 to 3.3% in ethanol; 11.26 to <.12% in ethanol-benzene mixture; and 20.35 to 16.57% in pyridine). This is pro- bably due to the higher reactivity of lignite to undergo higher polymerisa- tion in aqueous than aerial irradiation. In C Cl medium of irradiation (Fig. 3), however, the extra eta bility after radiation' chlorination increases upto 100 Mrad (from 10.72 to 18.66% in benzene; 15.84 to 22.4% in ethanol; 18.11 to 23.84% in ethanol-benzene mixture, and 22.76 to 24.72% in pyridine), after which it either levels off or decreases. In this case , the increased solubility is ascribable to the breajt-down of cohesive secondary forces rather than rupture of primary linkages , whilst the constancy or decrease might weJL be due to polymerisa- tion or cross-linking. The radiation chlorination is caused by chlorine evol- ved during radiolysis of C Cl , the overall radiolytic reaction being : 2 C ci2 However, it is not yet clear whether the chlonnatton takes place in the aromatic or aliphatic species of lignite. Further work in this direction is in progress. REFERENCES 1. P.S. M. Tripathi, L.C. Ram, and S. K. Jha in "Advances in Coal Chemistry", N.P. Vasilakos (Ed.), Theophrastus Publications, Athens, Greece, pp. 333-349 (1988). 2. L.C. Ram, S.K. Jha and P.S.M. Tripathi, Preprint - Volume, Symposium on Radiochemistry and Radiation Chemistry, Paper No. RE-05, Nagpur, Febr- uary, 1990. 3. E.J. Heneley and L. Korasyk, Fuel, 40, pp. 155-159 (1961). 4. A.J. Swallow, "Radiation Chemistry of Organic Compounds", Vol. H, Per- gamon Press, New York, p. 95, (1960). 5. H.H. Lowry, "Chemistry of Coal Utilization", Und SuppL Vol., Martina Elliot (Ed.), Wiley-Interscience, USA, p. 469, (1981). -oOo- SOLOBI1ITY OF KUTCH LIG NITE IN DIFFERENT SOLVENTS MODIFIED BY y-IRRADIATION 1 —— t_-— \ i» Si P-«^—-—«_ —cr- \ o -o tnim o—o TH*M* 0—O *—* IN -«•!(* •—• fr*"- -ftAT DOM .Nil - HAT D05( iHrMI . Fig. 1 - Aerial Inadiation Fig. 2 Aqueous Irradiation Fig. 5 - Irradiated in C Cl RC - 35.2 V-IRRADIATED ALKALI HALTDES AND FORMATION OP COMPLEXES OF TRAPPED Br2 AND Cl WITH ANILINE L. Bapat, G.N. Natu and G.Y. Rohokale Department of Chemistry, N. Wadia College, Pune - 411 001. SUMMARY : X-Trradiated KC1 and KBr react with aqua-aniline emnlsion and produce~CT bands of aniline-Br_ and aniline-Cl_ complexes in aniline layer. KEY WORDS : Extraction of trapped halogen in KC1* and KBr* , I. INTRODUCTION : In the earlier work1'2 CT bands of complexes of iodine with aromatic hydrocarbons were observed in the organic layer in which trapped iodine in KI* (/-irradiated) was extracted. II. EXPERIMENTAL : 5 g of KC1 and KBr powders of 3 0 and 150 mesh were V* -irradiated (10 kGy) and dissolved in 10 ml of 1:1 aniline-H.O (v/v) emulsion. Electronic absorption spectra of aniline layer were recorded. III. RESULTS AND DISCUSSION j It is proposed that V'-radiolytic product, halogen species, trapped in KC1* and KBr* react with aqua-aniline emulsion producing halogen-aniline complexes. Figs. 1 and 2 show absorption spectra of (1) aniline-halogen (pure) complex, (2) 'treated" aniline with KC1* and KBr* of different mesh. CT bands of aniline-Cl_ and aniline-Br_ complexes are observed in the case of irradiated salts. £Ssorption at all Xs increases with increase in "/-dose and amount of crystals. The enhancement in the yield of halogen in the case of crystals of 150 mesh may be due to the larger surface area in contact of fresh emulsion, IV. REFERENCES : (1) L. Bapat and D. Ravishankarj Radiochem.Radioanal.Lett., 50, 151 (1981) . (2) L. Bapat, G.N. Natu anmd G.Y. Rohokale? Radiochem. Radiation Chem.Sym. held at Nagpur (Feb. 19<*0). RC - 36.1 I 6f— 1-2 Olh 200 1.00 300 SCO >i/nr / nm Fig J. -Absoroticr ::c<:;ra cf aniline treated with -Absorption spectra of the 'treated' emulsions T-irradioU" KBr ot ii) 30 mesh and of water and aniline with (1) 7-irradiated KC1 (2! !b0 mesh (2) pure chlorine gas THERMAL ANNEALING OF GAMMA IRRADIATED AMMONIUM CHLORIDE CD. Kalkar, D.ftavishankar and Neeta Lala Department of Chemistry, University of Poona, Pune ^11007 * Novrosjee Wadia College, Pine 411001 SUMMARY: Amnoniura chloride prnduces N?H? and Cl0 as the main radiolytic products on gamma irradiation. Thermal annealing has a marked effect on the stability of N^ff* and Cl?. During the thermal annealing the chemical yield of nitrite and iodine was studied hy dissolving irradiated annonium chloride in aqueous sodium nitrate and potassium iodide respectively. The yield of iodine in isochronal annealing showed an exponential behaviour with temperature while that of nitrite showed a decrease and then increases at higher temperatures. The results are explained on the basis of dissociation and reconbination of N«H. with temperature. KEY WORDS: Chemical effects, Isothermal annealing of annoniun chloride I. INTRODUCTION: There aro very few reports on' thermal annealing effect on gp.nna irradiated annoniun halides. Isochronal annealinn of 2 ammonium chloride has been reported by Ravishankar and Chabria . In the present investigation the changes occuring during the isothernal /isochronal annealing of gamna irradiated NH/C1 was studied by observin|; the chemical effects produced in aqueous nitrit_« ami iodidr- solutions. II. EXPFRIMENTAL: One ?,m ammonium chloride powder (200 mesh) was irradiated to a predetermined gar-inn dose usinf; 2.5kCi Co gamma source. These samples vere subjected to thermal annealing at a desired tenperature followed by dissolution in 10nl of 0.1M )fa.MO and 0.5M KI solutions. The yield of nitrite was determined by modified Shlnn's method , The iodino liberated in KI was spectrophotometrically determined at 354nm. III. RESULTS AND DISCUSSION: The isothernal annealing curves in tarns of NO" and I_ yields show an exponential decrease with time (Flg.l). In isochronal annealing curve the nitrite formation shows a minimum at 90 C while that of iodino formation decreases sharply with increasing RC - 37.1 temperature (Fig.,2). The mechanise of reduction of nitrate to nitrite i« based on interaction of N0H+ with nitrate ions while iodide is oxidised by the N?U* and Cl2. The yield of nitrite decreases with temperature upto 90b. IXtrinp, this temperature ranrje the N H+ dissociates to A find 15 species as reported by ^arquardt1. Above 90 these 3pocies recombine to form N 'I* and hence the yield of nitrite increases rat higher temperp.tures. An regard.':- t'ie yield of jodine there is a continnou.s decrease which is probably due r.o th.- diffusion of gaseous chlorine, predominant over the reformation of N7H^ ?it higher temperature.'?, TV. REFEPFA'CF.S: 1. C.L.Marquardt, J.Cht-m. Phrs, V},(1970) T.'^tft, ir"~>7. 2. D.Pavishankar, N. Chahria, J. Padioanal. Uuc'l . Chen. Lett 134 (4), (1080) 26.-S. ' 3. M.B.Shinn, Ind. En?>. Chen. Anal.Ed., 13,(1941) 33. 0-20, 20 40 60 0 20 "40 60 Time of Isothermal annealingOmn ) Fig. I - Isothermal annealing of irradiated NH^CI • 80 120 40 80 • Temperature C Fig. 2-Isochronal annealing of /-irradiated NH^Cl RC - 37.2 KINETICS OF ISOTHERMAL Aim BALING OF HYWCBLURITX IN Y-IRRADIATED LANTHANUM CHLORATE HEXABYDRAIE SYSTEM R.S. lokhande, Miss S.8. Kelksr and Mies M.D. Bodas Department of Chemistry, University of Bombay Vidyanagari, Bombay - 400 098. SUMMARY Kinetics of isothermal annealing of hypochlorite formed in Y-irradiated lanthanum chlorate hezahydrate have been studied at different temperatures in the range of 85-155 C. The hypochlorite is found to anneal by a combination of first and second order processes, the former being i'ast reaches completion within *ew hours- The activation energy obtained for first and second •* - process from the data is 9.2 sad 5.82 KJ mole""^ respective Ke^ Words : Gamma Iri It is seen that the fraction annealed RC - 38.1 the particular species per gran of the sample and X. the amount of species changed by annealing at time t. Interaction of gamma rays leads to the excitation of chlorate ions followed by it's degradation to various products. The radiolysis of chlorate results in a variety of products such as Cl~, Clo~, C102~» CIO/", CIO, C1O2, ClO^, O2 and Oj. Among these the chlorine bearing fragments Cl~, CIO2*" are In higher yields. It is more likely that species containing both oxygen end chlorine will accupy lattice position and oxygen atoms or oxygen molecules will be trapped as interstitials either at or very close to the damage site of these way diffuse much further apart in the crystal matrix, depending upon the degree of excitation and the primary radiolytic event. These fragments do not undergo recombination reactions at room temperature due to an energy barrier. Only at high 1 emperattire of annealing they become mobile leading to different reactions. For the annealing of hypochlorite by the first order process, the reaction involving the decomposition of hypochlorite due to nearfcy oxygen is suggested, 010" + 0 J» Cl~ + o2 The fraction of hypochlorite which is distinct to anneal by the second order process is probably a random reaction controlled by the diffusion of oxygen CIO" + 02 —^ CI03" The activation energies obtained from the Arrhenious plots for the first and second oraer processes are found to be 9.2 and 5.82 KJ mole"*1. IV. REFERENCES 1. R.S. Lokhande and S. D. Daptardar; Proc. Symp. Kadiocherc. and Radiation Chem. Feb. (1990) N&gpur. 2. R.S. Lokhande and S.D. Daptardar; Proc. Symp. Radiochem. and Radiation Chem. Feb. (1989) Kalpakkam. 3. J.F. White., Am. Dye Stuff. Repr. , 31, 484 (1942). 4. I.M. Kolthoff, R. Belchev., Volumetric Analysis Vol, Interscience, M.Y., 267 (1957). RC - 38.2 Ei-'I r,CT5 OF GAI'D'.A-ITlRADIATIOr! Oi: ISOTHERMAL DECOi:POf ITICT-' TikTK AIJD ACTIVATION ENKR1Y 0? AMMOriUK PEROHLCKATE V.O. Dedgaonkar, University of Poo-ia, i'une 411 CC7, 2jdLr_Iiii lEJiLiJloiief J §.?_ L.'&5 La.. 2 PIIT. 2 2 _i _±.™£_ 411 _ Dpi and n. Dpvchandra Singh, Uovrosjee "ania College, Pune 411 001 Multistage complex isothermal decomposition of ammonium perchlorate (AP) is markedly influenced by gamma-irradiation. Dependence of acceleratory constants on gamvia-dosp do not bear a simple relation botveen the two when certain data are plotted. Logarithmic plots, yield straight line?. Activation energies Tor isothermal decomposition of AP decrease with rise in radiation dose- The strain and stress produced in voids of AP samples are responsible for lowering the activation energy. (Keywords* : gamma-irradiation., isothermal decomposition, rate constants, ammonium perchlorate, activation energy) J. iriThCi/.JCTiC;:: Themal decomposition or AP 3s influenced by gamma- -radiation dose /I,^/. Dependence or' acceleratory rate constant1- on rauiation dose is studied from certain known data', v.'hieh show lo(jar.1 thmj c relation between them. Activation energies computed by Arrhenlus formula reduced considerably on exposure of AP samples to ga.m:na-irradiation prior to its decomposition. II. EXPKMKr^'TAL In earlier experiments /3/ commercial grade pure AP samples (purity greater than &y >) of particle size 7.5 - 6.0 x 10~° min were exposed to gamma-dose at the rate of 5 kGy h~l from Co^'O source. These samples were subjected to isothermal decomposition. III. RRStJLTS A?1D DISCUSSIOr.' Acceleratory rate constants enhanced with rise in dose- The relation between the rate constants and gamma- -dose is not simple. The plots of rate constant (k^) versus dose are parabolic curves (see Fig. 1) and straight line plots for logarithm of rate constant versus dose are depicted by Fig. 2. The data have been analysed at various temperatures. Decay rate values are insignificant. The relation between acceleratory rate constant and dose may be proposed as log k, = CTD + Cg where C^ and C2 are constants and D is radiation aose. The activation energies for the same data computed using Arrhenlus treatment show a progressive decrease from RC - 39.1 74.4 to 27 kJ rriol-1 for a do?? range of 0 to 0-0 MGy (see Table 1). This is dup to the initiation of decomposi- tion of gamma-irradiation v;hich creutes /'!/ dislocation in AP samples. The strain and gtre-s produced in voids of AP samples provide energy for decomposition. B» Tides, dis- locations produced function as sources and sink-5 for the generation or trapping of vacancies und interstitial** favouring stereochemical environment for r.enerating decompo- sition nuclei. Concentration of decompo Ition nuclei increases with increasing dose. IV. REFERENCES 1. P.J. Herley and P.V/. Levy; J. Chem. Phy., 49, 14'JJ (1968) . 2. y.G. Dedgaonkar and D-U. Hajput, );-diochem. Kadioanal Letts., 57, 281 (1983). 3. D.U. Rajput, Ph.D. Thesis, University of Poona (1985). 4. N.B. Hannay, Treatise on Solid State Chemistry, Prentice Hall International Inc., London, Vol.4 (1976). Table 1 : Activation energies for the acceleratory stages of isothermal decomposition of AP for various doses. Dose/MGy 0 0.06 0.12 0.24 O.o 1 E* / kJ mol" 74.4 64.50 58.00 50.70 27.0 £• °)-y- 255 C 2 25 °C 3 « Z - * 0 B 200 t c •<2 OS 1-0 I 25 25 Dose/ MGy of loyarithm of rate constant versus dose. RC - 39.2 EFFKOT OF GA;J'A-IMIADIATIOI; O'-T ISCTlIEr.KAL DEHYDRATION A':D DECOKPCS IT10" RAT?;- C? MAGI'ESIUM PERCilLORATS D.U. Ra.jput, I'ovrrns.jee l.'adla College ,_.Pune_411 OC1, V .r> .Dedgaonkar Department of Chemistry, University of Poona, Pune 411 Oc.7, and R.T. Pincle, Iiov.rosjee -;adia Collrge, Pune 411 001 Acceleratory rate constants of isothermal dehydration and decomposition of magnesium perchlrrate are enhanced by exposure of samples of magnesium perchlcrate to ganca-irradiation prior to its dehydration and decomposition. Dependence of rate constants on gamma-dose is not direct. A logarithmic relation between the two is reported tv re at various temperatures. (Keywords : gamma-irradiation, dehydration, decomposition, rate constant, logarithmic). I. INTRODUCTION The effects of gamma-irradiation on isothermal dehydra- tion and decomposition of magnesium perchlorate have been reported /I/. Acceleratory rate constants for both the dehydra- tion and decomposition process of magnesium perchlcrate increase with rise in gamma-do?-" . There is a complex relation between the rate constants r.nc; iha r^mma-dose to which the samples are exposed prior to dehydration and dr-cornposition processes. II. EXPERIMENTAL Comnercial grud dry .luirnosjurn perchlorate samples of purity greater thun CO ,J wi-rc used in worV. reported earlier /I/ on thermal dehydration and decomposition of magnesium perchloraue. r'or expo:jui"i:, L. 'JCJ;:. ;U--TLXHL: Li .in source °^Co of ostinat^d dost.' rt-te 5 kOy h~l was err.plcyed. Dehydration of macnesiuni perchj.o- rate was effoctf d at .JCl> °C; a temperature below the value of decomposition of magnesium perchlorate. III. RESULTS AI!D DISCUSSION The analysis of enhancement of acceloratory rate constants with the rise In gaimna-irradiation dose reported /I/ do not show a direct relation between the two (see Fig.l). The parabolic curves are obtained at different temperatures for acceleratory and decay rate constants although decay rate constants are insignificant. Exemplary straight-line plots Ovig. 2) for logarithm of rate constants versus dose for isothermal decoiu^osi- tion of magnesium perchlorate at 3ii0, 335 and 350 °C indicate a complex relation between them. For both the dehydration and decomposition processes; the relation between the rate constants and the dose may be proposed as log kj =CiD + Cg where D is the dost and C-|_ and C? are constants. A number of steps are Involved in both the processes of dehydration and decomposition; which are known /2,3/ to be influe-nced by gamma-irradiation. Dehydration and decomposition RC - 40.1 processes are multistage complex processes. Important steps in the reactions are nucleation and growth. Gamma-radiation produce 3 damaged regions in the samples leading to the formation of dehydration and -decomposition nuclei- The enhanced nucleation in dehydration and decomposition processes is attributed /4/ to gamma-irradiation which produces voids in samples of magnesium perchlorate which fill with dehydration and decomposition gases that ultimately result in causing dislocation by strain and / or stress mechanism. The complexity of the processes may be considered to be reflected by the logarithmic relation between the rate constants and the gamma- -dose. IV REFERENCES 1. D.U. Rajput, Ph.D. Thesis, University of Poona, Pune 411 007, Indie: (1985). M.B. Choudharv, Ph.D. Thesis, University of Poona, Pune 411 007, India (1981). 3, V.G. Dedgaonkar and D.U. Rajput, Radiochem. Radioanal Letts., 57, 281 (1983). 4. N.B. Hannay, Treatise on Solid State Chemistry, Prentice Hall International Inc., London, Vol. 4 (1976). u r k r I 330* C 1- , b , 0 RC - A0.2 CHAKGE PLAT3 TECHNIQUE IN RECOIL STUDY : EVIDENCE FOR COLLECTION Ol" RLCOIL CHAHUE SPECIES FROM SOLID TARGETS ON METAL FLECTKODJCS S.P.Mishra* and (Miss) Jyoti Singh Nuclear & Radiation Chemistry Laboratory Depar tine:, t: of Ch> im s t i y , h.H.U. Varauasi - 221 0 0S (INDIA) SUMMARY The occu ranee of chuge on recoil Mri produced by the (n, •{ ) reaction in po 1 yci y bt a 1 1 i rie KMnO. has been examined using 'charge plate t erhri i que ' . From cons idera 1l oris of capture gamma ray decay schemes and internal conversion it appears that, in a condensed medium, the recoil atoms develops charge after loosing much of its initial kinetic energy winch allows collection on charged e1ect rodes. I. INTRODUCTION Continued problems in the field of chemical effects of nuclear traiisfonn.if ions have centred uruund the kinetic energy and charge of i\:.i.,ii aU;,,^ in tieter;iii. u i ;ig their ultimate chemical fate. In (n, / ) it;actions the neutron capture excites the resultant compound. nucleus to a level equal to the neutron binding energy which is emitted in the form of 'neutron capture gammas'. There are clear evidences that snme of these capture gammas are internally converted ( 1 ) . Subsequent results ( 2. ) indicated that n-capture followed by mterna.1 conversion or electron capture, are subjected to extensive decomposition as a consequence of coulombic explosion. This may give Iise to additional chemical effects. In order to study the charge nature of the initial recoil specie:; I ollowing ( u, -f ) reaction, W ex lor and Davies ( 3 ) worked with ijdbeouH ethyl ho lidos at low pressure in a vessel fitted with electrically charged collector plates. Yoshim and Davies (^ ) made a similar study on recoil atoms from the (n, •(') reaction escaping from thin metallic films. The use of Ag | AgX charged plates in the col Lection of radiohaloiji.n charged species (anionic & cationic) from (.'i,-/ ) activated liquid alkyl halide systems has been well demonstrated and results explained ( 5" ) . The present: work is an attempt to apply the 'charge plate technique1 in the collection of recoil species on metal electrodes when polycrystalline solid targets are neutron activated. II, EXPERIMENTAL A thin layer (thickness : 0. 'J em ) oi permanganate | In (acac) J V(acae), was packed in filter paper cover & irradiated with the help of a 300 mei Ra-Be neutron source having an integral flux of 3.2 x 10 n | cm | sec. Two nickel electrodes of ^2.5 cm. iiameter were placed parallel to each other in contact with the filter paper jacket. Distance between two electrodes was ca 0.5 cm. An electric field of 1000 or 1S00 volts (DC) from a stabilized power supply was applied across the electrodes during the end-period of irradiation for 3.0, 4.0, 6.0, 20.0 & 24.0 ho, The radio- activities collected on the electrodes were quickly counted by the help o£ an end-window G.M. counter and necessary corrections applied. RC - 41.1 III. RESULTS AND DISCUSSION The radioactivities collected for varing times (3.0,^.0,6.0,20.0 & 2 4 hrs ) during end-period of' activation on Hi-electrodes under 1000 V or 1S00 V following 24 hr irradiation of 3 targets (KMnO., In (acac)., V(acac)2) were measured separately with a thin end-window p, tY G.M. counter under conditions of constant geometry. The counting assembly was having a constant reproducible low background count of ca 10c]nnti. and hence the activities collected on electrodes, though of low order, could be well relied upon. No activities wi-re collected on the electrodes in the absence of an electric field. Reproducible collections only in case of KMnO. targets could be observed and no activity on electrodes set.m to nave been collected in the case of other two targets. The collection is apparently independent of time of collection and field applied and a slightly greater fraction of negatively charged species than positively charged ones grit collected. The: origin of charge on recoi 1 atoms has been discussed in some early papers (H,6 ) and idea used in subsequent works. The neutron capture gamma spectrum of Mn is fairly well known ( 1 ). In more than 27% of all cases the initial de -exci tation step is the (^mission of hard gammas of more than 6.75 HeV, which provides the Mn nucleus with a.n outward velocity of more than 3.86x10 cm|sec. Using the Seitz ( -j ) treatment for the determination of the energy loss of "hot" atoms in a lattice it would seem reasonable to assume that only Mn recoi ljspecius oriyi n.ati n ACKNOWLEDGEMENT We thank the CSTR, New ))<.-.)hj for the .iwatd of RA to \)r (Miss) Jyoti Singh. IV. Rlil-liHENCKS 1. G.A.Bartholomew, A. Dove i k S. P. Mishra Nuclear & Radiation Chemistry Laboratory Department of Chemistry, B.H.U. Varanasi - 221 005 (INDIA). SUMMARY : Single crystals and powder samples of HIO3 /DIO3 x-/V -irradiated at room temperature or at 77K and allowed to warm ^upto room temperature, give esr spectra characteristic of the 27-electron radical (HIG 3" or H(H JIOo") . The unpaired electron in predominantly delocalised over the iodine and oxygen atoms. The possible structure of the radical is discussed and radiation damage mechanisn involved is proposed. I. INTRODUCTION Salts of non-metal oxyanions have been extensively studied by radiation/ hot-atom chemists, and in particular, e.s.r. spectroscopy has been widely used to probe the n?ture of paramagnetic centres trapped in such solids in polycrystalline and single c.ystal states. Informations deduced have greatly helped in the understanding of radiation damage mechanism. Alkali-metal perchlorates, perbromates and periodates have been studied, and all the expected primary electron-loss (XO4 ) and - gain (XO* ) species have been reported ^~J' . Similarly, solid state radiolysis of chlorates and bromates^- have provided good information on the unstable intermediates including XO3 and XO3 centres^ . However, these has been a dearth of such information on ic-Jates . A species described as an iodine atom was ^trapped in x-/"^"- irradiated HI03 , and a comparable species, thought to be 10 -"" . was formed in r-irradiated LilO^ crystals . Both these identifications are open to question as the former is chemically improbable and the latter was confusing due to incorrect e.s.r. data deduced. in an attempt to resolve the identity of electron - loss and - gain species formed as a result of solid state radiolysis {x-l~f- rays) in HIO3 , single crystal and polycrystalline materials were irradiated at R.T. and 77K and resulting paramagnetic species studied by esr spectorscopy. II. EXPERIMENTAL* Reagent grade Iodic acid was recrystallised from water or D9 0 before irradiaton, and single crystals were grown from aqueous or deuterated solutions by slow evoporation . lodic acid as crystals or fine powder was "f ~ irradiated upto 2 hrs at room temperature or 77K with a 60 Co vick-rad source (cj» 1.7 Mrad h ). In a saperate series of experiment x-ray irradiation of the mateiial at room temperature or 77K was done with the help of a Philips PW 2184 tube (tungsten anticathode, 30 rnA , 30 kV) for 2 hrs. The e.s.r. spectra were recorded on a varian E109 spectrometer calibrated with a Hewlett - Packard 5246L frequency- counter and a Bruker B-H 12 E field probe, which were standardised with a sample of DPPH. The spectra have also been recorded on a Bruker ER 200 D-SRC spectrometer (coupled with a HP Microwave Frequency counter type 5342A) . Computer simulations on powder spectra were performed on L'NIVAC 1100 system of the University of Geneva, Switzerland. RC - 42.1 III. RESULTS AND DISCUSSION Powder and single crystals; of HI0 3 /DIG 3 x-/~C- irradiated at loom temperature and recorded at 77K gave clear esr spectra spread over ca_ 5000 G with one I nucleus interaction which is assigned due to "IO3 '' type species (e.g4 HIOj or H(H*" ) IO3 ) . In .the central portion of the spectra unresolved and overlapping features due to HIO^" /IOQ /uj" species also seem to be present. Irradiations at 77K gave no hints of "IO3 " type species but central part consisted the usual unresolved features. However, annealing of irradiated materials above 77K up to room temperature gave spectra somewhat similar to those obtained for room temperature irradiated samples. Alongwith one '*"• I _interaction additional coupling due to one HI H could also be realised in the "-O3 " species. Detailed computer simulations using second order corrections (quadrupole effects neglected) on powder spectra have yielded the magnetic parameters for this species in conformity with 102 F(7) • A*s F3 and A*s Cl 3 (8) radicals. The earlier identification of the centre as 1° in# irradiated HlOq appear to be untenable and another report (6) of observing 10 3 species in irradiated LilOj is open to question because wrong esr parameters are derived. Also, in both these works an authentic radiation mechanism has been missing. The fo^owinp mechanism seems to Lit well with observations. Hio3 ^^-^ay_> HI0+- + ,f ... (l) 1I1O3 + e -> [H10] ... (2) + + HI03 -> I0*3 + [H ] ... (3) + + HIO~ * H > H(H )IO~ i.e. H(T)I03 ... (4) Reactions (2-4) are inhibited at 77K but because operative during warming up *-f the low temperature irradiated material or during R.T. irradiation. The "10 3 + type centre (H(H )IC>3 H(H)I0 3 )+is stable at room temperaturt-+ due to a charge compensation effect rendered by (H ) from the neighbouring HICs species in the unit cell consisting of four HI0-, molecules. ACKNOWLEDGEMENT I thank Professor M.C.R. Syir.ons, FRS, University of Leicester, England ~nd Professors E.A.C. Lucken & M. Geoffroy, University de Geneve, Switzerland fcr encouragements and helpful discussions. REFERENCES + Most of the experiments performed at the chemistry Dept., Leicester University, England and Department de Chimie Physique, Universite de Geneve, Switzerland , 1. J.R.Byberg and S .J ,K .Jensen. J. Chem, Phys., 52, 5902 (1970). 2. J.R. Byberg, J. Chem. Phys., 55, 4867 (1971). 3. M.C.R.Symons and S.P. Mishra, J.Chem. Soc. (Dalton, 2183 (1981). 4. S.P.Mishra and M.C.R. Symons, J. Chem. Soc. (Faraday 1), 7£, 747 (1976) & references cited . 5. C.E. Bailey, J. Chem. Pbys., 59, 1599 (1973). 6. V.O. Martirosyan, M.L. Meilmen, I.N. Marov & V.V. Shukov, Phys, Status Solidi, BJ£, 791 (1975). 7. S. Subramanian and M.T. Rogers, J. Phys. Chem., 75, (22), 3479 (1971). 8. S. Subramanian and M.T. Rogers, J. Chem. Phys., 57^ (11), 4582, (1972). RC - 42.2 liFFHCT OK Ul'-.MICAI. KKACTIVITY OF MF.DIUM IN THE REACTIONS OF KNKRI;KJ/I: HKOMINJ-: SI-KCIKS S. f. Mishr.r- M.R. Tainan Di-pt. of Chein isti y Dept. of Applied Chemistry Banaras Hindu Uni vors i ly fi Chemical Technology Varanasi - 5, (mii:i. fi Raj shahi University Bangladesh SUMMARY : The buhav ii»ii' uf ener'getic bromine species ( Br, Br and Br) originated troin (11,/j activation of,, dibromoniethane (DBM) ard tribruaiouieUu.il- ('iiiM) has uoen studied by 'charge plate technique)' {CPT> . i liie iv)i(ir,i:i I' .' th'.th' ' i.-ijuiii;: siiccius Hiee order iiss CIIDr^^- Cli, Hi-, . It i-j cui'iuiuded tlun tin; i'oi-iiifUirtn -collectiun of charged species in these liquids iw governed by Uiij chnm icil ix'aci i v 1 ty of tlio niedium7target molecule. (Key iarjn' |il;i!i.' t.;chrii()iH.', refu:tivity of medium, energeiic i'om i 111; s 1 '(•(.ii•?.-,; . I. INTRODUCTION : luierj-jci io ii.dutien atom reactions in liquid phase are usually complicated mainly because of the considerable interactions between the medium and the recoil species. It is not surprising that the properties of the medium have extremely great influence on the final chemical stabilization of recoil species and hence on the electrode yields. The present paper reports the effect of eheiii ical rearti"iIv of medium .>n the stabilization of high energy bromine atoms originated from the (n ••/) activated DBM and TBM. It. KXriilUMKNTAl. The: experian.-ntal |irui.i:i]i!iv, h.i-1-up counting and analysis essentially the same as reported in our uirlior couimuaicalion'. ' III. ltliSULTS AN!) DISCUSSION From th.i time-d«i:ay analysis nt radioactivity data, the respective yields for 00m Br, tlUer and U^Br on each electrodes in DBM and 'I'BM wore obtained . It ib obsc:rvi:u that both 1 ..si lively cind negatively charged species are produced and collected on charged plates during (n,/) reaction in these compounds. KesuUa reveal that eoliec tion oi negatively charged species on the anode for all the; radiohromines fall in the order : CH2 B12 "?• CHBra whereas for the cat ionic species the 01•der is CUBr37-0112Br2 . These results are quite similar to earlier findings on chlorobromomethanes reported from our laboi'u lory . ' 2 )jt Wilii nbsi-rved lhat tor ^ BrCHCl2 and BrCCl 3targets, the (jollocied aciiv i! ii'::; on tin; caihock : full in the order BrCCl 33 yy BrCliCipp- HrCHoCl but revt'isi. is the ca::.e Tor ;mode i.e. Br CH.,Ct> BrCHClBCHC^^ BrCCl.BCCl 3 This observation w.i:i explained on [he basis that deprnton&tioii reactions would be favoured in case of BrCll2Cl and BrCHCl2 targets and no such reaction H C - . 1 is possible in case of BrCCl3 and hence one would expect a larger fraction of dimercations loaned in the order recorded above; the result is increasingly higher yields on cathode. A riverse order on the anode is explained on the basis of decreased formation of anions containing *Br (*Br or ;Bri!l ) or alkyl bromide anions. Accordingly, a larger fraction of *!SrlM than Cfo in the chlorobromomethane target is expected in the order : Brdi^CI > BrCHC II BrCCl 3. Our results may be explicable on the basis of the mrchanism proposed above for Chlorobromomethane if we consider one bromine atom in IJBM and TBM as a foreign atom and with the assumption tiiat the relative case of doprotonntion is higher in DBM than in TBM. The difference in the chemical composition between DBM and TliM is that one hydrogen atom has been replaced by ono bromine atom . If our proposed mechanism is corret then one can draw the conclusion that two bromine atoms in DBM or three bromine atoms in IBM are not undergoing (n,y) reaction simultaneously or in succession. Because, if this was the case then naturally one could expect higher yield on both the electrodes front TBM activation rather than from DBM. As our results in ossenre clearly start against this supposition wo conclude that the format ion/cul lection of charged, succies in these liquids has been governed by the chemical reactivity of the medium/target molecule. It is wxirth recalling herethat the similar were the findings of Glueckauf and Fay who observed that activities retained in organic form in DBM and TBM were 57% and 67% respectively hut they wore unable to give any satisfactory explanation for these variances. ACKNOWLEDGEMENT One of us (MR/.) is thankful to Govt. of India for giving him a 'bilateral cultural exchange scholarship1. IV. REFERENCES 1. (a) Mishra, S.P., Singh, N.P. Radiochiin. Acta 29, 75, (19)!1). (b) Mishra, S.P., Singh, N.P., Zaman, M.H. Radiochemis try and Radiat. Chem. Symf)., BARC, Bombay, Preprints Volume, HES 15-1 I 2. Singh, N.P. Ph.I). Thesis, Banaras Hindu University (19«1). 3. Glueckauf, li. , Fay. J.W.J. : J. Chem. Soc. 3H4,3'JU ( RC - 43.2 RADIOTRACKR TECHNIQUE IN ADSORPTION STUDY : A CASE OF EFFICIENT REMOVAL OF Sr(II) FROM AQUEOU5 SOLUTIONS RY MnO^ POWDER * Shuddhodan P . Mishra and Uhanesh Tiwary Nuclear arid Radiochernistry Laboratory Department of Chemistry Banaras Hindu University Varanasi - 221 005 INDIA . SUMMARY : The adsorption of microamounts of Strontium (II) t.'om aque.n:. solutions on manganese dioxide has been investigated and optimised a^ a function of contrct time, sorbate concentration and temperature. 1 hr- equilibrium is essentilly achieved in £a 15 min The kineti-s of a^sorpt i-,.-n is found to be of first order, the system obeys the 'Freundlich adsorption isotherm1 over a wide range of Sr (II) concentration (10~3 - 10~7 M) and the process of uptake is found to be endothermic in nature. (Key Words : Adsorption, Freundlich isotherm Sr-89 , Manganese dioxide.) I. INTRODUCTION A gi-eat deal of interest lias been manifested in adsorption stuai&.> of metal ions, especially of metals involved in fuel reprocessing ami fission products on metal oxide surfaces. Strontium is a product oT fission of uranium and has an importance from waste disposal point . The adsorption 1 of Sr (II) on metal oxides "3 [las been the subject of several investigations in past. In connection with the efforts in understanding the role of manganese dioxide in removal of Sr2' ions from aqueous solutions adsorption of Sr2+ ions by B-Mn()2 in aqueous solutions has been investigated under various physical conditions. II. EXPERIMENTAL Manganese dioxide (AR/B.D.H.) was activated at 573K in ai> for C£i 24 hrs, cooled slowly to room temperature and then sieved t. obtain particle size 120-170 mesh. The surface area was determined arm was found to be 9.6762 ± 0.1343 m^ g"*l . A careful x-ray analysis oi the sample revealed that it is B-M11O2 and its structure to be tetragonal (rutiie) . The adsorptive solution of desired concentrations were prepared from a stock solution of 0.1 M Sr (NO3 )z by successive dilution. The adsorption of Sr2+ ions was studied by taking 0.1 g of MnO2 in 10.0 ml of sorbate solution lebelled with Sr-89 in the form of Sr (NO^)^ in dilute nitric acid, obtained from BARC , 'J'rombay . The ex peri men till procedure for for measurement of adsorption and estimation of amount adsorbed were the same as reported earlier(5) . III. RESULTS AND DISCUSSION In preliminary studies, the influence of contact time between sorbate (10"-1 M) and sorbent (0.1 g) on adsorption yields was examined from aqueous solutions which indicate chat the time-growth of adsorption RC - 44.1 J i: •••••vy fast and chemical equilibrium is established in ca 15 min. The :.•• The temperature dependence study (iCn-TJB K) also ;;u;'p.'M :••; the above deduction where the ri !r.f.;n I adsorbed at equilibrium imrt-a^es with temperature (of Table 1) without any significant change in the ""."jcral nature of time-rate curves and equilibrium time. The plot of ! oj; (.->(. - at ^ -— ti-ne gives straight line which shows that the kinetic; of lue j-niccss is of first order. Based on Thermodynamic data the nature oi iduorption is deduced to be of activated type. These studies indicates tli.it MnOv powder can be used as a sorhent for efficient removal of Sr ions from aqueous solutions. 2 + Table : 1 Adsorption of Sr Jons on MnO? Powder 1 I; ' ti.il Concn . 3 [ 5 6 1 io~ io' ID' i.o"' of Sr(Il) (mol r- ) Amount adsorbed at O.4631 0 .90H0 0.9673 0.9771 0 .9803 1 4 5 6 7 equil ibrufn'tnol E x 10~ X io- x 10~ x 10" x lu Temperature 303 313 323 333 (K) Amount adsorbed at 0.9613 0.9672 0 .9728 0 .980V equii ibr nun ( mol gM x 10-6 Distribution Coef f . 24.82 .?:/. VI ./.S (mol g *) x 10 "initial concentration of Sr (NO3), = 1 .0 x 10 M) ACKNOWLEDGEMENT We thank U .G .C. New Delhi for financial support. IV. KEKKKliNCliS 1. R. Rao Gadde and H .A .Laitinen. Anal. Chem 46, 2022 (1M74) 2. S.M. Hasany and M.A. Qureshi, Int. J. Apfil . Kadiat . Issot . ij., 747 (IV81).' 3. M. Raslnd and M. Ezaz, Int. J. Appl. Radiat . It,ot 37, 501 (19K6). A. S.P. Mishra & S.N. Singh, Int. J. Appl. Rad. Isot . 38^, 541 (19U7). RC - 44.2 80 80m 82 RECOIL Br, Br I Br IN Ba(BK>3)2 .H2O Shuddocian P. Mishra and A.B.R. Tripathi Nuclear and Radiation Chemistry Laboratory Department of Chemistry, Banaras iiindu University Varanasi - 221 005 INDIA . 8 80n 82 SUMMARY : The rotenlion of °Br, Br and Br in Ba(BrO3)^ .II2O activated at room temperature with thermal neutrons is found to be 23.0t,, 2-4 .0 % 61 25.0^ and annealing at 150°C gives an increase in retention to 36.0%, 36.0% & 42.0% respectively for the three radicbromine species. A detailed kinetfc study of isothermal annealing is presented and mechanism of reformation of the parent molecules is proposed. INTRODUCTION The tention of radiobromines formed by thermal neutron capture in alkali/alkaline-earth metal bromates had been studied extensively in the past (1-7). The bromine activities appear to have been lecovered as BrOf ,BrO2 BrO~ , Br ~" and possibly BrO,j~ in somebromates but little or no BrO-T , BrO~ and BrO,j could be detected. In our present study re-visits on such findings 80 SOm on recoil Br, Br & 82 Br in Ba( BrO 3 )2 -H2 0 target and details on thermal anneaiii.g data are presented. EXPERIMENTAL Barium broruate monohydrate was synthesised : 350 g of KBrOj was dissolved in 600 ml of distilled water under almost boiling condition, -10' I ml hot solution containing 240 g of barium chloride dihydrate was added with continuous stirring when precipitate of barium bromate monohydrate was formed , after decontation the precipitate was washed with distilled water several tines until free from soluble chlorides. BrO 3 and cation estimation showed the purity of the product to be greater than 92% . 3.0 g of sample was activated with thermal neutrons for 24 hr using a (Ra-Be) neutron source having integral flux of 3.2 x 10 n cm"' sec"' with concomittand r'-dose rate of 172 rads/hr. Thermal annealing experiments were performed in an electronically controlled oven with a precision of ± 1 .01 for varying periods (15, 30, 45 and 60 min) . In each case the rat'iobromine fragments in two stable oxidation states (e.g. Br & BrO 3"" ) were separated by fractional precipitation method' ' Radioactivities for difierciit time intervals were recorded with the help of an end-window G.M. Counter under conditions of constant geometry and retention values obtained after necessary corrections. RESULTS AND DISCUSSION Percentage retentions were calculated for different recoils br, mBr ft Br as reported earlier9 . It was found that, retention first increases and 8urn then slowly reaches to a plateau in the coder of #2 Br ^> Br ^ 80 [>r al the temperature studied. Harbottle and Jaclr-*^ reported a very pronuunord difference in the annealing behaviour of the two bromine isotopes in m-i. ::•'•. hC - 45.1 irradiated potass)""' bromate. In this ra = e t:ie initial retentions w< re .i!most the 6arue, but the course uf the annealing reactions were quite different (' . 'i'vie heavier brom^n^ isotope &T TtfoY-nifrd hTomrtt moie readily I'raTi ".Vie during the earlier" fast portion of the ar.i,ea) ing process, showing that the effect is quite different from the usual kinetic effect. Sbahkar et al had explained this on the basis of formation and involvement of a risetastable species in their annealing studies in co-acptylai.etor.ate targets To le;irn more about kinetics of thermal anneal in;; j-i-jctions )n neutron x activated Ba(BrO^)2 •*'? ^ ^e rate con: tints were compiled tmni the slop.- of l"g(l^oo"^f ) against time of heating curves. The resolutio . of above • irves reveal;; the presence of first order kinetics. The values oi a< . iva' • n • ner;;y were calculated from the plot of log k \s_ 1/T. Annealing data coupled with Fletcher lliov.n model at oifferent Icmperatures (50, 100 & 150°C) are combined to obtain a single curve of i-'.(iiiva)eiit antii'a''nB at* single reference temperature showing the validity of Fletcher-Brown m^del'-l jn OU1- present study. A good agreement between theoretical and experimental curves were seen at lower temperatures, while at higher temr)«rature«»/ the deviation indicates tLtat some of ttie anuealiu^, process are governed by different mechanism Oiher than simple fii°t orJir recombinal i'in . The Fletcher-Brown activation energies were obtained from 11.f S 1/T ..lot of l-(!>7ef 'T(T) ^ - The calculated values of activation energy lio::i Fletcher-Brown and Arrhenius treatments were found in the order : 8^ Br < 8o Br <80mBr. The I'lotclmr-Brown activation energies are generally found to bo greater than the values obtained froin Airhenius plots, which is in line with an earlier report [J). !• letcher-l^rown activation energies are governed, by slow and fast two first order p. i.cesses and the values obtaine ' are in !act from a composite annealing curve. An.iealing may consist of recombination of the vacancies and the interstitials : BrOj * 0 (1) BrO . (2) Also, tl...' composite annealing intact, contains all the three recombination process iiontiiliu t ing to retention • ACKNOWLKDGKMENTS VVI: thank the CS1K , New lielhi foi a project grant. HCIKRKNCKS 1. C.E. .I'.oyd '„ Q.V.. Larson, J. Aii:. Chen. Soc .90 254(1968). 2. D..I . ,Api;rs,J .Jach et al , Radiochini. Acta, 4_, ~193 (1963). 3. T., Andersen, H.E., Lundager Madsen and K.,01esen Trans Faraday Soc. 4. ITU. ,C,,nij,bel I andC.II.W., Jones .Radio.:hiin, Acta, 9, 7 (1968). 5. K.Jach and G. .ilarbot tie ;1 Trans Faraday Soc, 54, 520 (1968). (<. V, ,llarl....tt le.J .Am .Chtn .Soc - , 82, 805 (1960). / . 5.K. .Vcljkovic it'id G.llarbott le.J. Inorg .Nucl .Chan. 24, 1517 (1962). K. G.E.,Uoyd and U v . Larson ,J ,Ani .Cheni .Soc . 91^, 4639 (1969). ') . S.P. .Mishra 8. A .B .R .Tripathi, Proc .Radiocheni. and lladiat . Qiem. Sym., Najipur, RE-:i, (1990) . '.0. J.,Shanl;ar, K.S. Venkat etiwarlu 4 M. l,al, I'Voc . Synip. Chan. Effects Nucl. Transf., IAEA, Vienna, \, 417 (1961). !!. R .C. ,t'l etcher, and W.I..,' Brown . t'hys . Rev. 9^, 5H5, (1953). RC - 45.2 ADSORPTION UP UAKIUM IONS ON SOUluWinTANAf RADIO FHAUEH o Shaddhodan P. Mishra l\ N. drinivasu Nuclear and Kadioohemistry Laboratory Department of Chemistry Banaras Hindu University Varanasi - 221 U05 (INDIA) SUMMARY : Adsorption of barium ions on sodium tit-anate has been ^tudied as a function of concentration (10 -10 M) 8 temperature (3O5-335K) using radiutrucor technique. The percentage adsorption increases with increase in bulk dilution -attaining a value of ca Jtf'fe at lower concjnrrat )• vis. The adsorption process .appears \.> ,je fast c - • jef;: :i.u to a saturation value in ca 30 uiiu. at all concentrations of adsorbate. The kinetics of adsorption follows first order rate law and obeys Froundlich isotherm. The .amount adsorbed on adsorbent surface decreases with t lie increase in temperature. (Key words : Adsorption, Inorganic ion e.:.:hu '.ger, Rndioactivo waste, Kadlotracer) I. INTRODUCTION Inorganic ion-oxchangers ha-/ found apol ica t ions in metal 1 jn seoarations mainly in the nuclear energy industry for separation of selected radio-nuclides from the wastes of reprocessing plants. These materials .. have good stabilities against hiijh temperature and radiations . Earlier workers have reported the sorptio-i properties sodium ti tanate towards soma ions 1} ife report here th? sorution af barium t vis on sodium Utanate as a function of c Mi\eutr:nion and temperature. II. EXPERIMENTAL The soroti/e properties of sodium tii.nidto was studied i la ing sorbent samples prepared by the reaction of rn.iniui;: tutrjchlonde with sodium hydroxide in aqueous solution. The adsirhunt was dried in 3ir and stored in a desiccator and then sieved to obtain size 120-170 mesh. Ba(Uu }} s ilutio.as of required cuiicun (ration wore obtained from a stcjtc" solution of 0.1 M BJ(NO.]., by succossl/o dilution. The adsorptioa of barium ijns was studied'' by taking O.lg sodium titanate in 10.0ml of Ba(No )„ labelled wim .lJa-M0 carrier free as nitrate in dilute nifii: acid uotained from the BAKC, Trimbdy. The experimental procedure for measurmeut oi adsorption and the estimation of amounr adsoroed were the same .as reported tarlior III. KBSULTb ANU DISCUSSION The vasu'i 13 jf concentrdtio.i dependence study on adsi>rption Gf barium ions on sodium titanate are rooorted in Table 1. The adsorption of bjrium ions increases very rapidly in the beginning and most of it is adsorbed within a short time uf ca 3D min. and there after increases vory slowly and a tea ins a psoudos i turat i in value. Table 1. Shows th.it percenia;!e adsorpUni in. leases '.vitli the increase in bulk dilution. This imuoise is due to tiu: avai !abi 1 ity of laryer sorbent kC - A6.1 surface sites for relatively lesser number of barium ions at higher dilution. The temperature dependence of adsorption of barium ions on sodium titanate (cu Table 1.) shows that percentage adsorption decreases with the increase in temperature. The plot of log (a -a ) yj> time gives straight Table 1 : Adsorption of Ba on Sodium titan.ite Initial concentration 10"2 10 "3 10 "4 10~5 10 "6 (mol 1 ) Amount adsorbed at 0.3465xl0~30 .9564xlO~4 0 .9722xlO~50 .9752xl0~60.9860xl0~7 equilibrium (mol g J Temperature (K) 3H5 315 325 335 Amount adsorbod at O.9H12 0.9787 0.97 45 0.9630 equilibrium (mol g~^ ;cl 0 ~^ Distribution Couf. fi2 .43 46.11 37.49 26.03 (mol g-'jxH)2 5 * Initial none, of Ba(No3)? = 1.0xl0" M. line which cloarly shows that the kinetics of this system follows frlst order rate law. Since the process is found to be favoured at lower temperature which indicates that the uptake appears to be physical typo. ACKNOWLEDGEMENT : We thank the DAE, Govt. of India for financial support. IV. REFERENCES 1. J. Lehto, L. Szirtes. Radio Che:.i. Radioanal. Lett. 50, 375 (1982). 2. J. Lohto, O.J.Heinoneit , J.K. Miettinen, Radiochem. Radioanal. Lett. 46, 381 (1981). 3. V.S. Dubrovin, S.I. Malimonova, Radiokhuniya 27_, 465 (1985). 4. S.P. Mishra, S.N. Singh, J. Appl. Rad. Isot. 38, 541 (1987). RC - 46.2 AR - Applications of Radioisotopes and Radiations Papers : - AR -01 to AR -51 EXTRACTION OF SILVER (I) WITH AMBERLITE LA-2 * A. Chatterjee & S. Basu Department of Chemistry, The University of Burdw3n, Burdwan 713 l'J4 SUMMARY s The use of Amberlite LA-2, a secondary araine^in the extraction of Silver(I)-thiosulphate complex from aqueous media at pn-J is reported. The effects of diluents, concentration of thiosulphate and amine, presence of diverse ions etc. have been studied. ^~Key Words i lAmberiite LA-2), SilverU)-thiosulphate complex, Ag-110m_7 INTRODUCTION j The use of high molecular weight amines as liquid anion exchanger for the extraction of anionic metal complexes has been reviewed from time to time (1,2). However, only a little attention 13,4) was paid for the extraction of Silver. The present investigation describes an att- empt for the extraction of Silver(l) labelled with Ag-llOm, t>y Amberlite LA-2 dissolved in iso-amyl alcohol as inert diluent. EXPERIMENTAL : Amberlite LA-2 £ Aldrich Chemical Co._/ was used as such. All other reagents used were of AR grade. The radionuclide Ag supplied by BARC, was measured by a / -ray spectrometer with a Nal (Tl) detector. In the general extraction procedure, the aqueous phase was a mixture of thiosulphate solution of desired molarity Ivarying from 0.1M to 0.5K) and the silvernitrate labelled with Ag-110m. The organic phase was com- posed of Amberlite LA-2 Ivarying from 0.01M to 0.0W) was dissolved in suitable inert organic diluent. Equal volumes of two phases were mixed in a seperating funnel. After 2 minutes, lml of each phase were counted by usual methods. Experiment was performed in different organic solvents and also in presence of different diverse ions to study their effects on the extraction. RESULTS AND DISCUSSION : Variation of aqueous pH shows that the extrac- tion is maximum at pH-3. The complex is known to be most staole at this pH also. Variation of thiosulphate and amlne concentration shows that the extraction is maximum at U.bM thiosulphate and 0.05M amlne concentration, the most efficient diluent being isoamyl alcohol 199.54%). AR - 01 .1 The presence of sodium or potassium salts of oxalate, citrate, tunga- tate, molybdate, nitrate, nitrite, sulphide do not affect the extraction. The effect of other sulphur acid or halogen acid can be removed by pr^e- treatment with concentrated nitric acid. Among the cations, carrier amount of Zn(II), Mg(IIi, Cu(Ii;, FeUIJ, Fe(IIi>, Altll-W, ZrUV) orTiUV) do not • affect the extraction at all. The stoichiometry of the; extracted species were determined both by substoichiometric and slope ratio method and results indicated a 1 t 1 composition. REFERENCES : 1. H. Green, Talanta, 20, 139 11973). 2. T. Sato, J. Inorg. Nucl. Chem., 28, 1461 U966> . 3. A. Alian, Mikrochim Acta, 981 (1968). 4. R.J.T.Graham & A.Carr., J. Chromatog., 46, 293 (1970K 120.00 -| C o 80 00 - X UJ o 40.00 - O Q_ 0.00 i i i r i i i i | i r i i i i t til11 0.00 2.00 4.00 6.00 pH Fig. 1 pH — variation of extraction 0; Siivsr(l) AR - 01.2 TriK iNf-'r.f/CHci; OK OIT.UKNTS ON SYNKKOISM IN THK KXTUACT ION ot" KllltOP I UM fi Uaijhupal.hy and M fiudersanan DepciiI m<:nl of ('hemi si i y K .('. College ot Science liorobay 400 0?0 ( Analytic,il ci,, mi::» < y Division, l)A):c, Jw.ml..• y loo on1,) SUMMAKY The e1 J ; poltii AR - 02.1 change of diluent was much larger than the change in extraction in the absence and presence of the donor. CONCLUSIONS: Extraction was more in the caae of aliphatic or non-polar diluents than in polar diluents, introduction of an electron withdrawing group l«d to a reduction in extraction. In dichloroethane and chloroform, ine ised chlorinatjon Jed to an increasing polarity as well as an increased tendency for H-bond formation. The incieased interactions between the metal chela tf:s and diluent as weJl as donor and diluent led to a decrease in extraction. On the basis (it .regular solution theory, a relationship was tound between the partition coefficient or the metal chelate and the solubility parameter,b . It was also tound that a linear correlation was observed between file nature of cheating agent and the stability constant of the chelate expressed by log >yN ' N log )'RTTA . PKd -f const where )' and fi represent the partition coefficient and slabilit;- constant of the metal chelate and pK represents I he dissociation constant of the cheiating agent. Some of the results ace presented in Tables 1 and 2. ACKN0WIiKJ»GKHKNT5!: The authors wish to thank JJr. A.K. iliindaram and Or.M.Sankar Das for guidance and laboratory facilities. The award ot a fellowship to S.H. under Faculty improvement Programme of UGC is gratefully acknowledged. TAUl,)•: 1 Diluent. 6 (: 1/I: log K. log iy ory Oenzene 9.15 2.27 0.44 6 .31 16.62 Chloroform <) 30 4.(11 0.21 Chiorobenzene 9.50 5.62 0. 10 5 .45 16.6.3 Nitrobenzene 10.00 34. (I2 0.03 4 14 16.76 liexane 7.30 1.89 0.53 6 .61 14.54 Dichloroethane *: Values for Eu-HTTA chelates - TAIILK 2 Diluent 1 og v J ( f 1 log 1'. . m " ' "HA1 " t. log 3 Denzene 1.73 8.01 16.62 20.00 22 93 Chloroform 1.(10 (!. 17 16 64 1(1 ')4 Chlorobenzene 1 .65 7.93 16 6.1 19 84 22.08 Ni l.iobenzene 1.60 7. (Kl 16.76 1>1.6(1 20 ')0 Hexane 0.U5 7.1.3 14.54 18.70 21 15 Dichloroethane 1.65 7.93 16.42 1(1.0(1 20.4r. CCi 1.43 7.71 15 61 22. 15 Values for »ono and diadducts of Ku-HTTA with AR - 02.2 EXTRACTION AND SEPARATION OF Zr. Nb. 8 Hf BY ALIQUAT 336 AND ITS MIXTURES WITH NEUTRAL DONORS FROM AQ. HC1 6 THIOCYANATE P.K. Mlshra. V. Chakravortty and K-C^Dash Department of Chemistry, Utkal University, Bhubaneswa India SUMMARY- Extraction of Zr(lV) by Aliquat 336 is achieved only above; 6M HC1, whereas presence of thiocynnate inns result in the extraction of the same from low HC1 concentrations (0. 1 -4. HM). Mixtures of Aliquat 336 with neutral donors like 1K)SO, THI1 fj I'UPO result in synergism in the extraction process. By suitable regulation of extraciant and acid concentrations, separa- 9ri ()5 9r> 1(11 tlon of ' Zr- Nb and ' Zr- III is possible. (Key words: l-lx t raotion, Zi-(IV), Nb(V) I) llf(IV), Ali(|iiat Hti. Synergism) I. INTRODUCTION- Extraction of Zr(IV) has been achieved by Aliquat 336 (tricaprylmethylainmoniuiii chloride) from strong IK.l media and from thiocyanate solution containing lower IK.l concentrations 1,2/. The present study reports extraction and separation of ' Zr- ' Nb fi llf by Aliquat 336 and its mixtures with some neutral donors. II. EXPERIMENTAL- Aliquat 336 supplied by M/s llenkel Corp., USA. was used after purification • ' . DOS! I was purified by the reported method !) r> 'IS 1H1 /3/. Radiotracers ' Zr1, Nb li III were obtained from HARC, Bombay. The activity measure'iients were per'formed by using a high purity lie detector coupled with multichannel analyser ND-6.S. The procedure for extraction and determination of these elements has already been reported /4,'.'>/. III. RESULTS AND DISCUSSION- Extraction of Zr(IV) by Aliquat 336 starts above fiM IIC1. Mixtures <>l Al irjua t 336 with neutral donors like DOSO (dioctyl yulphoxide) and THI1 (tributyl phosphate) result in synergism In the extraction process. It has been observed that extraction by mixtures of 10% (v/v) Aliquat 3 It) and 0.01 M I)OS(), or1 5% (v v) Aliquat 336 and 1% (v/v) ['HP is always higher than that by individual components at 7M MCI. The extracted species in the above extraction process is found to be fJV.iCl.nl,, where | (J - U.N(CII.) ami I. = DOS(» or THI'; lor I. - D(JSO. I b 3 3 n = 1 and for 1. - THP. n At low acidities, i.e. U.I lo 4.SM lli:i, ap[)ieci AR - 03.1 be associated with 2 TOPO molecules either by extraction of Zr(IV) with TOPO alone or its mixtures with Aliquat 336 from acidic thiocyanate media. By regulation of KC1, SCN , and Aliquat 336 concentrations, separation of 95 95 181 Nb, Zr S Hf is possible. It has been observed that under other- 95 wise identical conditions, extraction of Nb increases with increase in HC1 cone, whereas extraction in case of Zr and Hf initially increases, reaches a maximum and decreases thereafter. With increasing thiocyanate cone, at fixed HC1 acidity, the extraction follows the order Nb> Zr^> Hf. With increasing Aliquat 336 cone, the extraction follows the order: Nb> Zr y Hf. Thus with gradual increase of Aliquat 336 cone, from 0.8 to 8% (v/v) at 1M HC1 and 1M KSCN, Nb and Zr can be extracted into organic phase leaving Hf behind . The extraction of Hf requires still higher Aliquat 336 cone. (16% (v/v)) and higher thiocyanate cone. (2.5M). Thus at 4.5M HC1 and 1M KSCN, the separation factor ( Nb/ Zr) is 3675 and extraction with 0.8% (v/v) AMquat 336 assures complete separation of Nb from Zr. with 98% Nb being extracted into organic phase in a single extraction step. ACKNOWLEDGEMENTS- Thanks are due to Prof. S.N. Bhattacharyya, Head. Nuclear Chemistry Div., Saha Institute of Nuclear Physics, Calcutta for laboratory facilities. IV. REFERENCES 1. T. Sato and H. Watanabe, Anal, Chim. Acta, 49, 463 (197C). 2. T. Sato, H. Watanabe, S. Kotani and M. Yamamoto. Anal. Chim. Acta. 84, 397 (1976). 3. M. Chakrabortty, V. Chakravortty and S.R. Mohanty, J. Radioenal. Chem., 53, 131 (1979). 4. P.K. Mishra, V. Chakravortty, K.C. Dash, N.R. Das and S.N. Bhattacharyya, J. Radioanal. Nucl. Chem., Articles, 131, 281 (1989). 5. P.K. Mishra, V. Chakravortty, K.C. Dash, N.R. Das and S.N. Bhattacharyya, J. Radioanal. Nucl. Chem.. Articles, 134, 259 (1989). AR - 03.2 RPEC Separation of Zirconium, Niobium and Hafnium with HDEHP N.R.Das and Sushanta Lahlrl Nuclear Chemistry Division Saha Institute of Nuclear Physics 1/AF, Bldhannagar, Calcutta 700064 Summary A reversed phase extraction chromatographic (NPEC) procedure for radlochemical separation of zirconium, niobium and hafnium has been developed using d!(2-ethylhexyl)phosphorlc acid (HDEHP) as the liquid cation exchanger. A mixi-ure of radiotracers, 95Zr, 95Nb, 181Hf, were first extracted by HDEHP Impregnated on Kieselgurh packed in a column and were sequentially eluted out with mixtures of oxalic acid + H SO + H O solutions a' vnrylng concentrations. The purity of the separated species were studied through y-ray spectrometry. (Keywords: RPEC, IIDBHP, Zr, Nb, Hf, jr-ray spectrometry) 1. INTRODUCTION The elements, zirconium and hafnium, because of their homologous properties occur together in nature. Similarly, zirconium and niobium form an Intimato pair very difficult to separate. Thus the separation of these elements in a mixture, especially in microscale concentrations, always constitutes an important problem. The present study deals with the development of a simple radiocheitilcal procedure Involving reversed phase extraction chromatography (RPEC) for separation of Hie elements In tracer scales using dl(2-ethylhexyl)phosphoric acid (HDEHP) as the liquid cation exchanger. II.EXPERIMENTAL ' The reagent, HDEIIP, was obtained from iCN Pharmaceuticals. Inc. Radiotracers, Zr (65d), °Nb (36d) and Hf (42.6d) were procured from BARC. Trombay and was processed as per requirements, n-Heptane was used as the diluent for Uie exchanger. AR -• 04.1 For separation of tracers In a mixture by RPBC, Kieselgurh used as the inert column material was first hydrophobised with dlmethyidlchloro siJane and then Impregnated with HDEHP. The chromatographic column of the reagent impregnated Kieseiguhr (3 cm x 0.6 cm) was prepared in a stoppered glass tube and the flowrate of the eluates percolated In the column through a syringe was regulated with a peristaltic pump. Specific volumes of effluents were coiiected at regular intervals and the activity present in the effluents were characterised and measured with a ND66 8K channel analyser connected with a HPGe detector. III. DISCUSSION In the RPEC system, the separation profile oT the elements with 1 2 HDEHP closely resembles to that of solvent extraction.' The affinities of the elements towards HDEHP follows the order of Hf>2r>Nb which are in conflrmity with their properties of the complex formations under the experimental conditions . When the column was leaded with the tracer mixture, niobium which has least affinity for the catalon exchanger comes out almost unabsorbed and H it was eiuted out completely with O.'N ooxalic acid + 6N H SO + o°o solutions. Zirconium and hafnium wore later eiuted out from the column simply by lncresing the oxalic acid concentrations to 0.6N and 2N reflectively in the mixture. The purity and the extents of separation were verified by taking recourse to y-ray spectrometry. The method is very simple but highly effective. IV.REFERENCES 1. N.R.Das, B.Nandi and S.N.Bhattacharyya. J. Rad'ioanal. Nucl. Chem., 62, 53, (1981) 2. N.R.Das, B.Nandi and S.N.Bhattacharyya. Int. J. Appl. Radlat. Isot., 32. 306, (1981) 3. N.R.Das and P.Chattopadhyay, Bull. Chem. Soc. Jpn., 61., 4423. (1988) AR - 04.2 RAPID EXTRACTION AND RADIOCHEMICAL SEPARATION OP Cr(IIl) AND EuUH) WITH ALIZARIN INTO DIPPERENT ORGANIC SOLVENTS. S. D. Jadhav, A.M. Amlani and Z. R. Turel. Nuclear Chemistry Division, The Institute of Science, Bombay- 400 032, India. SUMMARY Alizarin has been used for the rapid extraction and radiocheraical separation of Cr(IIl) and Bu(IIl) into different organic solvents. 53-Cr and 152+154EU were used as tracers to study the extraction process. (Key Words : Solvent extraction, Or, 5 + ^Eu, Alizarin) I. INTRODUCTION Alizarin reacts with a large number of metal ions like tin, gallium, aluminium, calcium, iron, zirconium, etc. to form coloured lakes « A literature survey indicates that this reagent has not been applied for the extraction of Gr(III) or Eu(III). The present investigation deals with developing a rapid and selective method for the extraction and radiocheraical separation of Cr(III) and Eu(III) using their respective tracers. II. EXPERIMENTAL i) Extraction procedure for Cr(III) Al-ml solution containing 0.75 mg of Cr(lII) labelled with •* Cr was taken in a 150 ml separating funnel. 10 ml of 0.15^> solution vf alizarin in methanol was added and the pH was adjusted to 8.0 with concentrated ammonia solution. The volume of the aqueous phase wag made to 16.5 ml with distill- ed water and it was equilibrated with 16.5 ml of cyclohexane for 5.0 min. The phases were allowed to separate and a 2-ml aliquot of each phase was taken for counting on a gamma-ray spectrometer at the channel number corresponding to the 0". 323 HeV photopeak of 51(jr. The extraction coefficient value vvas calculated in the usual way. ii) Extraction procedure for Eu(III) A 2—mi solution containing 2.0 mg of Eu(lll) labelled with J52+154J3U was taken in a 150 ml separating funnel. 10 ml of 0.1$ solution of alizarin in methanol was added and lh«- pll was adjusted to 4.5 with the help of 1M NaOH. The AR - 05.1 volume of the aqueous phase waa made to 20 ml with distilled water and the mixture was equilibrated with 15 ml of n-octanol for 2« 0 min. The phases were al'owed to separate and 2-ml aliquot of each phase was withdrawn for counting on a gamma-ray spectrometer at the channel number corresponding to the 0.34 MeV photopeak of 152EU. The extraction coefficient value was calculated in the usual way. III. RESULTS AND DISCUSSION The extraction coefficient value(E) for the extraction of Cr(lII) with alizarin indicated that maximum extraction occured at a pH of 8.0 with an equilibration time of 5.0 min. The $> E was found to be better than 98$ over the pH ran^e 7.5 to 8.5 and decreased in more acidic and alkaline ranges- Determination of the effect of time of equilibration showed that the f& £ was better than 98?5 over the equilibration time of 5.0 to 9.0 min. The extraction of Cr(III) into different organic solvents using alizarin was studied and it was observed that of the solvent used, cyclohexane was the best solvent for extraction. The extraction coefficient value(E) for the extraction of Eu(III) with alizarin indicated that the maximum extraction occured at the pH of 4.5 with a reproducibility of 9.09$. The E value was found to decrease in more alkaline and acidic ranges. After equilibration, no phase separation was observed for the pH ^ 7. Determination of time of equilibration showed that $> E was better than 96.67$ over the equilibration time of 1.0 to 5. o min. The extraction of Eu(lII) into different organic solvents using alizarin was evaluated and it was observed that n-octatiol was the best solvent for the extraction. The effect of sodium, potassium or ammonium aaits on the extraction coefficient value of Cr(lII) and Eu(IIl; was deter- mined. The separation factor for Large number of elements In th« extraction of CriIII) and Eu(IIIJ was determined and it waa found to be better than 102 for most of the elements. The stoichiometry of metal to reagent was determined by the method of substoichiometric extraction and slope ratio method and was found to be 1:3 for both the systems, IV. REFERENCE i. Organic Analytical Reagents. W. Pranic and J. Welcher,, 4, 409 (1957). AR - 05.2 SOLVENT EXTRACTION AND RADIOCHEMICAL SEPARATION 0? Mo(VI) AND W(VI) WITH OKGAJ'IC DYES INTO NITROBENZENE. D. S. Bhaiia, S. Blnriam & Z. R. Ture]. Nuclear Cfteni r.; try Division, The Institute of Science, Bombay- 400 032, Indie. SUMMARY Rhodamine B end methylene blue have been used for the extra- ction and radiocheiucol seDnmtion gf_ Mo(Vl) and W(/l) respecti- vely into nitrobenzene. 99M<> arid 1°?W were used as tracers. (Key Words i Solvent extraction, MofVI), W(VI)f uhodomine B and methylene VLue) I. INTRODUCTION Literature review shows that vftodaraine B has been used for the detectioti of molybdenum-1- and m^thylene blue has been used for molybdenum and technetium^. However, the extraction of Mo(VI) with rhodaraine B and W(VI) with methylene blue have not been reported in the literature. The present work deals vrtth the extraction and radiochemical separation of Mo(VI) and W(VT) with organic dyeo. II. EXPERIMENTAL i) Extraction procedure for Mo(VI) ID a 150 ml aeparatory funnel 1.009 mg of Mo(VI) labelled with ^Mo nas token, and the volume was made to 16 ml with water after adjusting the pH of the solution with HCl or NaOH. 1 fill of 1# rhodamine D into nitrobenzene was added. The aqueous phasa was equilibrated for 3 min. with 15 ml of nitrobenzene. The phases were allowed to separate arid a 2-ml aliquot of each phase was taken in a counting dish, evaporated to dryness and counted on a G. II. coitrtter employing an Al-absorber of ^^.t^-^/o^ thick- ness to cut off the soft conversion electrons of ^mlc. The extraction coefficient value was calculated in the usual way. ii) Extraction procedure for W(VIj A L-ral solution of W(VIJ eonhnining 1.0 m^ of W(Vl) labelled with ^5w wag taken in a oeparatory funnel and the volume was made to 15 ml with distilled water after adjusting the pH of the solution with IIC1 or NaOU. The aqueous phase waa equilibrated for 3 inin. with 13 ml of nitrobenzene containing l£ methylene blue. The phases were allowed to separate. A 2-ml aliquot of AR - 06.1 each phase was taken in a counting dish., evaporated to drynees and counted on a G.H. counter. The extraction coefficient value was calculated in the usual way. III. RESULTS AND DISCUSSION The extraction coefficient value of Mo(VI) indicated that the maximum extinction occured at a pH of 5*5 for an equilibra- tion time of 3 min. The effect of various solvents on the E value was studied and it was observed that nitrobenzene was the best solvent for the extraction. The extraction coefficient value of WlVI) Indicated that the maximum extraction occured at a ).-'. 3-0 for an equilibration time of 3 min. The effect of different solvents on the E value of W(VI) using methylene blue into different solvent was studied and it was observed that nitrobenzene was the best solvent. The effect of sodiura, potassium or ammonium saltB on the extmoticn coefficient value of Mo{VI) and W(VI) was determined. The separation factor for a large number of elements in the extraction of Mo(VI) and Yi(VI) was evaluated and it was found to be bet her than 10-> for most of the element. The ntoichiometry of m*fcu.l to reagent was determined by t'hs method of substoichiometric extraction and slope ratio method and was found to be 1:2 for both the systems The stoichionietry was further supported by the analysis of solid complex. Decontamination factor at substoichiometric amounts was evaluated and it was observed that of the "JO elements studied only Au(lII) and Pt(Il) infcerferred in the extraction of W(VI). Phe process was made more selective by the U36 of masking rgenta. Decontamination factor for Mo('/I) revealed that Hg(II), ff(VI) and Au(lII) interfered. The interference was removed by masking with suitable masking agents. W(VI) was back-extracted quantifcnrtively with 1:1 NH3 to the extnnt of 92.656 and Mo(VIJ with 1:9 NH3 to the extent of 8l.04< in ;i gingle extraction. IV. REFERENCES 1. Organic Analytical Reagents, W. Pranilc and J. Weicher, 4, 549 U957). 2. Ibidp 4, 517 AR - 06.2 SOLVENT EXTRACTION OP Tl(lII) AND Sn(II) WITH 1,2,3-BENZO- TRIAZOLE INTO ORGANIC SOLVENTS. Jyotsna ITapaclia, V.B. Patil and Z.R. Turel. Nuclear Chemistry Division, The Institute of Science, Rombay- 4 00 032, India.. SUMMARY Aqueous 1, 2,3-benzotriazolo (1,2, 3-HBT) >:as been used for the extraction of TI(III) and Sn(ll) into n-heptanol and cyclohexnnone respectively. (Key Words : Solvent extraction, Tl(lII), Sn(II), 1,2,3-HBT) I. INTRODUCTION 1,2,3-HBT ha3 been used for the solvent extraction of a large number of metal ions such as Cd(ll) , Ag(l)^, Zn(ll)3 etc. However, the extraction of Tl(lll) and Sn(ll) using this reagent has not been reported in literature. The present work deals with developing a rapid and selective method for the extraction of TI(III) and Sn(ll) with this reagent. 204Ti ^a J-Ugn were used as tracers for studying the extraction process. II. EXPERIMENTAL i. Extraction procedure for Tl(lII) In a 150 ml separating funnel w;is taken 1.79 mg of Tl(III) d 4 labelled with Tl. 10 ml of 1# aqueous solution of lt2,3-H3T was added. The volume of the solution was made to 15 ml after adjusting the pH of the solution to 8. 5 with anunonia. The mixture was equilibrated for 2 min. with 15 ml of n-heptanol. The phases were allowed to separate and a 2-ml aliquot of ecjcli phase was taken in a planchet. It was evaporated to dryness in a planchet and was taken for counting on a GM counter. The extraction coefficient value was calculated in the usual way. ii. Extraction procedure for Sn(II) A 1-ml aliquot of the solution containing 2.0 rag of t) labelled with -^Sn was taken in a separating funnel. 10 ml of J-# aqueous :.;otution of 1,2,3-HBT was added. The raolarity of the solution was afijuateii t" 3& with HCl and the voLurae was made to 27 ml with distilled water. The aqueous phase was equilibrate for 5 ain. with 34 nil cyclohexonone. The phases were allowed to separate. A 2-ml aliquot of each phase was taken for counting on a r-ray spectrometer at the channel corresponding to o.39 photopeaJc of J-1JSn. The extraction coefficient value was calculated in the usual way. AR - 07.1 III. RESULTS AND DISCUSSION Tl(III) was extracted to the extent of greater than 99^ at a pH of 8.5 and equilibration time of 2 min. The extraction of Ti(IIi) into different organic solvents revealed that n-heptanol was the best solvent. The eifect of sodium, potassium or amnonium salts of various anions on the E value of Titlll) revealed that 100 mg each of perclilorate, nitrate, sulphate, broraate, bromide, nitrite and chloride, 50 mg each of chroraate, dichromate and acetate do not decrease the E value of Tl(lII)#,gulphite, thiourea, cyanide, citrate, thiocyanate, iodide, thiosulphate and fluoride interfere. The interference of these anion3 was removed by expelling them with pa regia and HC1 prior to the extraction of' TI(ITI). ?J:e interference of various cations was studied. Na(l), Cs(l), S(VI), Ir(IV). RbU), Calll), Sb(lII), P(V), AsUll), Ba(IT), Sb(V), Se(lV), K(I), Ce(IV), Cr(VI), Pt(IV) and WIVI) were extracted to the extent of less than 5$. Ru(III|, Pd(lIJ, Au(lII), Zn(ll), Felll), Hg(II), Eu(iri), Mn(ll), Cd(Il), Cu(II), Co^ll), Pe(lll> and Ir(III) were extracted to the extent of more than 75/6. These cations were removed by extrapting them with 1,2,3-HBT under the experimental conditions mentioned above, maintaining Tl in (I) oxidation state. After removal of these cations, Tl(l) way oxidised to Tl(lll) and was extracted an mentioned above. The percentage extraction of Sn(ll) with i,2t3-HBT into cvclohexanone was found to be better than 98^ over the HC1 moiarity 3.0 to 5.0. A mcximum E value of 70 was obtained at the moiarity of 3M and equixxbration time of 5 min. The extra*- ction of SnlXI) with 1,2,3-HBT into different organic solvents was studied and among these solvents, cyelohexanone was found to be the best solvent for the extraction. Separation factor for other elements was determined. IV. REFERENCES 1. S. Subramanian, Z.R. Turel, J. Radioanal. Nucl. Chem. Lett., 103, 365 U9»6). 2. I. A. Mendes, Z.R. Turel, J. Radic-anal. Kucl. C}!<*mt luett.^ 96, 343 U985). 3. J. Kapadia, Z.R. Turel, J. Radioanal. Nucl. Chem. Lett., 118, 15 (1937). AR - 07.2 EXTRACTION MECHANISM OF Tr(VII) IN TBP-HDKHP-MONOHASI C ACID SYSTEM E . A . P . S . Sasl r y , Sujatha.S, G.K.Mishra and K . N . S i nqh , Nuclear and Rad i ochemi BI ry Lahni tury, Department: of Chemistry, Ha^aras Hindu University, Varanasi 221 005, Fndia. SUMMARY The disl rihill ion cue f f i <• l cut < Kd ) for the extraction of Tc(VJI) by 30% TBP in n-dodecane and it R 1:1 binary in i x t u re ( v/v ) with HDF.HP in the inpseiice of UNO, , MCI , HClo and H 2^\ have been determined. R.IMCII on I he Mt o i < -h i omet r i c and HII >pe ana I y a I B methods, a possible mechan i sin has hcen proponed. Key words : Kx I t act i on /Tr ( V I I ) /T»l' IIDFMP /Horn >bas i<- a.-ids T . INTRODUCTION Although the extraction of Tc(V I I ) has been reported by several workers! I ,5 I but the possihIe mechanIsm has not been discussed in detail. The pi esent work wan unde r t aken with a view to study the ext ract ion of Tc (VII) by (l:l) binary mi xt.ure(v/v) of TBP-IIDEHP in n-dodec ine and to propose a possible me cha n i sin. T T . EXPERIMENTAL fiqiirf ! volumes of organic ami .KJIII-OIIH phases were laken in each experiment . The HIJNHHIH [ih.ise I'IIIIHIKIHII of NriTcIL SO I lit l on in mo no ha sic acid and the oi(jaiii<- phase consist ed of .10% TBP in n-dodecane in litsl set ami (1:1) mixluie (v/v) of TBP—HDEHP in another set of experimentK. Olher experimental details are similar to that of l.ieser et. a I . I 7 J III. RESULTS AND DISCUSSION The distribution coefficient values for the extract ion of Tc(VII) by TBP and TBP-HDFHP systems from aqueous solutions containing different mono basic Hinds are given in table 1. The results show that the Kd values increase with the concentration of the acid initially, attain a maximum, and thereafter start' decreasing with further increase in acid concent rat ion. The decrease in the Kd values at higher acid concentration is because of the loss of free chelal iny acl iv;ty of the ext ractant due to the formation of [acid • extractant ) adduct. This results in the step wise decrease in the font or three fold coord i rial: ion of Tc(VII) with extractant leading to partial solubility of the adduct already formed at the opl i mum acid coin ;MII rat ion. The extraction of Tc(VlI) is found to be enhanced by the acids in the following order, H2SO > HCl > HNO3 > HCI()4 The slope of I he plot of Ioy Kd va log [extra ctant ) was found to be 4 in (tie absence of acid, whereas this value is found to be < in presence of acid up to a concentration corresponding to the maximum Kd value, (a) In aqueous solution : [HAJ ^ -* IH^I^ * I A" laq ( i ) AR - 08.1 HNa i + [ToC 1 ( ii ) [HToO, J U + [TcO+ ]-_ (iii) (b) Extraction by TBP : + l ( iv) Tin the absence of acid) + [H lo + I A" i i•—•— [HTcO4 .3Sjlo>3 + I HA (v) ^ "• (in presence of acid °up ttoo optimum concentration) (c) At high concentration of acid [HTcO . 3Si 1^ + [H"M«^f [&- ) ^=± ~ *"*• — i "J_ ru+i _i_ I A • 1 HTcC^ .Sj] + [HA.Sj-.-.c, (vii) Q.1 I lu+l *x I a 1 ' I UTr-Tf:On, L t THA-Silo^ " (viii) (d) Tn mixture of extractants : ^ . 3Si .S2JO (x) (Si = ^fBP, S£= HDEHP) O>^ " The adduct (x? fol lows l.de KHICI^ m ism H• hIgher concent.rat ions of HCK! ris Kt.ated nlKivf. I Extraction of Tc(VlI) from acidic medium Dist.r l but ion Coefficient. ( Kd ', HA ; HNO 5 HCl HCIO^ H^S % 0 i; 0.78 0.99 0.56 0.76 0.42 0.85 ;o.63 ; o.89 0 5;1.14 1 .52 0.91 1 .33 0.93 1.29 ;o.86 ; l.23 l o;1.27 1.79 1.21 1 .71 1 .03 1 .62 ; i.10 ; l.51 I 5; 1.13 1.88 1 .31 1 .9 3 0.89 1 .73 ; i.30 ; l.72 2.o;0.93 1 .60 1.14 1 .71 0.67 1 .48 ; l.48 ; l.85 3.o; 0.49 0.67 0.46 0. 93 0.16 0.71 ; i.65 ; 2.01 IV. REFERENCES III J.V. Hol-der, Radiochim. Acta. 25, 171 (1977) 121 A.S. Kertes and A. Beck, Proc. 7th Tnt. Cong. Clin. Chem. Stockholm, 35 (1962) 131 T.H. Siddall, U.S. Atomic- Energy Coramun. document, 364 (1959) 141 K. Alcock, S.S. Grimley, T.V. Healy, J. Kennedy, H.A.C. Mckay, Trans. Faraday Soc. 52, 39 (1956) 151 K.H. Lieser and R.N. Singh, Kad i och i m. Add, VI, 20 A (1983) 161 Ka/.uyilki HoHh l mot.o, Takrishi Oinori Hiui Kenji YIIHII Radiochim. Acta. 49, 65 (1990) 111 K.H. I.ieser, A. Kriiger and H.N. Singh, Radiochim. Acfa. 2H, 47 (1981) AR - 08.2 SYNERGISM IN THE EXTRACTION OF Zr(IV) BY DIFFERENT SOLVENTS FROM VARIOUS MINERAL ACID SOLUTIONS .Sujatha.S, G.K.Mishra and R.N.Singh Nuclear and Radiochemistry Laboratory Department of Chemistry, Banaras Hindu University, Varanasi, 221 005, India. SUMMARY The distribution ratios for the extraction of Zr(IV) by TBP and its binary mi cures with Py or HDEHP in n-dodecane in presence of HNO5 ,HCljHClO^ and HjSO^ have been determined. The results show an enhanced extraction in the order TBP-HDEHP > TBP-Py > TBP Keywords: Gynergisir/extract lon/Zr < IV ) /mi nera 1 acids/ I. INTRODUCTION The extraction of Zr(IV) by several extractants h3S been a subject of many studies!1,4]. However, very little is known about the phenomenon of synergisrn or antagonism related to Zr(IV) species. The present work was undertaken to study the solvent extraction of Zr(IV) by 30% TBP and its binary mixtures with Py and HDEHP in n-dodecane. II. EXPERIMENTAL Iri the experimental work the organic phase consisted of either 30% TBP or its 1:1 binary mixture (v/v) with Py or HOEHP in n-dodecane were taken. Other exper iment.Hl details fire namilar to that of Liese - et:. al. [5J. 'Ar-9ri tra :er ws obtained from BARC, Bombay, India. The radio activity was merfjurcd by a Scalar rate meter(SR-7) or miiJti channel analyser (PHA-1) attached to a well type Nal(Tl) probe. III. RESULTS AND DISCUSSION The results given in table 1 show the solvent extraction of Zr(IV) from aqueous solutions containiny HNO^,HC1,HClO^. and HfcSO^ by mixtures of extractants. It is obvious that there is an increase in Kd values when 15% TBP is mixed with Py and HJDEHP in case of all the above mentioned acids. Thus a binary mixture of TBP and other extractants extracts Zr(IV) more than TBP or any of the extractants alone selected for the work. The maximum enhancement in the extraction of Zr(IV) was observed in the binary mixture of TBP-HDEHP whereas lower Kd values were obtained in case of TBP alone. The overall order in the enhanced extraction of Zr(IV) by various extractanta corresponds to the following order, TBP-HDEHP > TRP-Py >TBP. This order of influence may be ascribed to the presence of more active hetero atoms constitut ing*more efficient functional groups which are responsible for the difference in their solvating power and also due to the difference in structure and Length of C - chain. The experimentally higlv|f-»r Kd vdlueH in cane of TBP- AR - 09.1 HDEHP syst em may be ,itt ribut-ed to HDEHP HIOIPCUIP which b.-'have.s like a bident riip ligand in nridic medium due to lt.H d imer isationi& ! . Thf ext rncl ion of Z r ( I V ) WHH nlmn rv«i lo he influenced by the acids in I he following order: UNO, \\r lo. TAHI.K I Synergistie effect, on the solvent exlracl ion of Zri IV) by mixtures of extract.anta from aqueous Holul urns conl riininq different mineral acids. ; Ext. rant-ant TRP TUP - Py TRP - HOEHP :HNO3 1 fri 2 M 1 . 5 M i Kd ma x. 0 . 70 I . l 1 .48 ; HCI 1 M 2 M M ! Kd max . 0.74 I .29 :M HCIO4 1 . S M 1 M !Kd max. 0 . I lf> 1 .24 :M H2SO4 .? M i M I Kd max. I .4C I . 1 .HO TV. REFERENCES / 1 / P. K . Misilirrt, V . Clwkr.ivin I I y, K . (" . Dasli , N . l( - I)H T HIIH S . N . Bhrtt.l.acha rya , Knd i ucli i in . Ada, 4'J, 4 ri (14411) I'll K.W. Murbarh and W.M. McVey, I,HI, - I I c) , l.awerence I, i ve t mo r-e Laboratory, Apr i I I 4r>4 I'M J. Malvyn Mckibben, Had l oc:h I m. Acid. AR - 09.2 LIQUID-LIQUID EXTRACTION OF Mo(VI) & U(VI) BY LIX 622 R.N. Mohanty, S. Singh, V. Chakravortty & K.C. Dash Department of Chemistry, Utkal University, Bhubaneswar, India. SUMMARY — The order of extraction of Mo(VI) from 1 M acid solutions by 5% (v/v) LIX 622 (HL) in benzene is HC1 > iiNO, > HC10. > H^SO., and the extracted species is shown to be Moo?L. as established by IR data of the organic extracts and the extrac- ted species in the solid form. Extraction is' almost quantitative at end above 10% (v/v) LIX 622, and independent of concn. of Mo(VI) in the range 10~4 to 10"3 M . Extraction of U(VI) by 10% (v/v) LIX 622 in benzene becomes quantitative at equilibrium pH 5.9. Separation of Mo(VI) and U(VI) is feasible. (Keywords: Extraction, Mo(VI), LIX 622, U(VI)) I. INTRODUCTION — The presence of molybdenum in the acid or alkaline leach solutions, even in trace amounts, may lead to its contamination in the uranium end-product and pose a serious problem, for which solvent extraction process is employed for removal of Mo(VI) from uranium circuits. A commercial oxime- based chelating reagent LIX 84 (2-bydroxy-5-nony.]acetophenone oxime in kerosene) /l/ has been used ^s e.n extrectant for Th(IV), U(VI) & Mo(VI) in this laboratory. Freiser & coworl'^rs /2/ investigated the equilibrium and klm-tics of extraction of Mo(VI) with LIX 63 (5,8-diethyl—7-hydroxy-dodecat) II. EXPERIMENTAL — Commercial J./1.X 622, obtained from M/s lienkel Corp., USA, way usi-.-cl in benisi-n« a3 diluent. Tributyl phosphate, TBP (BDH), was purified, isoamyl alcohol and other organic solvents were redistilled before use. The procedures for extraction, method of estimations of Mo(VI), & U(VI) and of determination of distribution ratio, U, are the same as already reported /l/. The IR spectra of the samples were recorded on Perkin Elmer 983 IR spectrophotoiueter fitted with IK data station at IIT, Madras (India) using polyethylene films. III. RESULTS AND DISCUSSION — Solvent extraction of molybdenum(VI) was carried out from different acid solutions by 5% (v/v) LIX 622 in benzene. The concentraLion of Mo(Vl) was maintained at £ 10 M , since at higher concentration polymeri- zation sets in. The order of extraction of Mo(VI) from 1 M acid solutions is HC1 > HNO3 > HC1O, > H SO.. Extraction is found to AR - 10.1 decrease with increase in concn. of HC1 (1-0.4 M) and H SO. (1-9 H), and increases slightly with increase in conca. of UNO, (1-6 M, at higher concentrations the extractant oecompose.s), and of HC1O (1-8 M). The 1R spectra of the organic extracts and the solid extracted species show the symmetric and asynime-tric Mo = 0 stretching vibrations in the: 850-950 cm"' region. The extracted species exhibit IK bands characteristic of cls-Hr>0 and th': oxinie ligands. In case of LIX 84 and LIX 622 (in benzene) and the organic extracts, the v(C=N) appears at 1610-1615 cm"' and v(N-o) is at 1030 cm~V5/. In the organic extract corresponding to the extraction front aq. J-J SO., no IR bands were observed for the sulphate group. In order to explain slope-analyses and t hie natvire of the extracted species, MoO-L«, the following equilibrium has been proposed + h\ooi , , + 2 HL, , = Mo0oL_, , + 2lfj~ , (]) 2(aq) (ory) 2 2(org) (aq) Extraction from 1 M HC1 solution increases with increasing percentage of LIX 622 i ri benzene, is almost quantitative at \0"A (v/v) L1X 622 and remain:; constant thereafter. Influence of diluents such as benzene, CC1., chlorofcorni, hexanol , oyclobexariol, isoaiuyl alcohol and nitromethnne on the extraction of molybdenum (VI) from 1 M JJC1 solution by 2% (v/v) LIX 622 has been studied. Liquid-liquid exUaction of uranium (VI) (t> X 10 M) trom nitrate media by 10% (v/v) LIX 622 in benzene has been studied over the equilibrium pH range 3.0 to 6.0 (the pH was varied by addition of pyridine). Extraction increases with increase in pH and becomes quantitative at pH 5.9. The effect of variation of percentage of LIX 622 and TBP in their mixture on the extraction of uranium(VI) bus been studied. TBP can Oe used as a synergist upto 2% (v/v) after which) it acts as an antagonist. The following mechanism for extraction has been proposed. UO^t ^ + NO~ v + HL, > = U00(NO_,)L, . + nt > (2) 2(aq) 3(a ;1i-,s M.N. OmtK: 1 k.ar and A.N_.6arq .* ;>« [jii.costry .N.igpur ijiii ver sily, Nfv >'UH 44OO1O SUMMARY A r ad ! < i. h •.->^ hcen fpveiopf^ fcr the* i.'etei ihinji-. ic.n oi '>vJT;. ) •ie-!'iq '.> tra.'fr. ; i if- hisf! • llw compiexat inn t. f ' •' ( i • >. ) v. i t • • H liyiir oxyqu i' '•-> > ' np ( u •< i > it? ) a:iri acelylareldne .*-M1 putt !Ct.;r'.; ..'i rhliinifnnn. Er.tfect ci f -.•."_!« param*?'.ipr s 'in h .is f!.,l > m.-j ii) pqui 1 I'lrat-jnii, fii!'.11"' o * 5ol vei 11 . qu.it ' ; t .> t i •" u.tt'ire, ••":'•<.• '. lit it:, vprr.-"' ionr". !•••" 'n>f?.-i •-; t;;.1 i ^d - ! ( Keywords. !'n.' I . IN' Hi.'! • IC M ;jr; Chr o.r.i" solvent pytr.n'ti'.ii [.i ••'...fiiiir t- '• .'r I;; (III) using O tracer. 1 I .fe'XPEN > MEN: At. All :hp.nuil5 »<.pd were A!y dissulvji.ij Ci ;)•* ; t.-i -ick) MI mi<)iin ace t / i «•"' . .m- r •• .'i •;. I :' rttj.i : ?i titr.il t ed with Hm! chloroform. Hot f • or yai i ' i. pf:,.• •.• ..•>:•.•;•• » <.•.•' . <., .i , (U:- t *••- : 'I' :i.:'i • ' I . ! 1 I .,:;; "*-i ". >:• (•"•!-:> P' .>•'! 'V ii; UHIIT 'O develop the r ad iuc hemi La 1 p.- or.ei'.i; • -'. A l< - 11.1 concentration range 5-2OO fjg (Fig.2). c..Effect of InterferrtnQ ions— The possible radiometric and chemical interferences were evaluated by extracting >-— 2OO ^g Cr (tracer + carrier) along with *-»« 1OO /jg of different interferring ions or their respective radiotracers. Hn, Ba, Cs, Cd do not interfere at all. However, Hg, Sb, Co, Ag and Se interfere to the extent of 2O7.. d.Precision and accuracy: Replicate measurements were made in the concentration range 100-300/jg of Cr (Table 1). A set of 10 replicate analyses were performed by taking 2OOpg whence actual concentration was found to be 199.9£O.2/LKJ- 11 was observed that results are within <3X error and R.S.D. ACKNOWLEDGEMENTS; Our grateful thanks are due to IAEA, Vienna for financial assistance under contract No.5649/RB. IV.REFERENCES: 1.G.V.Iyengar, Bei.Total Environ., 86= 69 (1989). 2-E.Berman, Tnxic Metals and TJieir Analyses, Heyden and Sons Ltd. London, (198O), p.75. 3.V.M.Rao and M.N.Sast.i, Talanta, 27, 771 (lVBO). 4.R.R.Greenberg and R.G.Zeislwr, J.Radioanai.Nucl.Chem..124. 5 (1988). 100 - .. J I _. 1 50 100 '50 200 ccr(ino,py - FIG. 2 AK - 11.2 REMOVAL OF Cr(VI) FROM AQUKOUS SOLUTIONS BY ZrOi : A KINETIC BASED STUDY M.M.BHUTANI. RAMESH KUMARI, A.K. KITRA Department of Chemistry Indian Institute of Technology,Hauz Khas.New Delhi-110016 (India) SUMMARY: The adsorption of chromate ions on zirconium dioxide from aqueous solutions has been studied as a function of pH and chromate ion concentration. The system obeys the Freundlich adsorption isotherm and the mechanism of sorption is discussed in the light of ligand exchange model. KEYWORDS: Sorption, ZrO2, Cr(VI), Freundlich isotherm I.INTRODUCTION: In recent years, the use of chromium and its compoundn has Increased considerably in industry and agriculture. The discharge of chromium in industrial effluents and wastes into waterways has become environmental hazard*. Zirconium dioxide, one of the most promising ion exchangers, behaves as an anion exchanger in acidic and neutral solutions and cation exchanger in alkaline media. ZrO2 has recently become important in water desalination by hyperfiitration2 process. The present communication reports the uptake of chromate ions by ZrO2 under various experimental conditions. II.EXPERIMENTAL: All chemicals used were of the analytical grade. Solutions of the required concentrations were obtained from a stock solution of 0.1M sodium chromate by successive dilutions. Cr-51 in the form of Na2 51CrC« in HC1 was obtained from BARC, Trombay, Bombay (India). 0.32 gamma activity was measured on a single channel analyzer in conjunction with Nal(Tl) wel] type detector from Nuclear Enterprises Ltd. U.K. III. RESULTS AND DISCUSSION: Preliminary investigations were conducted to study the influence of pH on the sorption of CrO42~ ions on ZrO2(Fig.l). The sorptlon increases as pH increases and reaches a maximum value at pH 6.75 and thereafter it decreases. The hydroxy group of the adsorbent enters dissociation or association reaction depending upon the pH of the medium as fallows: MOH ?=* M* + OH- or M-OH ^=i M-OH2* (1) 0H- MOH ?==* M0-+ H+ or M-OH ^=* M-0" + H2O (2) The reaction (1) proceeds at low pH value where adsorption of chromate ions is favoured while reaction (2) is conditioned by high pH value favouring the cation adsorption or lew adsorption of anions. The isoelectric point (IEP) of ZrOi is 6.73. it appears that at lower pH the charge on the tracer and at higher pH the charge on the surface plays a decisive role in deciding the mode of adsorption. The kinetics of adsorption of chromate ions on ZrO2 was examined as a function of chromate ion concentration (10-6-10-2M) in aqueous solution at the room temperature. The general nature of time rate variation curves as well as equilibrium time, remain AR - 12.1 essentially similar in the concentration range studied but percental of adsorption increases with dilution due to avajlability of small number of adsorptive species. The dependence of adsorption on the adsorptive concentration at equilibrium was tested through Freundlich isotherm (Fig.2). The linear Hot of log ae vs log Ce where ae is the amount adsorbed at equilibrium and ce is equilibrium bulk concentration, suggests that th« sorption follows Freundlich isotherm over the entire range of adsorptive concentration. The fractional value of 1/n - C 854 and relatively high value of k=3.98x10-8 signifies that the oxide surface possesses stronger adsorption forces and higher affinity for chromate ions and distribution of energy sites are exponential in nature*. IV. REFEItTCNCES 1. S. Music, M. Ristic and M. Tonkovic, Z. Wasser- Abwasser-Forsch, 19. 186 (1S87). 2 C.B. Amphlett, L.A. McDonald and H.J. Redman, J. Inorg. and Nucl. Chem.. 6, 236 (1958). 3 G A Parks, Chemical Reviews, 39, 177 (1965). 4* R Sips, J. Cb^m. Phys., 16, 490 (1948). Fig. 1 CARRYING Vs pH 100 80 Fig. 2 FREUNDLICH ISOTHERM o 1UJ 60 CARRlER-ZrO2 u 8 o 01 o 20 O 2rO2 i 0.1 50g 7* 2 j i I |_ 0 2 U 6 8 10 0 pH -LOG(Ceq) AR - 12.2 ELECTROLYTE-DIFFUSION OF Cof HO_ ) IN AGAR OFL MFf>I: .1M : CONCENrRATION DEPENDENCE, OBSTRUCT I ON EFFECT AND ACT! VA I I ON E S.F.Patil, N.S.Rajurkar arid A . V. Borhnde Department of Chemistry .University of Poona. Pune 411 0€>7< INDIA) SUMrlAflY: The deviationa between the measured diffusion coefficients of cotjalt nitrate in IX agar gel over concentration range of 10 -0.25M and theoretically calculated valuer? are explained by considering different interactions occurring In tho water /!el- electrolyte system. The effect of concentration on the obstruct- ion effect has been explored and the decrease in the latter with former is interpreted in terms of competitive hydration between ions and agar molecules. The energy of activation for diffusion of cobalt nitrate in 1% agar gel at €>. 1M concentration is also determined arid is found to be 31.5 ± 0.2 kJ mole {Key words: Electrolyte-diffusion, Activation energy. Concentrat- ion dependence, Obstruction effect). I. INTRODUCTION: In previous coronnmicatonti ' , we have i eported the various aspects of electrolyte-dif fuciion of Co£>04and The present paper deals with the measurement of diffusion coefficients of Co(NOj) at various electrolyte and gel concentra- tions which enabled us to examine the applicability of OrLsager- FUoss ' equation. I H) I V<|. l.illDII Of I', ,,((|Oj)? Will. V|'i.lH- I cm I of i.UftLGril. ol ion ol cubolt AR - 13.2 MULTIELEMENTAL NON-DESTRUCTIVE NEUTRON ACTIVATION ANALYSIS OF THE CANCEROUS TISSUES OF THE HUMAN BRAIN EMPLOYING "^C NEUTRON SOURCE B.S.Shanbhag, Z.R.Turel, Nuclear Chemistry Division, The Institute of Science, Bombay-4U0tJ3i!, INDIA. AND K.E.Ture-1, Seurosutgery Department, J . J . Hospital, Bombay-400003,INDIA. SUMMARY Mn,Na and K have been determined in the cancerous tissues of the human brain employing INAA. (Key words : INAA, cancer, brain, 56 Mn, 24Na, 42R) I.INTRODUCTION Instrumental neutroj activation analysis invo- lving f ray spectronietry with a high resolution detector is a time waving method and could be used effectively for the determination of elements present in the biological samples. The present work describes the determination of Mn, Na and K in the cancerous tissues of the human brain involving neutr- on irradiation followed by f-ray spectrometry. II.EXPERIMENTAL i)Sampling Samples of normal and cancerous tissues were collectod in pretre"ted vials directly after the operation. The samples were lyophilised and ground in an agate mortar and pestle. The age of patients varied from 11 to 61 years and the type of tumours analysed were lymphoma, glioma, meningioaia etc. The presurgery medication, the case-history etc. were preserved. ii)'icirget preparation 50-100 mgs of dried, lyophilised powder of the cancerous tissues of the brain was taken in quartz ampoules. Standards were prepared by evaporating the solution of Mn(II), Na(I) and K(I) in ijuartz ampoules. The samples and AR - 14,1 standards wore sealed, placed in polyethene bag and irradi- ated with thermal neutrons. iii)Irradiation The irradiation of the sample and the standard was carried out in position 2 of the 252cf neutron source having a thermal neutron flux of 8b x 107 n.cm-2s-l The flux of the source was calibrated employing Au-foil by manual as well as pneumatic transfer method. i v ) P r a c e d u r e for the determination of Mn , the sample and the standard were irradiated to saturation activity for a dura- tion of 15 hours. The sample and standard were counted after a lapse of 5 mins from the end of irradiation at the 0.84 MeV photopeak of 56nn. The sample and standard were remeasured alternately with a time interval of 30 mins. After the complete decay of 5fc>Mn,the sample and standard were reirradia ted at same position for a period of 3 days followed by counting of the sample after 5 mins from the end of irradiation The activity of ^*Na produced by neutron irradiation was measured at the 1.37 MeV photope- ak and that of ^2K at the L-53 MeV photopeak. Peak area cou- nting was employed for both the isotopes. The sample and the standard were counted under identical conditions at an inte- rval of 3 hours. The amount of Mn, Na and K present in the samples was calculated from the weight of the standard and the activity induced in the tissue sample and standard at the respective photopeak. III.RESULTS AND DISCUSSION Irradiation ojf tissue sample produced ^o Hn by the(n,f)reaction on 55 Mn , '-^Na by t h AR - 14.2 STUDIES ON THE HETEROGEMITY OF MICRORETICULAR CARBOXYL.ATE ION EXCHANGERS - II G.S.Murty, Nuclear Chemistry section. School of chemistry Andhra University, Visakapatnam 530 003, INDIA and P.K.Padmanabhan, Analytical Chemistry Division, BARC, Bombay 400 085, INDIA. Summary Earlier studies revealed some unusual sorption of multivalent cadons when the exchange studies were carried out at micro concentrations which was attributed to the heterogenity of the exchange sites in these exchangers. This observation was further (.•'...:v'ii -d by extending these studies to other carboxylate e^ciiaij. '• c:: iiKe lndion-2Jb, BioRex-70 and CS-101. I Key words : carooxylate exchangers, heterogenity, metal ions. } INTRODUCTION While studying the preconcentration and separation of various elements like CaillWl/, cu i II) , Ni i 11) /2/ , In.{Ill) , Tl {III) etc/3/ from different matrices it was observed that these elements were tenaciously retailed on Zeokarta-226(H+), at trace concentrations. The element so tenaciously sorbed was found to be eluted under drastic conditions i.e. at higher acid concentrations and with larger volumes/4/. Our earlier investigations/5/ on carboxylate exchangers like lkc-50, Merck-IV and Zeokarb-226 revealed similar behaviour with Cu(II), Zn(II) and Y(III). These studies wi.-re extended to other exchangers like Indion-236, BoiRex-70 and CS- 101 and the results obtained are presented in this paper. EXPERIMENTAL Ion exchangers : iiidion-i.ib, BioRex-70 and CS-101 in their H + farms. Batch equilibrium studies : The equilibrium distribution (Kd) values were obtained by following the method mentioned earlier/5/. The metal concentration in solution/ resin phase were determined radiometricully in the case of Y(III) and by flame atomic spectrophotoiaetric method in the case of Cu(II) and Zn(Il). RESULTS AMD DISCUSSIONS A systematic study of K^ values were carried out varying the metal concentartions ,from 1000 ug to 10 ug in solution phase. The values obtained are given in table 1. The Kd values are found to incr^'se with the decrease of metal concentration ( in solution ) as observed earlier/5/. This tenacious sorption of AR - 15.1 metal ions at micro concentrations were confirmed by conducting column experiments and following the elution pattern. A probable explanation for this behaviour may be due to differences in the exchange site distribution. In other words the distance between the adjacent exchange sites may be varying. Sherry/6/ pointed out that the selectivity order for ions may differ depending on the inter site spacing. These studies show that there exists a small but significant ( particularly at micro concentrations ) fraction of the resin contains heterogeneous exchange sites which may arise due to che structural heterogenity/7/. in these exchangers. Acknowledgements he authors express their thanks to the Head, Analytical Chemistry Division and Director, Chemical and Isotope group for their keen interest in this ;:ork. REFERENCES 1. P.K. Padrnanabhan et.al., Ind. J. Chem 20. 428 (1981). 2. O.C. Dias, M.Sc Thesis. University of Bombay (1977). 3. S.G. Iyer et.al., Talanta 23, 525 (1976). 4. P.K. Padmanabhan and Ch Venkateswariu, Tech. News Serv. vol 9 (1977) 5. G.S. Murty and P.K. Padmanabhan, Proc. Radiochem. and Radiat. chem. symp., Nagpur (1989). 6. H.S. Sherry, Ion Exchange Vol 2. p 89, Ed. J.A. Marinsky, Marscel Dekker (1969) . 7. V.V. Varentsov and I.M. Abranov J. Chromotog., 102, 83 (1974). Table 1. K d values of metal ions at various initial concentrations on different carboxylate exchangers. Solution phase Exchanger H+ form concentration Indion-236 BioRex-70 Cs-101 10 ug Cu >104 2.4X103 4.1X102 Zn 2X103 210 185 Y 2.1103 1.1X103 520 100 ug Cu 5.2X103 190 235 Zn 60 30 30 Y 2.4X103 8.3X102 530 1000 ug Cu 140 30 30 Zn <10 <10 <10 Y 320 140 110 [ Cu, Y at pH 4,0 and Zn at pH 3.0. ] AR - 15.2 CARRIER FREE SEPARATION OF99Mo FROM FISSION PRODUCTS S.K.DAS. A.G.C.NA1R, S.M.DKSHMUKH AND SATYA PKAKASH Radiochemistry Division, Bhabha Atomic Research o.itre, Bombay 400 085, India. ABSTRACT.... : Mo in carrier free form has been separated from fission products in the neutron induced fission of U. Coprecipitation of Mo with Pd as the oc-benzoin oxime complex followed by extractici" and ion exchange chrcmatography have been used to separate Mo. a-activity and fission product imparities were estimated in the final Mo sample. The uppe; limit of these have been reported. INTRODUCTION: 99mTc is a very unique nucleus used for making radiopharmaceuticals for organ imaging and dynamic function studies because of its suitable half life and gamma line. This nuclide is milked out from Mo- Tc generator system which is technically known as "cows' '. Mo is produced either by neutron activation or by separation from fission products. The latter one gives 9Mo of very high specific activity'^ ( 1000 times more than that produced from activation) . The q q chromatographic technique used in "Tc generators presents inherent problems if y Mo of low specific activity is used because of the low absorption capacity of the alumina bed' '. In the present paper we have demonstrated the separation of Mo from other fission prod, -ta without using any Mo carrier. EXPERIMENTAL: '33U hss been U3ed as the fissile material. Two types of targets were prepared. 150 microgram of U was planchetted on a 1 mil Al foil covered with 0.03 mil Al spacer and a catcher foil. Another target was prepared without any separate catcher foil. The targets were irradiated in CIRUS reactor for 24 hours in a neutron flux of 5xlO12 n/s.cm'2. After irradiation the catcher foil and the target were separately- dissolved in HC1. Ru and I fission products were volatilised off1-3' by oxidising with KBrO^ after adding respective carriers. The solution acidity was adjusted to 0.1 M of HC1 in presence of 5 mg of Pd carrier and a-benzoin cxime was added to coprecipitate Mo with Pd . The complex wa RESULTS AND DISCUSSION: Solid content of the final 5 ml Mo solution was found to be very negligible. The chemical recovery of Mo was of the order of 60%. The decontamination factors achieved for alpha activity in both the samples of molybdenum separated from the catcher foil and the target as a whole were nearly the same. The fluoride content estimated by ion selective electrode potentiometry was of the order of 1-2 ppm. A typical analysis of the fission product impurities is shown in table, a- activity in the separated sample has been found to be rnore than the allowed limit. Further washing of the column with HCi+HF mixture trill reduce the level of cr-contamination * ' . CONCLUSION: A procedure for the separation of carrier free ho from fission products is developed. The final product is found to be free from any solid. Fission product impurities eg Nb, Ru and I are found to be negligible. The level of impurities can be reduced further by washing off the column with larger volume of respective solutions. The determination of pure (5-emitting nuclides and other fission product impurities e.o. Ru, 89'90Sr. 137Cs, 125Sb is in progress. ACKNOWLEDGEMENT: Authors are thankful to Dr. P. R. Natarajan, Head, Radiochemistry Division for his keen interest in the work. REFERENCES: 1. R. E. boyd, Jkadiopharnaceuticals and -abelled compounds. Proc. Symp. organised by IAEA and WHO, v.l, p.3 IAEA, VIENNA, 1973. 2. S. A. Ali and H. J. Ache, Radiociiim. Acta. 41, 65 (1987). 3. R. P. Larsen, L. E. Ross and Gwendohyn Kisser, ANL 5810 4. E. M. Scadden and W. E. Ballau, Radiochemistry of Molybdenum, NAS-NS-3GO9tl^bU) 5. K. A. Kraus, F. Nelson and G. E. Moore, JACS 77,3972(1955) TABLE 99 Contamination levels in fission produced Mo Nuclide Contamination level pCi/Ci of Mo ppm wrt Mo 103Ru 11.1 168 95Nb JO 384 13iI 0.7 2.2 Total a 0.5 AR - 16.2 STUDIES ON THK DECONTAM1 NATION OF FISSION PRODUCED "Mo FROM '"1 USING CHELEX 100 CHELAT1NG ION EXCHANGE RESIN. N.D. Vairiya , P.R. Unni , A.R. Mathakar , M.Subramanian. Isotope Division , Bhabha Atomic Research Centre Trombay , Bombay- 400 085. SUMMARY : Fission y*Mo prepared by the hydrochlorination - sub- limation mechod showed the presence of IJ'I contaminant slightly above the maximum permissible level for its use in nuclear medicine. This paper deals with the ion exchange studies carried out on Chelex 100 (100-200 mesh) in HzSO4 medium for the removal of iodine from molybdenum solution. The results obtained were quite promising. (Key words:Fission products, Molybdenum-99,Iodine-131 Chelex 100) I.INTRODUCTION Nuclear fission provides carrier free s Ci Mo which is very useful in preparing easy to handle *ai"'J.'c generators /I/. A method has been developed for the chemical separation of carrier free "Mo from fission products by hydrochlorination-sublimation technique /2/ . Except for 1Jix , the product coi,formed to che U.S.P. specifications /3/. Inspite of various techniques used for the elimination of '"I before and after hydrochlorination,the product could not conform to the U.S.P. specifications with regard to 1JII contamination. Hence, investigations have been carried out to remove UII from fission product solution in Hz SO4 medium using Chelex 100 resin prior to proceeding with the separation of S9Mo and also at the final purification stage. Chelex J 00 resin is styrene divinyl benzene copolymers containing paired lminodiacetate with chelating groups xn binding polyvalent metal ions /4/ . 11 . EXPERIMENTAL. : Radiotracers used in thm work were obtained from Isotope Division , B.A.R.C. 2.Reaaents : Reagents used are ail of analytical reagents grade. 3. Radiation L>e tector : For the assay of radioactivity the 4K intrinsic Germanium MCA was used. 4 . Procedure : The distribution ratios ( kd) of 'J'' Mo and ' J J I tor Chelex 100 exchanger in 1.5 M H2 SO-i medium at varying concentra- tions of WH4 SCN in the presence of baz Sv-.< were determined as fol- lows. 200 mg. each of the resin was equilibrated with the respective tracers pretreated with Na^soj followed by thiocyanate compiexmg in Hz SO4 , for one hour. Final volume was adjusted to 50.0 ml ; the carrier loading was kept below 10% of the total capacity of the resin. After equilibration , the solution was filtered , a known volume was counted along with the rfterence tracer solution. From the counting data, 'Kd' values were calcu- lated and the most favourable concentration of NhU SCN (u.bM) was selected tor column adsorption studies. A glass column of 1 cm diameter and 18 cm length containing 7 gm of Chelex 100 pretreated with l.b M H2SOi was used tor column studies. The adsorbed molybdenum was eluted with 1 M NaOH. The results of experiments are given in the Table. AR - 17.1 III. KESJJLTS_AND DISCUSSIONS The distribution ratios were obtained over a range of concentra- tions 10.1M to 1MJ of NH-iSf.N in 1.5 M H2 SO* tor both * * Mo and 1 J ' I oa Chelex luO of different mesh sizes. The results showed that appreciable difference in distribution ratio for "Mo and 1J1 I exists at O.b M UHi SON ; the actual values of Kd being A'l and 0.01 respectively. This marked difference in Kd values are utilised here to get a good decontamination of 1; " Mo lrom ' •' ' I. 'I'm s observation is corroborated by the column experiments shown in Table. About yb% of ' •' ' 1 is released into the eltJuent whereas of 'aMo gets adsorbed on the column as shown in table Also, CiieJex 100 of mesh size .100-200 is found to be well suited for this operation because of its high retention tor »-Mo l^b ^%). Mixtures of various compositions of •*''Mo and ' '' 1 loaded on to the column also gave the same results; and hence this ion ex- change procedure can be used for the effective de<"fnr ami na t I • >n of "MO tl O!!i ' ' ' 1 . 1 V . ACKNOWLEDGEMENT The authors acknowledge w 1 t fi thanks the keen interest tak^-n in this wotk bv Shri A. C . hapen . HCH-I , Isoi <^,R in vision. V. L.P.Richards, kadioisotope '., 1 .-11 <, 1 s for c 1 1 11 j c a J us* BN1, I J 4.5b uyoM 2.N.D. Vaidya and T.S. Mm tliy, Unclear Chenustry and kadiocliciinstry Symposium. BHU (198L), pp.bUl. J.United States fh.;,maf upca , XXI , kevismn , U . S . ¥ . C'tivention kockwilie , USA (Ly>i'il , pp. I u I o lull. 4 . Samuh . A. Ali , Knapp .i . , US patent , ps 4044 ybJ (|y/a) TABLE Adsorpt i on-t'lut ion characteristics of "Mo and 'J * J on Che 1 ex 100 H2 SU< : l.b M ; NH-i SCW : O.b M ; Volume of eluent : iuu.u ni 1 . ; Flow rate : o-lo drops/mm. IChelex 100 I «•• Mo adso I "'I 111 i Imesh size rbed in eluate 111 percentage percentage bO - loo 100 - ^00 ] 200 - 400 I AR - 17.2 SEPARATION OF CARRIER FREE "mTc FROM 99Mo OVER CERIA COLUMN D. X. Bhattachuryya and N. C. Dutta Nuclear Chemistry Division Saha Institute of Nuclear Physics IIAF, Bidhannagar, Calcutta 700064 SUMMARY Ccria e*<-h;>n>!t*r v:>s prepared !iy precipitation with ammonia from a solution of reiii sii!(>!'.:iic in hut dil. H SO . Radiochemical separation of '"Tc from Mo lias b?n\ achieved by applying a very simple chemical procedure. y-spennim obtained for the separated '"Tc was of high rarJi«fi'"-!i'iic p"i'iv Tl^ piucosN is irpi Keywords : Ceria exchanger, sepai'ation of Mo - mTc I. INTRODUCTION hi.'julu'ili1. tiii.ulii'.ii'ent oxides such as SiO ,Zr() . TiO :\ II. EXPERIMENTAL Mo in equilibrium with "'Tc was supplied by BARC, Bombay. Ceric sulphate, anim III. RESULTS AND DISCUSSION y-spectruni showed that the separated 9 mTc was of high radionuclidic purity. The chemical procedure was extremely simple, quick, with quantitative yield. 99m Tc ( K0.6keV) (b) o IA o O Mo(740keV) I 99Mo (778.5keV) oo u 107 453 800 Channal Number Fig. 1. Gamma Spectrum of (a) 99Mo -99mTc ',n equilibrium , (b) Separated "mTc from the mixture . AR - 18.2 DEVELOPMENT OF A COLUMN GENERATOR FOR 9»™Tc USING LOW SPECIFIC ACTIVITY B9Mo FOR RESEARCH IN TECHNETIDM CHEMISTRY K.K.Kothari, M.R.A.Pillai Isotope Division B.A.R.C, Bombay 400085 INDIA SUMMARY A simple and convenient column generator system for the separation of 9B'nTc from low specific activity 93Mo produced by BB irradiation of M0O3 is described. MoO4~ at acidic pH was loaded BBm on a preconditioned alumina column and Tc0a~ was eluted with saline in an evacuated vial. Molybdenum breakthrough was negligible and the radiochemical purity of the eluted as>IT>Tc0.»~ was >95%. The preparation of the generator and subsequent elution process is simple and rapid. The method can be conveniently adapted for preparing BB""Tc upto 5 mCi for research purpose. (Key words : Technetium chemistry, generator) I.INTRODUCTION Research in technetium chemistry is often carried out with the long lived isotope BBTc spiked with tracer quantities of O9mTc. The concentration of BSTc is taken such that when replaced with BB*"Tc it will give the required amount of radioactivity for nuclear medicine investigations. BBmTc is used at low concentrations just sufficient for giving adequate counts and hence the amount of BBmTc needed is very small, the daily requirement does not exceed 5 mCi. The generator currently available in India is based on solvent extraction separation to obtain the high quantities of BOmTcO*~ from low specific activity B9Mo for nuclear medicine applications /I/. Though well established and widely accepted, the extraction procedure is tedious and time consuming compared to the column generator technique. We have developed a simple .and convenient semiautomated column generator using the low specific activity BO 9B > Mo produced by the irradiation of Mo03 for obtaining " Tc0d~ in quantities needed for research studies. II.EXPERIMENTAL 5 g of alumina was taken in a doublf ended glass vial fitted with a sintered glass disc at the centre. The vial was sealed from both the ends. A reservoir containing 0.9% saline ("250 niL) was set up and connected to the column through an infusion tube and needle. The column was washed with 100 mL of dilute HCl (~pH 4) by attaching a 100 mL evacuated vial through a tube and needle BS to the bottom of the column. 5 mCi (~200 uL) of Mo04- (specific activity varied from 93 to 379 mCi/g from batch to batch) was loaded on the column after carefully adjusting the solution pH to 4. The column was washed with another 100 mL of saline t <> remove B9 BB u9 S! m unadsorbed Mo as well as -"Tc and Tc. > TcO4- was eluted for use from the following day onwards. Elution yield was estimated by comparing ""Tc activity with BBMo activity on the BB 1 column. Radiochemical purity of " TcO4" was checked by paper AR - 19.1 chromatography in 0.9% saline. The rat! lonuc 1 idic purity of the 99l"Tc was checked in a gamma spectrometer. The quality of the 99l "Tc04" was checked by labeling studies using diethyiene triamine pentaacetic acid (DTPA) which was labeled with ®aTc using stannous tartrate as reducing agent by following a standard recipe /2'. Labeling yield was checked by gel chromatoyraphy over Sephadex G 25 (0.5x10 cm) column using saline at- eluent / 3 / . III.RESULTS AND DISCUSSION 98 '"TcO4~ gets eluted between 4 and 10 mL. Elution yields, radiochemical purity of ""TcOa" and label ing yield for DTPA from different batches are given in Table 1. Elution yield ranged from 80 to 93% in different batches. Radiochemical purity was more than 95% in all batches. aBMo contamination in the eluted TcO.»~ was negligible (0.0003%). The OBmTcO,»- prepared by this method gave good labeling yield with DTPA. aaMo-n9rnTc generator based on the procedure described above are regularly prepared and used for labeling of a number of chelating agents at the Isotope Division /4/. IV.REFERENCES 1. R.S.Mani, D.V.S.Narasimhan, 'Radiopharmaceuticals and labeled compounds' STT/PUB/344, Vol 1, IAEA, Vienna, pl35 (1972) 2. W.A.Volkert, D.E.Troutner and R.A.Holmes Int.J.Appl.Radiat.Isot, 33, 891 (1972) 3. W.C.Eckelman, G.Meinken and D.Richards J.Nucl.Med.,13, 577 (1972) 4. K.K.Kothari and M.R.A.PiJlai Abstract in Ind.J.Nuc1.Med, 1990 (In Press) Table 1: EJution efficiency and rau j ochetnica 1 purity of nTc0< ^r,d labeling yield of DTPA Batch Sp.act. <\ Elution RC purity OBTc-DTP£ no. of9'Mo % yield (%) rnc i / g 1 2 3 4 (average) 1 '"Teo" " 87 80 _ _ ... 2 180 - - 84 84 98 96 3 160 91 85 83 93 97 95 4 93 88 • - - - 99 5 133 - 86 93 - 98 - 6 197 80 89 - - 99 - 7 lbO 87 89 - - 98 95 8 379 88 87 - - 98 95 9 15J 90 81 92 90 98 _ AR - 19.2 RAUiOI,ARFI/;NG OF TESTOSTERONE WITH !Zl!iI FOR USE IN I Grace Samuel and Meera Venkatesh Isotope Division, B.A.R.C, Trombay,Bombay,INDIA, PIN - 400 085 SUMMARY Radio-iodination and purification of Testosterone, an important, wf.ercid hormone, is described. Tesi.osterone-3-c.a rboxyw't.hy 1 ox i me was coupled to 125I-histamme by mixed anhydride cond^nsa!. i on reaction /],2/.The radioactive products were separated by bolveni extraction 'ollowed by TLC. x25I-testosterone of higti specific act. ivity, good immurioreact i v i ty and long stability was obtained. (key words: Testosterone, radioiodination, radioimmunoas^ay, ' •*s 1 , mixed anhydride, histamine) i.INTRODUCTION P .si! i o i mmunoassays ( RIA) are commonly used for the measurement of must of the hormone levels in biological fluids. Unlike the large protein or polypeptide hormones, steroid hormones do not have convenient chemical groups for direct radio-iodination with 125i which is the most ideal radioisotope for RIA. Hence such small molecules are chemically modified and the derivative is coupled to a rad io-iodmated molecule. Rad i o- iodi nation and purification of Testosterone, for use in RIAs is described here /3,4/. II.EXPERIMENTAL 2my Testosterone-3-oxime was dinsolved in 50ul dioxane(anhy.). To this, 10 ul each of 1:5 diluted t. r J buty lamine and 1:10 diluted i aobutylchloroformate (dilutions made in dioxane) were added. The !>.i :<•• ijre was stirred, incubated for 30 minutes at 4-6°C, and then i.h'-1 activated oxime derivative was diluted, to 2.8ml with dioxane. To prepare iodinot^d histamine,1OOng of histamine in lOul of 0.5M phosphate buffer,pH 8.0, was taken in a tube.To this 37MBq Nai25I and 50 uy Ch lorami ne-T in 10 ul of distilled water were added arid mixed. Aftfir 1 minute, the reaction was arrested by adding 300 ug of Na2H^O5 in 100 ul of distilled water. 50 ul of the activated testosterone-3-oxipe was added to the iodination mixture. 10 ul of 0.2N NaOH was also added and the mixture was incubated for 2 he at 4°C. The mixture was acidified with 0.9ml of 0.1M HCl and extracted with 1 ml toluene. This toluene layer was discarded and the mixture was neutralised with NaOH. 1 mg of Na2S2O3 was then added and the mixture was extracted with 0.5 ml toluene. TLC was carried out on this extract using Benzene: Ethanol: Acetic acid in 75:24:1 proportion, as the solvent system. The radioactive peaks were identified and were tested for specific binding to the anti-testosterone serum and non-specific binding (NSB). This was done by incubating an aliquot of the product with and without anti-testosterone serum for 18 to 24 hr at ambient temperature and then separating the bound complex of testosterone from the agent to adsorb free testosterone. The fraction with lowest NSB and high immunoreactivity was extracted w th 1-2 ml ethanol and stored at -2 0°C for further use. AR - 20.1 I I I.RESULTS AND DISCUSSION The yield of histamine labeling was; 45%, as estimated by paper electrophoresis.The distribution of radioactivity after TLC separation is shown in fig.l and Table-1 lists the data on NSB and immunoreaetivity of the three major fractions constituting the radioactive peak. Of these, the fraction at Rf 0.50 is the best with lowest NSB, arid maximum 1 mmunebind lr.g indicating that pure intact molecules retaining its immunebinding constitute this fraction, while the other fractions could be contaminated with damaged molecules with low binding and high NSB. Hence the TLC strip at. Rf 0.50 was extracted and used for further studies. Fig.2 depicts the quality of the tracer with time, w.r.t. NSB and Immunoreaei ivit y. The tracer was stable for 6 weeks without any deterioration, after which the NSB started increasing slowly. However most of ':he irairiunoreactivity was still retained indicating that there could be some radiolytic damages giving rise to small fragments of active non-1 iirnunoreact i ve molecules. A repurification on TLC could perhaps give good quality tracer in such cases. IV.REFERENCES 1. B.F.Eiianger, F.Borek, S.M.Heiser and S.Leiberman, J. Biol . Chem., 228, 713 (1(»57>. 2. P.W.Nars and W.M.Hunter, J. F.ndo., 57, XLvm (1973). 3. J.I.Thorell and S.M.Larson, 'Hadloimmunoassay and related techniques'. The C.V.Mosby Co., St. Louis(1978). 4. R..Edwards, E.D.Gilby and S.L.Jt'f fcoate, ' Rad loi mmur.oassay and related procedures in medicine',11, IAEA, Vienna (1974). TA8LE-1 Rf XNSB Xltnmunebinding j 0.50 6.3 70.0 i 0.56 10.9 64.0 0.63 29.9 24.5 Distribution of radioactivity Study on the tracer quality with time after TLC purification * immunebiodifW' . NSB activity % 25 (MBq) 3i nding / is 20 40 SO —*• time (days) fig.1 Fig-2 AR - 20.2 LABELING OK F.STRADIOL WITH 125I AT DIFFERENT POSITIONS AND THEIR USABILITY IN RADIOIMMUNOASSAYS Aruna Korde, Meera Venkatesh and Ramji Lai" Isotope Division, B.A.R.C., Trombay, Bombay, INDIA, PIN 400 085. <" Radiopharmaceutical Operations,BRIT, Thurbe Complex, New Bombay, INDIA, PIN 400705) SUMMARY Three different radioiodinated Estradiol preparations were made by direct labeling at the phenolic ring and labeling at positions 6 and 17 through histamine conjugation. These were tested for their usability in radioimmunoassays (R1A) in terms of imrnunoreacti vi ty, non specific binding (NSB), stability etc. (Key Words: Bstrddio1 . Rad ioimmunoassay, radioiodination, i-!!ST) I.INTRODUCTION Radioimmunoassay, a sensitive and a specific tool for hormone level measurements in biological fluids, employs radiolabeled analyte as the tracer. 125I is an ideal radionuclide for RIAs; but most of the steroid hormones do not possess convenient chemical groups to facilitate direct radioiodination. Estradiol which has a phenolic ring was radioiodinated at three different positions and compared to assess their usability in RIAs. II., EXPERIMENTAL Direct iodination of Estradiol was carried out following the general procedure of Hunter and Greenwood /I,2/. Estradiol is allowed to react wiAh Na123I in presence of the oxidising agent Chloramine-T for the required time duration and Na2S2Os in added to arrest the reaction. To iodinate at positions 6 and 17, Estradiol -6-CJ rboxyinethy loxime and Estardio] -1 7-hemisuccinato were at first activated by the mixed anhydride method /3/ and then conjugated with the radioiodinated histamine. Histamine was labeled with l:2SI as reported by Nars & Hunter /4/. In all the three cases the products were purified by solvent extraction at first followed by TLC . After purification on TLC, the fractions were tested for non-specific binding (NSB) and binding with anti- estradiol serum.The fractions with low NSB and good imnranebinding were extracted with ethanol and stored below 0°C for fusrtl^er use. Two different antisera were used for testing the binding. One was raised by us against Estradiol-6-oxime-BSA conjugate in rabbita and the other was against Estradiol-6-oxime-KLH from a commercial source. The preparations with good immunebinding were tested for their usability as a tracer in RIA. The stability of the tracers was studied by estimating the NSB and immunebinding at regular intervaIs. III.RESULTS AND DISCUSSION The details on yeild, specific activity, immunebinding, NSB, etc. of the tracer preparations are listed in the table-1. The yield of direct iodinations were higher compared to the others. But, the binding with excess ant i-Est raidiol sera was just ~60%. AR - 21 ..1 Stability of tracers obtained by direct iodination was also poor as seen from the fall in immunebinding from 60% to 36% and rise in N£B from 5% to 22% within 20 days (Fig.l). Iodinations were carried out using lesser radioicdine to check if the quality improves on reducing the specific activity. However there WAS no oerceptible improvement. The tracers labeled at the 6 position were the best of the three, wLth low NSB, good immuncreactivity and shelf life ol more than 6 Weeks. The racer behaviour was identical with both the antisera. But this :racer did not yaeld a sensitive RIA system with either of the antiserum, which may be due to the use of same 6 position for conjugation for preparation of the tracer and antibodies. The iodinations were repeated with much smaller amounts of oxime and the products were found to be comparable.This was to ensure that TLC separates the radiolabeled oxinifc from the cold oxime giving high specific activity tracer. Thus, in this case, if an antiserum with grcrter avidity which does not exhibit bridge binding is used, better results would be obtained.The tracer at position 17 did not show any immunebinding though the NSB was low. Thus, it car> be seen that the position of conjugation for antibody eiicitation plays an important role in deciding the type of tracer to be made. Many commercial kits use the same conjugation position for both tracer preparation and antibody raising though there have been reports advising the contrary too /5/. These studies also suggest the use of aame position ,with a caution to choose the right antiserum. IV.REFERENCES l.W.M.Hunter and F.C.Greenwood, Nature 194, 495, (1962). .M.Malmquist and J.I.Thorel1,J. Clin. Endo. iletab . , 49 , 655 , (1 979 ) J.E.T.Corrie and W.M Hunter,, Methods in Etymology V 73, edited by H.Van Vunakis and J.J.Langone, Academic Press (1981). P.W.nar," and W.M.Hunter, J. Endo., 57, XLVii, (1973) . E.H.D.Cameron, J.J.Scarishriek, S.E.Moris, S.G.Hillier and G.Read, J. Steroid biochem., ": ,749, (1974). TABLE-1 Tracer stability studies 80 > • • • .a Position of phenolic ring position 6 position 17 labeling a & d - immunebtndmg & NSB - position 6 SO b b ft c - wnmunebniding 8: NSB X yield 80 40« 40* ' N - phenolic ring \ specific activity 2.5 ~80.5 "80.5 (MBo / nM) •40 • X NSB 5 2 2 c ' 20 % Immunebinding 6." 85 " 0 X binding d . . . . t * this yield is for histamine labeling; total yieW is " 15% K> 20 30 40 50 age of tracer (days> Figure - 1 AR - 21 .2 LABELING OF INSULIN WITH n7Co FOR RADJOIMMUNOASSAY Grace Samuel, M.R.A.Pi 1lai Isotope Division B.A.R.C. Bombay 400 085 INDIA SUMMARY Simultaneous RIAs can be carried out using i=5i and B7Co labeled antigens. The procein hormone insulin was first modified by the addition of the chelating group OTPA to it by controlled incubation of the hormone with bicyclic anhydride of DTPA. The modified protein was purified by gel permeation chrornatography and then labeled with 57Co. The labeled protein was purified and reacted with insulin antibody to see the immunoreactivity. ( Key words : Radioimmunoassay, bifunctional chelating agents, DTPA anhydride, 57Co ) I.INTRODUCTION In simultaneous RIAs two hormones are measured in the same assay by labeling the antigens with 12Si and 5"'Co separately /1,2/. The assay system contains the antibody specific for both the antigens and also standards of known concentration for both of them. RIA is carried out as usual and at the end of the assay quantitation of the extent of the antigen-antibody reaction is done by counting the separated antigen-antibody complex at two different channels adjusted to the corresponding energies in a gamma counter. While protein hormones can be labeled with iodine with ease, 37Co cannot be incorporated to the protein directly. Though simultaneous RIAs for hormones are available from commercial manufacturers, the method of labeling the proteins with the 57Co has not been disclosed due to the commercial importance. Labeling of protein hormones with metallic radionuo1ides using bifunctional chelating agents originally described by Sundberg et al is now extensively used /3/. Hence we have tried labeling insulin with 57Co using the bicyclic anhydride of DTPA as the bi functional clielating agent /4/. II.EXPERIMENTAL Ins-.*! in was coupled to diethylene t r i ami nepentaacet j c acid (DTPA) ctt a ratio 1:3. 3.6 mg of insulin was initially dissolved in dilute 1IC i < pH 3). The pH was then adjusted to 8.0 with 0.5M bicarbonate buffer < pH 8.0). fine mg of DTPA anhydride WdS added as solid to it and allowed to react for three hours at 25 ° C. The reaction mixture was then loaded on a 25x1 cm pre-conditioned column of Sephadex G-75 and eluted with normal saline. 2 ml fractions were collected and the protein peak was ident ified by the Eiolin-Ciucalteau's method. 50 ug of the insulin DTPA was 57 mixed with CoCl2 (~10-20 uCi). The pH was adjusted to 6.0 with citrate buffer. After a reaction time of 0.5 h at 25 " C, the unreacted cobalt was removed by rjel filtration as earlier. A />« - 22.1 blank experiment, was also set up parallel by incubating 50 ug of 5 insulin with the same amount of ToCl2 in order to rule out the possibility of any radioactivity nonspeeifica11y adsorbed to insulin. The labeling yield was estimated by paper ehromarography i n sa 1 i ne. An aliquot of the labelled insalm < 20,000 dpm) was allowed to react with in:;ulin antibody ( 1:600 diluted) for 3 h at 2 5° C. The antibody bound insulin was separated by precipitation with polyethylene glycol. The radioactivity in hound fraction was measured. I I I.RESULTS AND DISCUSStON The unreacted and hydrolysed DTPA was separated from insiil in coupled DTPA by column ehromatog raphy over Sephadex G-75. The labeling yield was 81% as determined by paper chromatcgraphy. The radioactivity adsorbed to protein in the absence of DTPA was only about (>'%. This showed that the radioactivity associated with insulin was essentially through the chelating agent DTPA. The binding with insulin antibody was 40% on the day of labeling. The radiochemica1 purity reduced drastically after two dayi to 69% and the tracer showed no binding with antiserum indict ,ing the poor stability of labeled insulin. The poor stability may be due to the release of cobalt from the chelating agent with time as indicated by I he: poor rad l ochein I ca 1 purity. One of the main reason for the low stability may be due to the extremely small concentration of the 1 i gand-nieta 1 complex present in Rlh studieu like this. The above studies demonstrate that while it is possible to label protein hormones with 57Co through bifunctional chelating agents ,the bieye lie anhydride of DTPA may not be the right choice. IV. REFERENCES 1. M.T.P.Rock, M. S-Schneider , [) . M . F l a 1 kowsk l , Clinical Chemistry, 33, 922 (1987). 2. R.K.Desai, W.M.Dcppe, R.J.Norman et al.. Clinical Chemistry, 34, 1488 (1988). 3. M.W.Sundberg, C.F.Meares, D.A.Goodwin, C.I.Diamentl, Nature, 250, 587 (1974). 4. D. J .Hr.at.owich, W.W.I.ayne, H.L.Chi'ds et a\ . , Science, 220, bl3 (1983). AR - 12.2 STUDIES ON PREPARATION AND ASSAY OF CAKilON-14 LABELLED OXAL.TC ACID T. V. Ramamurthy, V.Jayakumar, and Harish Chander, Labelled Compounds Operations, Board of Radiation and Isotope Technology, Nsw Bombay 400 705. Oxalic acid-1-14C is prepared in 70% yields by oxidation of glycollic acid-1-l4C by alkaline permanganate and purified by ion exchange chromatography. It is shown that the sensitivity of the colorimetric method for the assay of oxalic acid is increased when the reagent indole is substituted by 5-bromoindole. Key words : carbon-14 labelling, oxalic acid-l-C-14, 5-bromoindole. INTRODUCTION Carbon-14 labelled oxalic acid finds applications as a radiotracer for d.ilineating its role as a metabolite in biochemical studies and is also used tor radiometrie assay of enzyme oxalate decarboxylase. Several procedures have been reported for the synthesis of C-14 labelled oxalic acid using l*CO2, formate-14C, propionic acid-l-14C 3~hydroxypropionic acid-1-1 4C and acetic acid-1-14C via glycollic acid-1-1*C1-5• We have earlier prepared oxalic acid-1-14C using the 3-carbon compounds by permanganate oxidation ; though the yields are good to moderate, 70% from propionic acid and 45% from f$- hydroxypropionic acid the specific radioactivity of oxalic acid is lowered owing to the clevage of C-C bond involving the labelled carbon. On the other hand using glycollic acid-14C, oxalic acid of same specific activity has been obtained6 . Since glycollic acid-1- 14C with high specific activity is an easily available labelled precursor in our laboratory we considered it worthwhile to evaluate the procedure from carrying out small scale synthesis7. Our studies in this direction are presented in this paper. EXPERIMENTAL Glycollic acid-1-14C is prepared from potassium cyanide-1*C and purified as per our earlier procedure. Known aliquot of glycollic acid (l-5mCi, 0.1-1.0 mmole) is rotary evaporated to dryness, taken up in 0.2-2 ml of of KOH (1 g/ml) , potassium permanganate (30-260 ing) is added to the solution and the mixture rapidly refluxed for 30 minutes. After cooling the excess permanganate is destroyed Ly hydrogenperoxide the solution is centrifuged three times with hot water and the supernatant solutions are combined. The solution is treated with a slurry of Dowex 50 X8 resin in H* form and the oxalic acid-14C obtained as an aqueous solution is counted. Aliquots are analysed by paper chromatography in ether: acetic acid: water (13:3:1) V/V and ether : formic acid: water (5:2:1) V/V followed by autoradiography. If glycollic acid is present (the first solvent system would reveal it) oxalic acid is again chromatographed on anion exchange resin loaded column and purified. Assay of oxalic acid-14C is made by adding 1 ml aliquots of oxalic acid to a solution (1 ml) of 5-bromoindole in concentrated sulphuric acid { 10mg in 10ml) and heating the solution at 90°C for 45 min. The absorbance is read at 526nm using a capillary cell (2 ml) and the amount is computed from the calibration curve. AR - 23.1 Ill RESULTS AND DISCUSSION The oxidation of glycollic acid is a convenient reaction for scaling down to submillimolar levels as shown by our experiments. (SeeTable) Owing to the small quantity of oxalic acid, the purification of oxalic acid is effected by anion exchange column chromatography. We have noticed that glycollic acid remaining unreacted is in trace amounts usually around 5 % or less and could be separated from the product oxalic acid by anion exchange column chromatography very efficiently. The detection of unreacted glycollic acid in the reaction mixture is effected by suitable choice of paper chromatographic solvent system which would give distinct spots for both oxalic and glycollic acids and well separated from each other. We have tried many systems and found that ether : acetic acid : water (13:3:1) V/V is good for this purpose and ether: formic acid: water(5:2:l) V/V is satisfactory for radiochemical purity determination of oxalic acid-14C . Assajr of Oxalic acid. Initially v>e attempted8 the colorimetric method of Bergerman using indole but as it required large amounts (400 ug ) of oxalic acid we attempted related compounds such as 5-bromoindole and 5- benzyloxyindole to see if sensitivity could be improved. While indole and 5-bromoindole are found to give the same colored product, 5-bromoindole gives nearly twice the absorbance for the same quantity of oxalic acid ( 5-benzyloxyindole fails to give any coloured product) thereby enabling to use less of the labelled oxalic acid for chemical assays. Good linearity is observed over the concentration range 50-250 ugm of oxalic acid. Table 1: Synthesis of '*C labelled oxalic acid. Precursor used Amount Radiochemi cal mmole Radioact i viv yield Glycollicacid-1-1 * C 0. 25mmole 1.0 mCi 70% -do- 1 mmole 5 7 70%» -do- 0.1mmole 0.57 70% -do- 0.lmmole 0.50 65%b a : Specific activity by 5-bromoindole 5.8 3 mCi/mM. b : Slight excess of KMnO4 is used for oxidation. Acknowledgement: The authors thank Mr. C. N. Dtsai, Senior Executive, Labelled Compounds Operations, BRIT, for encouragement. IV REFERENCES 1. R. J. Speer, A. Roberts, M. Maloney, H. R. Mahler, J. Am. Chem. Soc., 74, 2444 (1950). 2. P. Nahinsky , S. Ruben, J. .V i. Chem. Soc . , 63, 2275 (1941) 3. M. D. Kamen et.al., J. Am. Chem. Soc.,__64, 2229 (1940). 4. F. Kogl, U. Halber°Ladt, T. J. Barendrest, Rec. Trav. Chim., 6fc, 387 (1949). 5. K. Bernhard, r.Brubacher, H. Jacquet, Helv. Chem. Acta, 36, 1968 (1953) . 6. H. L. Peeters , N. A. Rumbaut, Radiochimica Acta, 5, 109 (1966). 7. T. V. Re.iiamurthy, K. V. Viswanathan, J. Labell . Compds . & Radiopharm. , 24. , 995 (1987). 8. J. Be.german, J. S. Elliot, Anal. Chem.. 27, 1034 (1955). AR - 23.2 HOFMANN ELIMINATION OF p-NITROPHENYL3THYL-l-C-14-TRIHETHYLAMMONIUM BROMIDS-A CARBOM-14 ISOTOPE EFFECT STUDY. T.V.Ramamurthy* and Arthur Fry r Department of Chemistry and BiocherIstry, University of Arkansas,Fayetteville, AR72701, U. S. A. SUMMARY The alpha carbon isotope effects in the Hofmann elimination of the title compound have been measured under changing buffer concentrations with a vie* to correlate mechanistic change. Since there are alpha-carbon isotope effects and the effects are small it is quite likely that the reaction is of the ElcB type, predominately irreversible, with the incursion of slightJy increasing fractions of reaction by the reversible mechanism as the buffer concentration is increased. Key words: Kinetic isotops effects, Carbon-14, Hofmann elimination, ElcB mechanism. I INTRODUCTION: Traditionally, the Hofmann Elimination has been considered to proceed by an E2 mechanism1. Our previous aC and nC isotope effect studies on a series of p-substituted (3-phenylethyl trimethylammonium bromides in ethanol/ethoxide ion clearly show that there are bonding changes at both the a and (3 carbons in the rate- deteriisining s-tep of the reactions, confirming the E2 mechanism2. However, the magnitude of the aC effect decreased as the substituent became more electron-withdrawing, indicating increasing ElcB character. Keefe and Jencks have reported, that in aqueous buffers the mechanism for the p-nitro compound is of the ElcB type, changing from irreversible, k2 > k-1, at low buffer concentrations to reversible k-1 > k2, at higher buffer concentrations3: /^T\ + B-,ki /~\ ~ + NO* -OJ>-CH2 -CHJ -NMes > K02- ( Q) -CH-CH2-NMe3 .£ \ I HB, k-i ! I E2 Mechanism CH= "^ ~ * ' * ~ ' > NO2 -(O)" CH2 + NMe3 II EXPERIMENTAL ^ The p-nitropheny]ethyl-l-C-14-trimethyl ammoniumbromide available from previous work was recrystallised to constant molar activity end (0.5-1.0gm) used for kinetic isotope effect studies. The isotope effect experiments in buffer solutions were carried out in a constant temperature oil bath at 65+0.1»C. The kinetic isotope effect k/K k was obtained by measuring four parameters, namely (i) fraction of reaction, f, (ii) molar activity of the reactant, Ro , (iii) molar activity of recovered reactant Rr, and (iv) molar activity of the product R,> at each traction using the equations of Tong and Yankwioh4. For each buffer concentration and fraction k/°k, the value was obtained by four independent equations (See Table 1). Snch value in Table 2 was the average of 20-25 and this increased the confidence in our isotope effect numbers . The fraction of reaction f was determined by UV spectrometry (the product of the reaction, p-nitrostyrene was monitored). The molar radioactivity Kaasurtments were made on p-nitrostyrenedibromide after exhaustive purification by crystallisation and vacuum sublimation. At every fraction of reaction, the reactant and the product were separated by AR -- 24.1 liquid-liquid extraction with n-pentane. The pentane Isyer containing the product, jj-nitiostyrene was brominated and the dibromide was measured for RP. The molar activity of the starting material, Ro as well as the recovered reactant, Rr were made after treating the starting material and the aqueo.a layer with sodium hydroxide and converting the product p-nitrostyrene into dibromide by bromination,, III RESULTS AND DISCUSSION The alpha carbon isotope effects under all buffer concentrations and in the absence of buffer studied are small. (See Teble 2) For a reversible ElcB mechanism (or an B2 mechanism), an isotope effect would be expected at aC, whereas for sn irreversible ElcB mechanism none would be expected. (The corresponding isotope effect for this compound in the ethanol/ethoxide ion system was 1.019 + 0.003.) That there is a "C isotope effect at all means that there is rate- determining bonding change at the alpha carbon, at least for a fraction of the reacting molecule'-. This result requires at least partial reaction in all cases by the reversible ElcB mechanism (or by the E2 mechanism), and is inconsistent with 100 % reaction by the irreversible ElcB mechanism. At the same time, the low values of the isotope effect probably rule out 100 % reaction by the reversible ElcB mechanism. To the ex' enr that the tr^.id in the above numbers is significant, it is in the direction pxpecced from the results of Keefe and Jencks: increasing isotope effect v.ith increasing fraction of reaction by the reversible ElcB inechanisn as the buffer concentration is increased. Table 1 : Carbon-14 Kinetic Isotope effects, k/°k, in the Elimination Reactions in 0.08 M Ace tohydroxarate Buffer at 65^. Fraction k/a k ca 1 culated from: Of ]reaction Ro # Rr , Rp Ro R, , f Ro , RP , f Rr ,RP , f 0..30 1.015 1 019 1.014 1.016 0..43 1.012 1 0 2 2 1. 009 1.01 5 0..46 1.015 1 020 1. 014 1 .017 0..55 1.009 1 018 1. 005 1. 012 0..66 1.019 1 026 1.014 1 .022 Average k/'" k - 1.015 + 004 (std . d e v .) Table 2 : Kinetic Isotope Effects Wi th change of [Buffer] [ Buffer] k/"k + std . dev. 0.14 M 1 .016 fl 00, 5 0.08 M 1 .015 0. 004 0.04 M . 1 .012 0. 004 0.02 m 1 .Oil 0.002 (NaOH only) 1 .011 0. 003 IV REFERENCES: 1. J. F. Bunnett, Auyew. Chem., Inl.i. Ed. fclnyl . 1, 225 (1962). . 2. A. Fry, L. B. Sims, J. R. 1. Eubanks, T.Hasan, R.Kanski, F. A. Pottigrew, S. Crook, Proc. int. Syrup. Syn. Appl. Label. Compds, W. P. Duncan, A. h. Susan (Ed.), Elsvier Scientific Publishing Company, Amsterdam, 133 (1983). 3. J. R. KeeZe, W. P. Jencks, J. Amer. Chem. Soc, 105, 265 (1983). 4. Y. Tong , P.K. Yankwich, J. Phys. Chem.,61,540 (1957). ' Post-doctoral Research Associate , Permanent address: Labelled Compounds Opeiatiuns New Bombay 400 705. A R - 2 A . Z ION-EXCHANGE SEPARATION OVER THORIUM TUNGESTATE COLUMN B.Sarkar and s. 3asu Department of Chemistry, University of Burdwan, Burdwan 713 104 140 140 137 SUMMARY : Separation of carrier-free La from Ba and Cs from 23B over Thorium tungestate column is described. (Key words : Thorium-tungstate, carrier-free separation) INTROOUCTION : In continuation of our studies with insoluble tungstates, ' we have synthesisei amorphous thorium V.ungstate ion-exchanger. Present work describes carrier-free separation of 140 140 a genetically related pair ( Ba - La) and a fission-product f Cs) from natural uranium, using this material. EXPERIMENTAL : About 2gm thori ,,m tungstate(Th ; W = 1.0 : 1.8 ) was packed in a glass column(0.5 x 10 cm) and the equilibrium mixture ' Ba - ' La (supplied by 3ARC, Bombay), pre-treated to pH 1-3, was fed on to this column. Unabsorbed La-fraction completely washed out by deminerlisei water and measured for y — activity. The parent was latter elute i with «2(i Hei • Again in another column, a mixture of cs and U (taken as uranyl nitrate) previously conditioned at pH 2-3, was fed and washed with water to remove all uranium. Then Cs-13 7 was eluted with 4tC He land measured for i ts 'y -activity. Both the measurement were carried out in a Princetony -TECH Qe co-axial detector of active volume 30crn with a system resolution of 2.06 KeV FWHM for l.33MeV/-ray of ' Co coupled with a ND-8K M.C. analyser. DISCUSSION : A comparison ot: T-spectra of La with that of mixture .shows that the characteristic peaks are absent of parent indicating a complete absorption of ^a-140 by the exchanger. Distribution co-efficient values are also in agreements with this conclusion. Similar conclusions are drawn from U - Cs system. The peak at 657 keV corresponding to Cs-117 is absent in uranium fraction. The over all methods are rapi I, simple and quantitative. ACKNOWL'-:OGK:-1.£NT : * 3000 00 ( a / ; ooo no 1000 00- 0 00 *1" * f -•100 00 ?00 00 (b) o o 1 JO 00 • aoo- 500:00 1000.00 150000 20UCOO 0.00 Energy in KeV .- c , u.O W UUO0 y /bpecira o< t]) Bo —La mixture bKeparuted la AR - 25.2 IMMOBILIZATION OF 3ARIUM, ANTIMONY AND CADMIUM FROM AQU'CUL'3 SOLLTIJU 3Y ilKCC MIUM OXIDE D. K. Bhattacharyya & N. C. Dutta Nuclear Chemistry Division Saha Institute of Kucl-ear Physics 1/A7, Biclha:magar, Calcutta-700064 India. SUMMARY The i mmobilization of Barium, Antimony and Cadmium ions waa studied ovc the stablfo and dense type adsorber of •zirconiuia oxide. Very hiyh and appreciable uptake of 140ga, l?%b and H^Cd by zirconium oxide was observed. Soxhlet flow leach test in de-ionis^d witer at 97oc showed that amount of leached out cations were- ve,y negligible. X-ray diffraction pattern of the oxide samples afcer ad-surpt i. on indicated some structural changes, nead3 further ir.vsstigati on. (Keywords % ''''as, "sh and Cd, Immobilization, zirconia) I. INTRODUCTION Study on the immobilization of radioisotopes from nuclear wastes over solid materials such as borosilicate glass, sjnroc, cement, insoluble metal oxides, etc. has attrac- ted in recent times much attention in connection with the dis- posal of hazardous and toxic elements. Here the extent of upkake of 2*°Ba, 1253t,(. U5cd isotopes and the immobilization in welghaible quantities over zirconia have been studied. II. EXPEPTMSNTAL 140 125 115 . — -•» Ba ^ sb and ii c Bombay. Zi'-. .•-.•lum oxychloride, barium nitrate, antimony oxide, cadmium nitrate and all other reagents were of AnalaR grade. Zr-hydroxide was prepared*' frc!i an aqueous solution of ZrOCl-. 8H-0 by precipitation with ammonia. Uptake determined by sna- king each of the isotopes with 0.5 g zirconia in 50 ml water for 24 hours, p -activity of the solution part was then measured by the Philips tyye G.M. liquid counter and Kp values were calcula- ted. Extent o£ adsorption of each of the cations was more than 95 percent. Each of Ba, Sb and Cd elements in weighable quantities were separately co-precipit?.ted. Ba as nitrate, Sb as oxiae in HCl, and Cd as nitrate were mixed separately with ZrGClj.QH^O solution and were precipitated by amraonia. These wera filtered, washed and dried at 70°C, the filtrate were ana- lysed spectrophotometrically for Ba, Sb and Cd ions. Adsorption were appreciable. The dried samples as pellets (13 mm x 0.5 ram) were calcined at 1200°C for 24 hours in furnace, cooled and leachablity (L) of the element were examined at 97°C in a Soxlet apparatus refluxed for 24 hours. Leached out elements were estimated in the solution by spectrophotometry and calculated as AR - 26.1 A •» amount of 'Leached' element(y), C = constant of element, in the Immobilizer (g), s =» apecific surface area of i. .nobilizer (cm '^), T • immersion tima (d) . Surface area before and after adsorption were mea- sured by applying N, gas adsorption BET method- x-ray diffrac- tion patterns have Been obtained by Philips Diffractometerif-^-i) III. RESULTS AND DISCUSSION 14 125, iiJ Uptake of °Ba, ""sb, Cd ware 95.6*%, 95.18% and 99.43/4 respectively. The adsorption at macro scale were 8a- 82.92%, Sb-97-17% and Cd-79.11%. The amounts leached out were 7 for Ba(3.16 x 10" ), Sb(12.l7 x 10-7) and Cd(2.7 x 10-7). Tne X-ray diffraction pattern showed that the zirconia as prepared was mostly of Baddeleyite type, hag suffered some structural changes due to fixation of B-i, Sb and cd and immobilized which can be used for the disposal of wastes and toxic isotopes. IV. R3F2R2NCE D. K. Bhattacharyya and N. C. Dutta, Proc. Int. Conf. Ion-Ex '90 NEWI, Wrexham, UK, p.67, 9-11 July, 1990. 10 20 30 40 50 SO 28 ( Cu - Koc) Fig .1- X- ray diffraction patterns of a) ZrO2 and b) Ba c} Cd d) Sb absorbed on ZrO2 matrix ( heated at IOOG'C) AR - 26.2 2f ADSORPTION OF ZK) IONR ON TiO2 POWUFR G.K.Miahra, Sujatha.S ctnd R.N.Singh Nuclear and Radiochemistry Laboratory Department of Chemistry, Banaras Hindu University Varanasi 221 005, India. SUMMARY 2< Adsorption studies of ZrO iona on TiQ2powder have been carried out at various concentrations (10"'° -10"BrO. A probable mechanism has been proposed. Key words : Arisorpt ion/ZrO f/TiO^ I. INTRODUCTION Adsorption \s on^ of the important methods for the separation and isolation of fission products. The preHenJ work deals with the adsorption behaviour of 7.r(TV) on TiO^ powder under varying exper imenl.rt I rondit ions. II. EXPERIMENTAL The adsorpt ion exper intents were performed by mixing 0.5 g of TiO^. powder as an adsorbent and 10 mi of aqueous solut. ion of ZrOCla. containing appropriate concent: rat. ion of 7.r-95 as a tracer. The experimental details are similar to thai of Singh ei.. al.Ill. The experiments were also performed at various pH. Ill* RESULTS AND DISCUSSION The results obtained on the adsorption of ZrO2"* ions on TiOi aurfs.ce are returned in table 1. The time required to attain saturation is approximately 15 minutes. It haa also been observed that the rapid and smooth adsorption of ZrOz'f iona on TiO^ surface at the initial stage becomes considerably slower with the lapse of time till saturation. The concentration of ZrO1*" jona harf marked influence on the adsorption as the adsorption increases with a decrease in concent rat ion* A plot of log (K/tTi> Against log C giv^s a value of l/n = 0.999. This uuggecits the possibility of the monolayer formation of i)rOz+ species on the surface of TiO2 and also the Freundlich adsorption isotherm is obeyed. The results obtained at various pH for the adsorption of ZrOz+ ions are returned in the table 2 which indicate that adsorption of ZrO1*" ion i is favoured in the alkaline range. The process of adtiorpt. ion involves the following Htepul2]|: -&« t °" + - ° Ti C -~ *Ti W "— " Ti C + 2H • ( >vm surf ace) (neutral murfmw) AR - 27.1 (a) Dissociation of Ti(OH)^in alkaline medium : - H+ HO -• O - -W HO ^ O" Ti (OH) ~ ^Ti \ ;=i-^_^ ^ Ti ^ + H * HO" XOH -t H* HO^^ ^ O_ (-ve surface) Thus , the dissociation of hydrated surface givea rise to a charged surface as following: 0 + O ;M— OH (.surface) —==-^ M O (surface) + 2H O O H I've charge) (-ve charge) (b) Di HHOC 1 at i on in acidic medium : Ti In al rciny acidic HDIIII ion, T ' O^ does not ex i si aw Ti*' or Ti + but polyrnerlseH to form polymeric tion i I i •j'-' solid surface |.-->adiny to a 1a 1 1 ad HO rpt ion of cat I oriH pa r 1 icularly /. r() IOIIH. Ti TI ... n - TaJ>J_e i Time v.-i r i nt i on of Zr()2t ions on TiOj powder al concent raLionK of ZrOCl^. al 'Ml ° C T l me Amount o f Zr ( IV) Jons adsorbed per 0.5 of 7 (mi n ) y X 10 5 g A 10^ g *. 10 g * io'° 5 0 .175 0. 204 0.212 0.228 10 0 .349 0. 359 0.372 0.385 1 5 0 .484 0. 503 0.515 0. 529 35 0 .484 0.504 0.514 0. 5 29 55 0 .484 0.503 0.515 0.529 b -7 -iu initial Cone. 10"5 1o" 10 10 1Vib) H 2 2 Adsorpl i on of '/.rO ions on TiOz powder al. di f ferent pH at .3 0" Time Amount o f 7,v < (V) 104 ) ions ad Horbed on 0.s g of Tio^> (mi n ) l.i 2.8 3.9 5.4 b.7 f .5 9 . H 5 0 .108 0 .131 0 . 147 0. )H2 0 .178 0. 206 0.21 9 10 0 .269 0 .30 2 0 . \18 0. 3 U 0 .344 0. Wlf> t) . 149 15 0 .409 0 .44 3 0 .458 0.4 75 0 .4Hb 0. 5 04 0 . 5 1b 35 0 .409 0 .44 3 0 .45 8 0.4 75 0 .4Kb 0. 5 0 9 (l.r> IS /I/ H.N.Slnyh, D.l.ahiry, Sujalha.K and G.K.Mishra t»ro<: . Had 1 ix-hem. and Radiation ehem. Symp. Naypur AI.-47, 1989 I'll J.O. I.ee, 'Concise Inorganic Ohemi«(ry', Van NOHI rand Reinhold Coinpiiny Ltd., London, 2nd Ed. 19 (1965) AR - 27.2 SOME ASPECTS OF IR STUDY ON FIXATION OF CHROMATE IONS ON HYDROUS ANTIMONY TRIOXIDE M. J1J311U.TANI, P.N. REDDY, A.K. MITRA, RAMEHd KUMAR I Department of Chemistry- Indian Institute of Technology, Hauz Khas, New Delhi-110016 SOMMAIiY: Infrared spectroscopic technique has been used to obtain a structural model for the surface reaction between hydrated antimony trioxide and chronrjate ions. The IR spectra of chromate species adsorbed on hydrous SbzOs shows the presence of CrO*2~ ions. KEY WORDS: Sb2Os, CrO«2 - , IR spectra I. INTRODUCTION: i'he study of fixation of various ions due to reactions that occur at the surface of oxide has been limited duo to 1-tck of direct evidence on which a structural model may be based. Infrared spectroscopy is potentially one of the most direct means of investigating the surface structures. The extensive study of IR technique on adsorption of phosphate and chromate ions on various other oxides^-3 has been directed particularly towards the interpretation of the shapes of adsorption model. In order to understand the mechanism ot the behaviour of antimony trioxida- solution interface, the sorpti<->n of chromate ions from aqueous solutions ha3 been studied by infrared spectroscopic technique. II.EXPERIMENTAL: Ml the chemicals used in the present investigation were of analytical grade. The samples used for IR spactroscopic analysis were prepared in accordance with the conditions shown below: Sample Preparation of Samples 1 ^Precipitation from SbCl3 by~addition~ojriM~NaOH 2 Sample 1 immersed in O.iM NaaCrO* at pH 3.2 3 Sample 2 washed with bidistilled water 4 Sample 1 immersed in 10- •* h Na2 CrC>4 at pH 3.2 5 Sample 4 washed with bidistilled water The precipitates were filtered and washed with double disti]led water and dried at 80°C. The specimen were pressed in KBr disks. The infrared spectra was recorded using Nicolet DX5 Spectrophotometer. RESULTS AND DISCUSSION: The radiotracer adsorption study shows that the maximum uptake in the present system occurs in the pH range 3-6 and sorption is postulated to take place by specific anion adsorption1- • 2 . The findings of IR studies are presented in Table-1. The effe<:t. of the adsorbed molecules are seen in terms of (i) frequency shifts in the bands of certain functional groups present on the surface (ii) new bands due to adsorbed molecules*. Antimony trioxide shows characteristic peaks5"7 at 870 cm" 1 and 742 i'trr1 which correspond to the surface hydroxyl groups (=Sb-OH). When chromate ions interact with antimony trioxide surface, its characteristic peaks appear to shift to 652 cm l and 694 cm'1 AR 28.1 respectively due to the formation of Sb-O-Ci >'• • species. The decrease in the frequencies may be due to incret- _n bond length of Sb-Q-CrO3H as compared to the -Sb-OH. The adsorb ion ol chromate ions on the oxide surface is also confirmed by the appearance of new peak at 558 cm"1 which corresponds to Cr-<~> bond. The possible mechanism of sorption of chromate ions on antimony trioxide surface is proposed ns follows: Sb-OH H* #Sb-OH2 (1) HCrO4- £=i | + H2O (ii) =Sb-OH =Sb-OCrO3H H2O (iii) -Sb-OH "^Sb-O'~ Before adsorption After adsorption 742 cm-1, 870 cm"1 694 cm-1 , 852 cm-1 CSb-OH sf sb~°-\ _^° + CrO«2- p=i I Cr^ + H2O (iv) Sb-OH ^Sb-Cl-"" ^0 Table 1: Peaks 0.1M 0.1M 10~ 960 (S) 870(S) 852(S) 852(S) 858(S) 852(5) 894 (S) 742(S,B) 694(S) 702(S) 72O(S) 720(S) 400 (W) 606(M) 588(M) 594(M) 588(M) 588(M) 498(S,B) 498(S,B) 498(S,B) 498(S,B) 498 (£>B) 384(S) 384(S) 384(S) 384(S) 384(S) 558(S,B) 552(S,B) 546(S,B) 552(5^) m,W-Weak, B-Broad) SIS: 1. R.L. Parfitt, R.J. Atkinson and R. St. C. Smart, Poll. Sci. Soc.Amer, Proc., 39. 837 (1975). 2. S. Music, M. Ristic and M. Tonkovic, Z. Waaser Abwasser- Forsch, 19, 186 (1986). 3. K.K. Pandey, G. Prasad and V.N. Singh, J. Chf-m. Tech. Biotechnol., 34A, 367 (1984). 4. K.S. Narayan, "Chemical Application of Infrared Spct.roscopy", Ed. C.N.R. Rao, Academic Press, New York, (1963). 5. V.F. Bentley, L.D. Smithson and A.L. Rozek, "Infrared Spectra and Characteristic Frequencices" Interscience Publishers, New York (1968). 6. N.T. McDevitt and W.L.Baun, Spectrochim Acta, 20, 799 (1964). 7. F.A. Miller andC.H. Wilkins, J.Anal. Chem. , 24, 125') (1952). AR - 28.2 CONCENTRATION OK CUKuMATK IONS ON V2O:. - A RADIOINDICATOR SORPTION STUDY ^ JBHOTANI, A.K.MITRA, RAME5H KUMARI Department of Chemistry Indian Institute of Technology, Hauz Khas, New Delhi- 110016 (INDIA) SUMMARY: The radioindicator method has been used to concentrate Cr(Vl) ions from the aqueous solution. The study shows that under optimum conditions, 250 mg of V2O5 carries 0.005 mg of chromium. Further, the activity carried under the specified conditions can be recovered more or less quantitatively by leaching the carrier. KEY WORDS: V2O5. Cr(VI), Sorption I.INTRODUCTION: Effluents from industries like chrome tanning, chrc " Dlating, alloy preparation, wood preservation and corrosion ion introduce excess of chromium(VI) in the natural waters is toxic to human beings as well as to animals*> 2. The liqui I effluents from pressurized water reactors are reported to contnin 5lCr = 2 x 10"2 (Ci per year). In the past, various physical and chemical methods have been used for concentration and analytical determination of chromium alongwith other elements particularly at trace level*. Besides preconcentration, the disposal of radioactive waste in solutions of low as well as of high activity is also an important aspect from the point of view of nuclear hazard and environmental pollution control* to a certain exten t. Vanadium pentaoxide which is widely used as a catalyst and oxidising agent, has been used for the concentration of 5lCr species. An attempt has been made to arrive at the optimum working conditions for the method and factors which influence the primary adsorption as well &s desorption processes have been studied. TI.EXPERIMENTAL: All the chemicals used in experiment were of afj{ily!.in«l grade. The radioactive tracer in the form of NaaSiCrO^ in ilil. HC1 solution were ol !,ained from BARC, Trombay (India) and used as supplied. The activity was measured in a single channel anal y-:er(HP.5) from Nuclear Enterprises Ltd., U.K. Tn principle, the procedure consists of two stages: first, the radi< ti\ a tracer in dilute aqueous solution is carried under spec' I'j p*c i ty of adsorbents, carrying efficiency as a function of amount of carrier arid extraction of the adsorbed activity as a function of pH c>\' the eluarit. MI M'T.Uf/TS AND DISCUSSION: The relevant findings are presented in Tf.ible ] . The optimum conditions for maximum adsorption (—-^ 9$%) V.--«s t'een worked out at pH about 2.25. Using 4 successive eiulii'iis around 85% of the overall activity is desorbed with dil. AR - 29.1 NSW on at Pli •'--'8.0. It is found thnt under these conditions, the O. 25 gm of VoOs is capable of carrying 0.005 mg of Inactive chromium. In terms of radioactivity, radiochromate species prese nt in 12 ml aqueous solution are likely to be ca> i led in a single operation to the extent of 90% and these- amount work out- to be 0.5 curie. For larger volume containing the same amount of radioactive tracer, a repetition of the operation woujd be necessary. The low»»r limit of estimation of these specie.", is about 0.1 ppm. Thus, the method of determination is not directly applicable to .".olutioris of higher dilution and preiroui •<-nt t » t i < >u 1 .i quit.M Important. The procedure amrins to be quick mid quite simple i ri operation and is likely to find some applications in certain raciiochemical processes involving concentrat ion and disposal of these radioactive species at low lnv^l. The adsorption of tracer chromate ions has also been carried out in presence of 10 2M concentration of various cMion.i a3 we]] as an i uns . 11 is observed that, the cations e.g. F'b?-l , Mg2 * , Zn2 * , Ni 2 • , Pe3*, Cu2 • , Bi3 * , * , Mn2' improve the adsorption significantly whereas certain other cations e.g. Ba>j * , Sr2 * , Cd2», Ca2+, Co2*, Crr+,A13+, Ag* do r:oc ?>'Ow any change in adsorption percentage. Further, in presence of certain c 'inp] «-x i rig anions e.g. N03 - , Cl" , Bv , , 1O3-, GOs?-- , CNS , F- , oxalate and tartarate ion3 the sorptlon is suppressed insignl1 icantly, 3 whereas an ions e.g. AsO4 -, Mo<)* 2 , i'u4 3 , :>z0;-)2-, oltrnt.) retard the carrying conslderably.Based upon the above study, in presence of most of the added ions, the quantitative s«p«tal i<>n of chromium appears to be feasible provided owirentration uf these species is not very high. Table-1 ADSORPTION nrc:".OKI"]'I ON Pll ( 5 pil ( % ) 0.95 64.2 1. 1 2.3 1 .35 76.0 ] .6 3. 1 1.9 87. 2 2. 1 7.9 2.05 93. 1 2.6 9.4 2. 15 94.7 3.0 12. 1 2.25 95. 1 3.5 2 5. 1 2.58 87.3 4. 1 50.9 3.05 73.6 4. 5 66.0 3.65 62. 1 5.0 78.0 3.85 50. 1 5. 4 84.0 4. 10 27.3 r>. 8 87.0 5. 10 18.3 6. 3 89.0 6. 15 12.3 7.0 91 .0 IV RKKKRENCES: 1 . MA. licnui t, Water Res. , 10, 497 (3976) 2. T. Maruyama A.H. Si dney and J.M. Coho.i, J. Water Poll. C'OJ-II ro] . Fed . , 47, 962 (197b). 3 ft. Wnnni nen "The Determination of Tracy Metal.-i i t, Natural 1 WH I.B rs " , lid . T.G. WOJS .. arul II.W. Nurnb«r«, Hloi:kw«l ! :;.j i in t.i f 1 ..: I'ubl ioations, London, 10 20 (jyfJH). 4. U.K.James and R.J. Bartiett, J. Environ. Qua).,\2, 169 (]9«3) AR - 29.2 DBSGRPTIOiU OF I" I Olid CH£kI30xi.UKD ON KOLYBDEUTUM ti .A.Daniela and (1-iiss) P.G .ite Department of Chemistry, Universi :,;,• of Poona, G-aneshkhind, Pune - 411 <-'"7 Desorption of 1 deposits on. molybdenum was studied at tempe<- ratures between 3A-8K and A1'5Y- in air. Iodide desorption followed the first order rate law. The rate constants were determined for desor— ption of I" films with coverage, © = 0.25/ 1.0, 1.5 ana 2.0. INTRODUCTION - 1/esorption Ji' I~ ions from silver^1' in argoa and air under different pressures was investigated ana it was. shown that the desorption results in decomposition, of Agl because silver halidea are photosensitive even at room temperature. I~ uesorption from- copper^ ^occurs in a molecular form (Cul) in the temperature raii^e 373-675K. .Both the processes of desorption 'follow the first order •kinetics. ^iuultaneou-s desorption in substantial amounts can influence the kinetics of surface as well as VOIULIS diffusion as both are coverage dependent. BXPEKIMENTAl --l-iolybdenum (99«9> purity, polished, supplied by M/s U-oodf ellow-iietals ,• Ln^land) metal used in the present study of deaorption was-in the-form of-a ribbon (breadth=0.64cin; thickness= Q.12mm)« 2cra long strips of the jietal were used, i'hese strips were subjected to chemical treatment in a solution containing 15ml perchloric acid and 5ml HG1{1;1) for'2 aec.'i'he strips were thorou- ghly washed thereafter with distilled water and dried. . , Per deposition of deaired 1~ coverages (0 = 0.25, 1*0, 1.5 and 2.0) the strips were immersed in 10~ 1-1 unl solution for restricted time intervals (coverage corresponds to the ratio of number of I" ions adsorbed to the atom population on molybdenum surface, viz., .1.08x10 atoms per cm2 x area of strip). The ual solution was "labelled with 1-131 isotope, provided by the Bhabha Atomic xiesearch Centre. These iodide covered molybdenum strips were then subjected to heat treatment'at constant temperatures in"the range 348K to 475K in a muffle furnace, i.esidual ft activity on the strips upon I" desorption vaa recorded at fixed time intervals usin^ a G~ii counter till the measured activity reuained almost constant. Contribution of the p activity towards the other side of the strip was accounted for» ileat treateu strips were washed and dried before recording their activity. RESULTS and DISCUSSION - I" ions form chemisorptive bond with the substrate molybdenum,hence its desorption will involve evaporation presumably in the form of_14ol2 molecule which is relatively more stable than other molybdenum iodides (Hoi decomposes above 823i£) . The I" desorption in a given time, t, corresponds to the difference in measured activities, viz., (ao-a+) whe-:. a<> denotes the initial activity and a,r that after • jsorption after"time t. The desorption followed thf first order rate law with respect to the concentration of tracer I" ions in accordance With the"relation AR - 30.1 2.303 a. v;er e Plots of log (a,/a0) vs. ^ " linear- The kinetic plots displayed two straight lines with different slopes representing a two-sta^e desorption.. The fir3t stage of desorption which la a fast one, imp- lies the removal of metal iodide molecules from the exposed surfa.ce while the second stage, that from the 3ub-s.urface of the substrate. The -second stage desorption involves diffusion of I~ ions . The desorption kinetics of this stage being' complex, only the first stage 13 taken into consideration in this work. Table 1: Kate oonstznt of desorption of I~ ions from molybdenum 1 1 5 t Temp./k V :1O"Vsec" kdx6x1O /sec~ 0=0.25 0=1 .0 e=i .5 2.0 e=o.25 0=1 .0 6=1 • 5 0=2 .0 1 ,; 16.0 1 .78 1 .54 1 47. 5 21 .34 16» 2 14. 1 ;73 21 .0 2. 5 1.93 1 .2 48. 2 21 .61 1". 8 14. 43 388 27 .0 3. 67 2/6 1 .8 48. 6 22.4 19. 4 14. 65 423 38.0 4. 83 4.05 2 .45 49. 6 23-5 20. 33 17. 11 458 58.1 6. 12 4.4 2 .9 59. 8 26.13 22. 9 Iti. a 473 83.2 7- 01 5.2 ';,• .51 67. 8 33-81 25. 23 21 .41 1 E/kJmol"" 2.4 1. 8 1.6 1 .23 It is observed from Table 1 thai; the rate constant of desorp- tion, k^, decreases with.increasing 0. ' xmilar*observation/was rep- orted'in the case of I" desorption on 3ilverv ' and copper* '. This is so because the desorption measurements are made in respect "of the saini geometric area of the strip though for the different coverages. It is obvious that the sublimation of the iodide film is. slower from the film with greater thickness. The corrected rate constant^,.©) appears to be independent of (^ except for the lowest coverage. Interestingly,(k,.0) value is greater for 6=0.25 than that at 9 = 1 (atleast by a factor of 2). This unusually fast sublimation from the lowest 0 is not clearly understood. The condition of sublimation flroiu the exposed surface at highehigner coverages is- different from that from the lowest 8 in which case the iodide is in direct conU .ct with the metal substrate . .The binding states of tli< iodide ion in the two cases therefore differ. The observed effect ma: be attributed to such different states of i binding. i''rom the data J' available on temperature effect on moiy bdenuiu fluorides ana chlorides, sublimation of molybdenum iodide film at the temperatures studied is well expected. The very low activation energy of desorption (Table 1) also supports the view of molecular desorption - 1) i; .A.Daniels , Ph.D. thesis, University of I-oona (1967)- 2) Ji.AoDaniela and P.G.Keddi , Proceedings of P.adiocheudutry and Al-59. itadiation Chemistry yymposiuifi, Nagpur, DAJbi (1990) Inc 3) R-CWeast, 'Handbook of Chemistry and Physic a1 , Baca Raton, 59th Edition, Florida 33431 (1978-79). AR - 30.2 IODINE - 1ODATE ISOTOPIC EXCHANGE REACTION STUDIED BY RADIOTRACER TECHNIQUE K.D. Ram Department of Chemistry, NERIST, Ranagar - 791110. India • md R. Tripathi Department of Chemistry, Uanaras Hindu University Vin-anasi - 221005, India SUMMARY The; isolopic exchango redaction of iodine between molecular iodine and iO'!;UR i::ns in neutral aqueous solutions (PH: 7.0 * 0.01'.) obey the rate Iji'. :• .. k f ',, ll2 i TO" j'Mif 175°C. The rate constant (k) was found in !••• .infi'i.s;i11u :tt nf Uio 'tola! ionic strength of the reaction mixture in presence o! K.i.i a ;..: KW(.),. Tho activation energy wns found to be 06 ± 3k.l/iiiuL Activation parameters were also calculated, ('n the basis of these roi.uil, an association-dissociation typo reaction mechanism is proposed for Uiis exchange ri;ai:tion in neutral aqueous solution. t-'.V w-iiu^. }ix<:):;>nm: i < :ir.l i<•n/idLiine/LoditUr.i/UiuUolrncer ttiL'hnique. ) i. i-'.'.iiLnjiicnin; Tho stuuy !i; isiinpa; •M.ii.iiiL;:; of iodine, b< Lwomi differen! oxidation states of iodine in I'qmvms solutions, molten salts and solid stiito, was planned to understand Iho role of isotopic exchange in the retention phenomenon fr>) low ing (n,"V" } procosf; in iodates and per-iodalos. Some of our results have already been ropi-i-i-'i /I-.)/. Here we present our results on iho iotiine - iodali! isuloph; c .di,nii|i! in neutral aqueous sohitinir; n- '.,•• ! l.i) an raciiotracer. II. EX PliKl MENTAL All the chuidicais used were of'AK grade. The iodine solution was ;;repared by dissolving iodine in KI solution. It was labelled with carrier-free 1-131 obtained from BARC, Bombay, India. The specific activity of iodine solution was determined in each case. The experimental procedure is described elsewhere /!/ in detail. The exchanging species were quantitatively analysed by radio paper cbromfitography. III. RESULTS AND DISCUSSION The exchange rate (R) was computed using appropriate form of McKay equation /A/. Tho order of the reaction with respect to iodine and iodate were determined from the linear plots of log R vs. log [I,,] and log R vs. log AR - 31 .1 [ 10" J, respectively, and were found to be 1.2 ± 0.05 and 0. 4 ± 0.05. Thus^the rite expression could be represented as: 1.2 -,0.4 R = k [I. 10 The exchange rate is found to increase in presence of KC1 and KNO.,. However, the rate constant (k) was independent of the total ionic strength of the reaction mixture in presence of KC1 and KNO,, signifying the involvement of at least one neutral species in the rate netermining step. Activation energy, obtained from the Arhenius plot, is found to be 86 ± 3 kJ/mnl. The values of free energy of activation, enthalpy of activation and entropy of activation, obtained on tho basis of transition state theory, are found to be 136 ± 1 kJ/mol, 79±3 k.I/inol and 13()±3 J/deg-mol, respectively. The exchange mechanism based on the redox reactions, proposed by earlier workers/5/, takes place exclusively in acidic media involving the species of intermediate? oxidation states. On the contrary, the experimental conditions in the present investigation in neutral aqueous solutions do not permit redox reaction to proceed. The rale law suggests a complex mechanism and tho fractional order might be duo to I . species interacting with 10., bionuil«cul- arly. ., The proposed mechanism and plausible structure of the transition stale complex is given below: 1 2- I o The mechanism envisages the formation of an asociative intermediate transition state complex resulting in a temporal y increase in the coordination number of iodine (V). Since iodine in iodate ion has a positive charge (<• 0.0532) /6/, the I., preferably adds to iodine (V). In the transition state a tempo- rary bond' formation between oxygens of iodate ion and one of the iodine of I involving vacant d- orbitals of iodine and filled p- orbitals of oxygon, may be speculated. The shifting of electrons towards central iodine (V) and simultaneous transfer of oxygen towards outer iodine within tliu transition state complex leads to the formation of exchanged products. The catalytic effect of neutral ionic salts ma\. lie due to electrostatic inter- action of Cl and NO., ions with iodine (V) within the transition state complex resulting in nearly tetrahedral arrangements of bonds. IV. REFERENCES 1. K.D. Ham and H. Tripathi, Appl. Radial. Isot. 40,505 (19H9). 2. K.I). Ham and H. Tripathi Ibid., In press 3. K.D. Ham and K. Tripathi, .1. Radioanal. Cheni. Articles. I)!) (2). 347 (1990). 4. 11. A.C. Mckay, Nature, 142, 997 (19311). 5. O.K. Myers and .I.W. Kennedy. .I.Am. Chum. Sou.. 72. H97 (19!i()|. G. b.l). lil-hlssa and A. Hinchliffo, .). Mol. Struct. 67, 317 (19)1(1). AR - 31 .2 SORPTION OF MANGANESE AND Clll>.<'HI MM i.,N INDIAN :;.<'IF.:-. 2.A. JiilAM arid SATYA i-KAT Pr'>i>sJ Engg. & Gy.^tem.i f1 iv i -- i- •!) Bhabha At-mic [fesem-.-h C.'-ntrr-. Tvomb--". . (•• ml •••;.- -hVi i1l '.'• )]j'! i ii ,.,-'1 mangel 11-•::.••=• .nid •], romt xn\ • rr<;satioe of iiiiin^anese and clirominni in shallow aquifers of. ly-jny places h?.s been reported by several investigators1 ^ . Tiiese iunic si-^.-oies present in the wastes cf certain industries interact wi tli the subsurface formation cin^i ->re a part of recharge water of the area concerned. We therefi.Tc studied the svrption behaviour of tb^se ionic species on tw>-> Indian soils. JI. MATERIALS AND METHODS : Silty l-am soli from the iJangetiu Doab and ijilty clay loam soil from coastal Maharashtra wer^ used in this study. Physico--chemical and mineraJogical characteristicer s of k tlies-5 S'-ils were ^etermi^ned a»,id arer reported <-ls<-vdiere'- ' . Five solutions of 10 ", 10 -', JO ''. 10 °, lu 'M .:•• .i.cerutr :< f,i. >n -..f nese v;ere prepared by dissolving A.k. gr*ide Klin1.')^ *- "tin in distiiLed water. I.'iiiu Lari ly five solutions -.if ohr< .-mi uin wei-= piepar.jd by di..'..-" •Ivin^ h.k. groA>s K \i ..'> , ' ' 'Cr. Thv til of aj 1 s. .In !: ions . wa.:. adjusted t •.• v-^ 7 . i.-HA. " C-.i t •;.-li tests as d'_:scr i h«--d eai'i i -3)'1 J ' wer-r carried ..-ijt. at rC'Ciu I.'.•inp'j tat ur.t . ! - • .> 11 <_-1-111 1 • > L i • -n •_• i nr.iiigaii-;!:.- cns-j -.!ir. •iui uiu in t Ii- fr-rrJ -ind pi'-dn t pliv-tsrrs w-.-1 •_ d-rt.ei ..in'-d Ly i:i>. "i iii r i iiii '' fin nirA '~'*~Ci ;. • i i vi t i •-£. in .:i Mult i •; lianu-:; 1 Analy ,r-j- and rjorptj on of th-i-^j o[i tbo .-:.!• i 1 .:• '.:;dcnLat.'-i|. ill! ! /.;•.•-..•;. nvit ion : 17W and !.M.-^uiidlich c-juat i on : S] - ^ X - - • " C ^ V.'ij-.re X is I-lie equilibrium c'- AR - 32.1 to th^ .v.-oonci and not to the f u 51 equation in dieat•=-.•? i.hat sorption (.if both managaiiese and • •hrorrii. uui "ii the two 5.-;pits i.°. governed by more than one mechanism i.e. 11;.- si-.i-ptiuu i.? due :.. • the me' -liaui sms of adsorption, ab-v-rpt i on -iii-J i-.n exchange, ail taking t11. t < •« simultaneous] y . The values .if r. ].<-»pe and int'-i'>['t. for si j ty clay loam for both niants'uie.'-.t-- and •.••hr.'-nii urn * r~ highrr than thor.e for silty Joam, indioat i ng that th^?.^ will be m. iv sorbed •'ii the former as compared to the latter. This is i-'s si LI *- due t<> mvi <--. organic matter, clay an.; in- ••ntiuori 1 1 oni te C"iitent. introdn-•• t i •'ij ii. g^ii'-r-i I ly livr/i ne; • i J t y J •:--i»ii f •.•uuat i •.•ut,, therefore, tlie industrial act. j vi t i .-.••. h-.i-.-.- j-.--.-..u J1 •-..) in the i r--..r-uce •.•f thes-- t...;ic.:iiit5 in shallow .-tqui f'_-rj.. T)ii.=. t y[-i .'f •iLt-i.iati-.-u is f^{-:Ct rrd !e.S:5 i )i S> i i t y olay f < I ii).":i t. i . n .s . I. V.p. Kudesia, Aota Cienoia Indi-a V..-1. Vll]C II.-.:: r I Hf-T: i , 11 y 1 :;0. ;i. A.K. iJahgaJ and V.p. Kudesin, Aot.a Ciencia Indica, Vol . IXC , No . 3 , (1 982 ;,• 1 5t. IW . •3. Mohaimned Ajmal anil fiaz;i uddin, Enviroiiiuftitoil Monitoring and Assessment, Vol.V, No. 2, (IHCGj, 1>.J1 194. 4. B.X. Handa, I'roo. National Seminar on Env. Pol 1. contract and Monitoring, Chandigarh (llHu1.;, 47M 492. b. Z.A. Khan, Ph.D. Thesis, Bombay Uni., (1989), 88 104. G. Z.A. Khan, Satya Brat and B.M. Misra, J. Indian lust. oci. (I'JSS), 411-418. V. K. W. T. G>,-uldinfef and u . Ta 1 ihinjecn . J.S...il. lj..-i . ( 1 9ti4 .1 , 3y7 408. B. V. Tai-^ and CD. H-jkil, J. Knvir-.-n. Uurt. il^fclC:), o^f. Gk1^. 9. 11. G. Iir'..wmiMii and E.<. I'. H-p-s I ding , •) . En v i . wu11.. ( 1984 ) , lOo 172. tO. H.L. L-ahn, ii.h. M-.:N'-.:.J and G. A. *'''>: 'juior 'Soil Ch^nu stry' , Wiley Inter :..V i *.-nce , 1 U'Y'.J ,1 . 117 XhklAL U I''."I-! i.'l1' HArMAliE:":!.1; /\!U> '.'IlI.cHHII! VIN OV'IL::- des -.'f I, I"i«~ da t.0:i u.-^j lifi l-'i'i nn-J I i-.-.li c.-..n.Kit ] • -n Fi ,-il i h t'.-i.s. i • -n f-? •Hot an t Hi. l-n 1 y C ^;3i i ty oiay luai y 14 1 llj. \i\ IJi 1 t.y l'. AR - 32.2 FLOW RATE MEASUREMENT IN A CRYSTALL IZER DRAFT TUBE BY MEANS OF A NEUTRALLY BUOYANT SEALED RADIOACTIVE FLOW FOLLOWER H.J.Pant Isotope Division, B.A.R.C. Trombay Bombav-tOO 085 SUMMAHY This paper describes flow rate measurements maoe in a pilot "lant for crystal 1 izer to characterise' the per f oimarico or two different types of propellers using ? ne1.:f. rs I 1 y buovant rsdioaci ivc i low follower. P«si'lts i n^H KEY WORDS Flow rate. Red i ot racer . or vs ta 1 i i z ei . INTRO 0 ilCT_LP N M i 3 c i b I e rsdiotracers are extensiv-ly used to measure i • w • te in industrial process systems. But tht- use of inisc i : • i *= radiotrsc^rs is 1 imi ted i/i s<. DESCRIPTION OF SYSTEM The Crystal 1 izer consists or a vertical draft tube ot diameter 115 mm and 1830 mm height enclosed in a vertical cylinder of diameter 355 mm. These two vessels are enclosed in < rnnicsl shaped outer vessel closed at the bottom end. A |.i^,H-i!^i diametor 102 mm is fitted at the bottom end of the •: j -if t t ,,... The system is filled with water and made leak proof. ..." ,station of propel!er causes water to circulate in the system. NEUTRALLY BUOYANT RADIOACTIVE FLOU FOLLOWER The neutrally buoyant radioactive flow follower con-- i •: t s or a polypropylene bead (Specific gravity = • •. !• C> ) •••< d i .. -.-r f.mm, having a small opening through it and a small •, -.; i od.o >. i v e cobalt-60 pellet < act i v i t y - 37MBq ) of weight 0.006" g. Tr,e radioactive cobalt pellet was encapsulated in the bead l> ,• «:•??• 1 i ng from both sides with araldite epoxy. The radioactive- (••••ad was tested for neutral buoyancy by releasing in a wj'.^r- t i I led cylinder. The density of the radioactive bead was mat^Ktd with the density of water by addition or removal of aralditt? epoxy so that it followed the flow in the system reasonably truthfully./!/ MEASUREMENTS AND RESULTS Two coSlimated de.tectors separted by a known distance id=153.5 cm) were placed outside the system and the outputs of both the detectors were fed to a microprocessor-b?sed data acquisition system connected to a printer set to record 1000 events with an interval of 200 msec. The count rate responses of both the detectors were plotted and the time eiapsed between he two radiation pulses recorded by the detectors was found out./2/ Representative response of the detectors are shown in fig-1. The AR - 33.1 fiowrate is calculated multiplying the linear velocity by crsssectiona1 area (A=11.5 cm'2) of the draft tube. The Reynold number of fluid flow was estimated to be of the order of 10'5, which indicated high turbulent flow having a flatter velocity profile. The flow rates in the draft tube were measured at different operating conditions such as coverage over diaft tube and RPM ot the propeller. The results obtained are given in table-1. CONCLUSIONS Results indicate that L-type propeller develops higher flow rate than R-.type propeller and hone© could be used in industrial process vessels where higher flow rates are required to be gpnerated. REFERENCES l.J.C.Middleton, '3rd Europen Conference on mixing' 1CI Corporate Laboraatory U.K (1979). 2.H.Cramer et al. Cheat. Eng. Sci. Vol-2. PP 5-42 (1953 TABLE-1 L-Type Propl !er R-Type Propl 1er R.No. RPM < t S.E Q R.NO RPM C t S.E Q 1. 500 3 4 3 .8 16. 0 4. 5 1 500 3.4 4 .8 6. 0 3.3 2. 500 6 9 2 .6 4. 4 6. 1 2 500 6.9 4 .5 9. 0 3.6 3. 500 10 3 2. 3 2. 5 6. 9 3 500 10.3 4 . 1 11. 9 3. 9 4. 600 3 4 1.8 0.0 8.8 4 600 3.4 3.8 6. 6 4. 1 5. 600 6 9 1.8 3.2 9.0 5 600 6.9 3. 4 4. 8 4. 7 6. 6OO 10. 3 1. 7 6. 0 9. 1 6 600 10.3 2 .8 6. 0 5.6 7. 750 3. 4 1.6 5.0 9.8 7 750 3.4 3 .0 15. 7 5. 3 8. 750 10. 3 1.5 2. 7 10. 0 8 750 6. 9 2 .8 16. 0 5.7 9 750 10.3 2 .6 13. 7 6.0 Where RPM : Revolution per minute, C : Coverage over draft tube in cm. t : Mean trani'c time in sec, Q : Mean flow rate in liter per sec. S.E : Standard Error. 90 100 110 120 130 1(0 1S0 160 170 1M Tim« (MC.) —•- A REPRESENTATIVE COUNTS/MINUTE VS 7!ME PI OT Boilom Detector (D, ) Top Detector (0, ) Fig 1 AR - 33.2 RADIOTRACER"STUDY OF SOME n- S.P.Mishra and V. Upadhyaya* Chemistry Department, B.H.U. Varanasi-221005 ^Chemistry Department, S.C.College, Ballia-277001 INDIA SUMMARY: The radiotracer results on some strong n-o" complexes indicate that the "specific" donor-acceptor interactions are predominating in such complexes. Key Words: Radiotracer Technique, Molecular Complexes,Charge-Transfer Complexes I. INTRODUCTION: Our radiotracer studies in the field of molecular complexes /I,2/ clearly indicate that the experimentally deduced equilibrium constant values for some weak 7C - II. EXPERIMENTAL: Labelled tetramettiyl ammonium pentaiodide was synthesized/2/ using carrier free radioiodine (I ) which was equilibrated in pure and donor containing solvents in a thermostat. All the donors and the solvent were purified by a procedure already described in literature/3/. III. RESULTS AND DISCUSSION: The equilibrium constant values obtained using the radiotracer technique, by the procedure already described/2/ along with the heats of formation (- £± H°) and entropy changes (- ^S ) for tne complex formation have been returned in table I. Tt is observed that the K values increase in the order of the electron donating ability of the donors,, Replace- ment of the hydrogen atoms by the alkyl groups increases the electron density at nitrogen' atom and consequently the electron donating ability of the donor molecule. Unlike the weak complexes the aliphatic amine-Ij complexes yield equilibrium constant values which are quiet comparable with those obtained by the conventional procedure (Table I). The conventional spectral method accounts for only "specific" donor-acceptor contacts occurring in excess of random collisions/4/ while the radiotracer technique includes both the "specific" and "non specific" interaction contributions/2/. Since the aliphatic amines are the most powerful electron donors their complexes with iodine yield very high value of equilibrium constants (ranging from ca 500 to 3500 jdm'mol )(Table I) in comparison K values for "31 -o" complexes (^1.0 dm'mol )/2/. It is therefore reasonable tS assume that the "specific" solvent effects comprising of comple- xation, hydrogen bonding, polarization/4/ are predominating in the strong molecular complexes and the contributions of the "non specific" interaction including solvent cage strain, dispersion etc. are negligible in their comparison. AR - 34.1 IV. REFERENCES: 1. S.P.Mishra, R.A.Singh and V.Upadhyaya, J. Radioanal. Nucl. Chem. Letters 96, 5, 481-88 (1985) 2. S.P.Mishra, R.A.Singh and V. Upadhyaya, Radiochim. Acta, 39, 205-210 (1986) 3. 'Techniques of Chemistry' Vol. IV, Edited by A. Weissberger, Wiley, New York (1970) 4. R.II.Lane, S.D.Christian and J.D.Childs, J. Am. Chem. Soc. 96, 38 (1974) Table I Thermodynamic data obtained for aliphatic amine-iodine complexes in n-heptane K (dm3 mol ) - AH° Donor c (K cal mol ) (fc. u.) 25°C 30°C 35°C 40°C ri-propylamine 743 598 486 395 7.80 13 .80 n.-Butylamine 980 775. 621 498 8.33 14.50 (770)a (8.4)a (14 • 8)a Diethylainine 3507 2684 2072 1575 9.60 17.00 a (9.7) (18 • 4)a Diisopropylamine 3774 2878 2173 1668 9.73 17 .90 Triethylamine 4449 3205 2326 1712 11.80 21 .20 (4690)b (3310)'' (12.0)b (23, • 5)b a. If.Yada, J. Tanaka and S. Nagakura, Bull. Chem. Soc. Japan 33, 1660(1960) b. S. Nagakura, J. Am. Chem. Soc., 80, 520 (1958) AR - 34.2 HOKOGct.EOUS 1SUTCP1C EXCHANGE BETWEEN NICKEL (il} AND BIS(RESACETOPHErJCt;E PHENY Li iYLRAZONE) NICKEL (il) COr.KLEX T. GAKGAIAH, P. RAHALEVI and (J.R.K._Jj DEPARTflE:JT CF CULT,! STRY, C\ LLEGE OF Et!GI NEERING SKI VENKATESWARA UNIVERSITY TIRI'PATI - 517 5'')2, INDIA. SUMiMARY * Isotopic exchange behaviour of hi s (resacetophenon** ph^ny lhydrazone) nickel (II) complex with nickel (II) in trl-n-butyl phosphate and methanol medium was studied. The studies were carried out at different tempera- tures varying the concentration of both metal ion and the complex. The results show that the complex is labile in th^ kinetic sense. Increase in temperature increasppthe isotopic exchange rate. 'lhe increase in concentration also results in enhancement of the rate of reaction. Key words: isot.opic exchange, labile, resecet ophenone pheny lhydra zone, nickel (II), Tri-n-hutyl pho:-phnte, metha nol. pyridi ne . INTRODUCTION s PhenPhenyy l hydrhydr a zonezss havhvee been i«y<: EXPERIMENTAL i Reagents Resrtcetophenune pheny1 hydra zone was prepared according to the method. Tti-n-butyl phosphate, methanol, pyridine, nickel (II) chloride were used. ° Ni tracer supplied by Board of Radiation and Isotope Technology, ElAE, Bombay was used. 63Ni was received as nickel (II) chloride in hydrochloric acid medium. The solution was evaporated to dryness and taken into rnethanol. Bis(resacetophenone phenylhydra^one) nickel (II) complex was prepared according to the method described in literaturev*/ Solutions of nickel (il) chloride (1.0 x 10~2M) in methanol and nickel complex (1.0 x 1C~2M) in tri-n-butyl phosphate were prepared. Reaction mixture was prepared by mixing 15 ml of nickel (II) chloride and 15 ml of nickel complex and 0.2 ml of pyridine. The tempeiature of the reaction mixture was maintained at 25 + 0.1°C and known amount of 63Ni tracer was ad Bis(resacetophenone phenylhydrazone) nickel (II) complex is inert towards nickel (II) exchange in the absence of base as well as in the presence of aniline. Hence the study was carried out in the presence of pyridine which is more basic compared to aniline. Results indicate that the complex is labile towards substitution in the presence of pyridine. It may be due to the formation of the adduct with the complex which renders the complex labile. The exchange behaviour studied at different temperatures keeping the concentration of the reactants constant, shows that the rate of exchange increases as the temperature of the system is raised. When the concentration of the reactants is enhanced by five times there is nearly four and half times increase in the rate of exchange. The effect of concentration is significant compared to temperature. The rates of exchange were evaluated and the activation energy determined using the Arrhenius equation. The activation energy of the reaction was determined from the slope of the graph log R (reaction rate) vs. 1/T is 8.40 KJ/mol. REFERENCES 1 M.W. Moon, Ger. Offen., 2, 058, 156 U.S. Appl., 93, 27, 496. 2 M.W. Moon, E.G. Gemrich and G. Kaugards, J.Agr. Food. Chem., 20,888(1972) 3 J.G. Pecca, J. Dobrecky and S.M.Albonico, J.Pharm.Sci., 60, 650 (1971). 4 S.Srihari and N.A. Raju, J. Indian Inst. Sci., 64(2), 45 (1983). 5 P. Umapathy and N.A. Raju, Curri.Sci., 29, 428 (1960). AR - 35,2 A BINARY COCKTAIL FOR THE DETERniNATION OF THORIUM AND NATURAL URANIUM USING LIQUID SCINTILLATION COUNTING. R.V.Subba Rao.A.G.Rafi Ahmed,N^S Ji.Singh and G.R.Balaaubramanian. Reprocessing Programme , IGCAR , Kal p^kkain • 60 3 102. ABSTRACT The present scinti1lator cocktail avoids the use of secondary solute.The efficiency obtained by this (1) scintillator was better than PPO/POPOP .The ef ficiency calculated wa3 82% for natural Uranium and 104% for Thor iuro. For the natural Uranium the RSD at 2 me level WAS 2.16% and at 8 mi: was 1.17%. For Thorium the RSDs at 2 mg and 8 ng were 2.64% and 1.94% respectively. INTRODUCTION The conventional liquid scinti1lator consists of primary solute whose emission response will not match with the spectral response of photo multiplier tube,thereby necessitating the addition of secondary solute which will also act as quencher.The present method eliminated the use of secondary solute and is very simple to be udpoted in any condition.The nuclides were extracted by the cocktail itaelf,which contains 30'. TBP( Dod ecane) as extractant. EXPERIMENTAL Urany] r, i trat e 100 gpl in 4N Nitric Acid Thorium nitrate 100 gpl in IN Nitric Acid 30 % TBP (Dodecane) 5 0 mlper liter of Xylene BBOT 0.7 gpl of Xylene ECIL 2 0 was used for counting purpose in aiit i - coincidence mode. PROCEDURE Suitable aliquot of Tli/U vaa put into liquid ai'hil: i llation vial. To t.h* vial 10 ml of cocktail wati added.The mixture was equilibrated for 5 inititues. After equilibration,the mixture was allowed to stand for 20 ini n tu es . The aijueoui phase was not seperated from the organic.The vial was ki'pl asj aucli in the sample chamber and counted.The back ground corrector! was applied. A caliberation was carried out to test the proportionality with change in the concentrai: ion of U/Th. AR - 36.1 RESULTS AND DISSCUSSION. After detailed parametric studies,the optimum parameters wet e f i xed( ( i a.. i ) This method «ives excellent results for the determination of natural Uranium and Thorium in the absrencp of other Alpha emitting nuc1ides.This method can be directly applied to the aqueoua as well a« organic nanip 1 es . The efficiency calculated for natural Uranium and Thorium wore 82'. and 104%(Higher efficiency for Thorium was due to bulid up of datrthters) reapt'cti veljr.Thn rproinmended range was 1-10 mgi for both tuitutal Uranium and Throuiin. REFERENCES 1 . Dotirt ] d , \, . Horiocks , Chi n . Tzu - Peng , Organ i c Scintil lators and t. icjuid Sc- i n t i J ) A t i on Cuun t i up . Acadam i c. press, New Y<;i. k .1971. AR - 36.2 Effect ol BBOT Ettectof TBP(30°A>) UO0 I»00 2 i « * 10 0-3 OX 04 G-t 1-0 Concantration (my) Volof TBP (ml) .— Jh —• - u ^Th. , U Optimum value 7 mg. Optimum value 0-5 ml. Calibration graph o o •-* O c e 10 of Th./U (mg) U __^-Th. AN - 36.3 EXTRACTION OF IRON FROM LUBRICATING OIL M.P. Chacharkar and Mrs. B.B. Tak Raksha Prayogshala, Jodhpur 342001, India. SUMMARY The extraction of iron into aqueous medium from the used lubricating oil of aeroengines due to wear and tear was studied, having dissolved in hydrochloric and nitric acid, using iron-59. Different variables, e.g., complexation with EDTA and thiocynate; iron and acid concentrations; diluent addition were investigated. The extractiion was quantitative distribution. The data was used for development of the 'metal wear moni tor'. KEY WORDS: Iron, wear debris, acid ex tract ion,lubricating oil I.INTRODUCTION The determination of wear and tear is quite important from the point of effective maintenance and operational confidence of the machines. Wear debris due to oil dipped component is combination of mechanical wear (metal particles), oxidative corrosion (metal oxides) and chemical corrosion (dissolved or organometallic compounds) /I/. The estimation of metal wear is dependent on the lubricating oil and particle size and shows variation when analysed with the different atomic absorption / .emission spectrometric techniques /1-2/. The spectrometric oil analysis by using particle size independent method /3-4/ also reflects the influence of above parameters.The dissolution of metal from wear debris, extraction in aqueous media and the development of specific colour with suitable reagent can provide quantitative results and enlarge the scope of its application. This paper presents some of the data on extraction of iron from used lubricating oil from the aircraft engines. II. EXPERIMENTAL All the chemicals used were of analytical reagent grade. Used lubricating oil was generally used in the present studies. In some cases, unused lubricating oil together with standardised ferric chloride solution was employed. Fe-59, obtained from BRIT, Bombay, served for radiometric assay. 5 ml oil + 5 ml 50% HC1 + 2 drops 5% nitric acid + 1 ml Fe-59 soln. were shaken for 5 minutes. Further 20 ml each of filtered kerosene and water were added to the above, and shaken again for 5 minutes before the layers were kept undisturbed for half an hour for seperation of the phases. The deviation from the said procedure and treatment of the aqueous phase was carried out as per need. Fe-59 activity in either layers was determined via gamma (1.29 Mev ) counting, AR - 37.1 III. RESULTS AND DISCUSSION The following table shows the results of above studies. Experimental seperation condition % Extraction in aq. phase As Fe (III)- chloride* 99.1 As Fe (III)- EDTA* 99.6 As Fe (III)- thiocynate* 99.7 Fe (III) variable(50-250 ug)* 99.5 - 99.9 HC1: (0.025-1.0 m)** 99.8 - 99.9 Kerosene 2-20 ml/5ml oil** 99.6 » 99.9 Kerosene addition start/last* 99.3 - 99.5 * Fixed acid concentration 0.4 m ** Fixed Fe(III)-250ug and acid concentration 0.4 m The experimental accuracy depends on the various physical and chemical factors apart from counting stastics. In the present investigations, that involves syrupy dark organic p!iase, the absolute observational error that was found was in the limits of 1%. Therefore, it can be concluded that iron that is present as a consequence of the various processes of wear and tear in the used lubricatig oil medium of aircraft can be extracted in the aqueous acidic medium quantitatively. In spite of large number of chemical species of iron (III) and chloride /5/ present in the aqueous hydrochloric acid medium and the extraction of these species in organic phase can be achieved in higher acid molarities or chloride concentration, and the extracted species can be easily stripped off in the very dilute acidic aqueous phase /6/. The experiments have clearly indicated in accordance with the literature that none of the Fe(III)-chloride species have contributed significantly in the back extraction of these entities in the lub.-oil+kerosene medium, or adsorbed on suspended particulate matter,or exchanged with any entity of the total organic phase. This data having supplemented with the spectrophotomeric studies has helped us in devising a suitable kit /!/ for estimation of the iron in lubricating oil system. IV. ACKNOWLEDGEMENTS The authors are grateful to Sri JV Ramana Rao, Director and Dr. AR Reddy, Jt. Director of our Institute for the encouragement. V. REFERENCES 1. K.J. Eisentraut, R.W. Newman, C.S. Saba, R.E. Kaufman and W.E. Rhine, Anal. Chem., 56, 1086A (1984). 2. C.S. Saba, W.E. Rhine and K.J. Eisentraut, Anal. Chem., 53, 1099 (1981). 3. J. R. Brown, C.S. Saba, W.E. Rhine and K.J. Eisentraut, Anal. Chem., 52, 2395 (1980). 4. C.S. Saba and K.J. Eisentraut, Anal. Chem., 53, 1927 (1979). 5. D.E. Chalky and H.J.P. Williams, J. Chem. Soc, 1920 (1952). 6. M.P. Chacharkar, Records Geol. Surv. India, 119, 61 (1989). 7. M.P. Chacharkar, N.K. Soni and Mrs. B.B. Tak, (Unpublisheu). AR - 37.2 HAO(OCtllM(CAI 01 TIHHtNAl (ON 01 Si I I CM I) IHACI II I Ml N I S IN HI010G1CAI MAT IH1AI S K . R . Krrshnamoorthy and I). K. Iyer Analytical Chemistry I) iv i sion , llhauha Atomic, keseaich Centre Hombay-400 08'j SUMMAKV:A radtochomica1 scheme fur the separation (if soven trace elements in biologies] matuvials add its application tu Uiruu quality control siimplcs i <; pit: merited in this paper. (keywords: bio log ical materials, traco elements, radiochemical vepaiatiori scheme) (NlltOOUClfON Determi na ti on of essential and toxic tract- el cine n I s in biological malei Uls, especially foodstuffs, has become I.his nued of the day. The stiiis i t.i vj IJ os fox many of the i.raci; elements are quite favourable for their determination by neutron activation analysis, but the matrix activity poses piobli'ms not only in handling but also in the computation of the peak areas fiom the gammu-ray spectra. A radiochemi,i:al sopaiation scheme h«<:omi;;> a nuces;: ily j n such (.uses. (his paper dosci Ibes su( ti a schume for I tit; determination of seven trace elements (As , Cd, Ct , Cu, llo,, So and /n) in t> I o 1 09 i cH I i>ia t (!I x a I s . IXI'I H (MINI Al. I xpe* imc'iil v. were init.tally cairied out usmy radioactive tracers and the correspond ing carrier:;, to evaluate the roctuvei.es. the yields wore in the ian$je 9^ to Hfl per cent. The flow-sheet, (l.itf. 1) yives the steps in the procedure. Iho procedure was then applied to two standard ruferencu materials, MA-A-I and MA-A-?. The results are given in Table 1. later, tlitee quality control samples from £Al A wure anaLysed by this procedure. I able 2 gives the results. Activities were measured using a 4'•<:<: lll'Ge detector coupled to a multichannel analyser. I he enuryi.es (in koV) chosen foi the evaluation of the peak areas: As-Y 6(60'/) ,C DISCUSSION The r(!sult.s show that the scheme! works sat i sfactor iJ y for the samples studied. I'hu main features of tit is procedure are I i.) separation of elements of low specific activity (Cr. (Id, Se) into separate groups and (it) use of pi oc i pi tat ion methods which aie fast and quantitative and (i i.i.) isolation of nuclides «>f mutually interfering gamma-rays (e.g. llg-?03 and Se-/0, 260 t-feV) in different groups. (n tho case of As-/6, the less sensitive K'il keV peak w«s chosen since Sb-1?? .iritefores at. 009 keV. Somewhat, differing values were obtained for Kg and Cd. this is due to tho problems of standard preparation for llg and the low specific activity of Cd-itO. I he procedure takes only ?. days from digestion to measurement. It is proposed to adopt the method for analysing food-stuffs. ACKNOWI lOGMlH) this work was carried out as apart of the Research Contract A3Y0/K'll of International Atomic Inorgy Agency. The authors are also thankful to Or. S. (liinciiidharan, Head, Analytical Chemistry Division tl. A. K. C. , liombay. * AR - 38.1 Wuighl of sample: 0.'Jg uai:h, 3 to 4 sample:: i n a sot.. Irradiation: 24 \\ (0 . 'J l.o 1*10 n t:w s ) Cooling period: 7 1 . Siifnjj_l_t; and tar; J era I) I si i 1 Jai u : I vapor a It* l.o dryness l)eiompose with UNO arid llflO I uai h with IIC1 , add to filtrate in llelhu.f's apparatus from Ci prei-ipi I at iori( s( ep ?). 7 Heiidue: Add HC1 to distil off 0J st i llatc: Crer.ipjtattJ HaCtO CiO Cl . fiI tor. Count for Cr-SI . • i 7 Combine fill/ale with step 4. J .KusiiJuo :A(J just, acidity to 4M IIC1. l'ri)i:ip i.ta te : Count for S»- I 'j . add Na SO soln . , filter. < LilUaiil-- adjust acidity In 1N KCi : Count, for As-Y6. add t ti i !i .J:i 1 tratu l-vaporatc to drynass. decompose amen, salts with UNO Co Cd-11!i and /n-6'j dilute, add Nil H l>0 . ad jus i: pll to 6, heat and filter. 6. I .iltiate: K«j«cl . l.ig.i: I .lowsheul for the Had.ioctitiniit.a.1 Separation Schomt; labl.o 1. Results of Analysis of MA-A-1 and MA-A-2 (wa lues in ug/9) l.:lORIOIll. MA-A-1 MA-A-2 mean s . (1. cm . value* mean Si .d. cor value* As B.9V 0. 43 (i. 1 2 .b3 (1 .24 2. 6 Cd 0.64 0 . 04 o. rs 0. W 0 .09 0. 0/ ( r (1.4 0 - - 1 . 1 1 . 2 -- 1. W Cu b.89 0. 36 / .6 3.18 0 .29 4 . 0 Iiy (1.19 - - (I. 28 0 . '• 'j (1 . 08 0. hi So 2 . !i J 1. 16 3.0 i. B:I 0 . 19 1 . 10 /•• 143 •j 1!,« 29. I !i . :>. .0 *from Y. Miiramalstt ami K.M.farr, IAlA/k'l /1 ?B (1981.) table 2. Mosult.s obtained for duality Control samples from IAI A (values in 119/9) I iwumtTil Sample 1 Sample ? Sample 3 ruf . luf . vo luis 1 of. vJJ no As 0.38 0. 41 6. ?r b . / 2. 66 7. .b Cd AR - 38.2 SUBSTOICHIOMETRIC ISOTOPE DILUTION ANALYSIS OF MERCURY USING POTASSIUM n-BUTYL XANTHATE P.D.SATPUTE and A.N.GARG Department of Chemistry, Nagpur University NABPUR-44OO1O(INDIA) SUMMARY- A substoichiometric isotope dilution metbjd has been developed for th« determination of Hg(I1) by solvent extraction employing n-butyl xanthate in chloroform from borax buffer of pH 8.2. A minimum amount of 5/jg Hg can be determined with a fair degree of accuracy. (Keywords: Substoichiometric isotope dilution analysis, Mercury-2O3 tracer, potassium n—butyl xanthate, solvent extraction) i.INTRODUCTION- Hg is known as a toxic element in biological and environmental systems. Solid (dust particulate), liquid (effluent) and gaseous (smoke etc) wastes from industries and automobile Are the major source of its entry into the biosphere/1-3/. Because of very low concentration (usually < 2ppm ) in natural samples, specific and sensitive analytical methods Are required for its accurate determination. We have developed a method using potassium n- butyl xanthate (KBX) for the determination of Hg. II.EXPERIMENTAL: All chemicals were of AR, GR or of high purity grade. KBX was prepared by mixing solid KOH with dry n- butyl alcohol until whole? content was dissolved. Then CS2 was added to get solid mass of KBX which was then recrystal lized. wHq tracer (8.5mCi as Hg(NO»)2 Sp.activity 579mCi/g) and other isotopes were procurred from BRIT. These were diluted and then suitable aliquota were used. Carrier solution was prepared by dissolving O.3383g HgCl2 (AR, BDH) in 5ml HC1 and made to 23Oml. Radiochemical Procedure: lOOjul of Hg tracer (J5.34 g) was taken in a beaker containing lml (2OO/jg) carrier solution and 9ml buffer of pH 8.2. After stirring with slight warming, the contents were transferred to a separating funnel. A mixture of lml IX KBX in methanol and 9ml of chloroform were added. A yellow coloured complex formed which was extracted. 2ml aliquots of each phase were counted at 279keV photopeak of 2O3Hg using 2" x 2" Nal(Tl) detector and SCA. 111.RESULTS AND DISCUSSION: In order to develop the radiochemical procedure various conditions Mere optimized as follows. a.Effect of pH; Several systems in the pH range 1-12 were prepared and extractions were carried out. It is observed that maximum extraction (98X) occurs at pH 8.2. On further increase in pH,extraction decreases.However, extraction can be carried out in lower pH range also b.Nature of solvent: The effect of different solvents on the «K tr diet ion was studied and following order was observed. Chloroform (937.) > Carbon tetrachloride (94%) > tri n-butyl phosphate (927.) > n-Butanol (89X) > n-oc'.anol (8&7.) > n-amyl alcohol (847.) > n-pentanol (837.) > Toluene (827.) > Nitro benzene (75X) > Isobutyl methyl ketone (717.) > Benzene (617.). c.Quant itative Nature: Several systems containing different amounts of Hg (tracer + carrier ) were extracted with excess of reagent Excellent linearity is observed in the concentration range 1O—lOOng AR - 39.1 and l-10jjg. In most biological samples Hg present concentration range of 100ng-i^/g. d.Effect of interfering ionsi The possible radiometric and chemical interferences were evaluated by extracting v/"v 2OO pg Hg (Tracer carrier) alongwith 2OO/jg of different interfering ions or __^ activities using respective radiotracer. It was observed that Cr, 76As, 75Se, 65Zn, 5"r,n, 133Ba, 134Cs, 11OAg. 11OSn, 125Sb,2O4Tl do not 115m W 60 interfere at all. Howover, CdF Fe and Co interfere to the •xtent of 157..Co and Cd can be masked by using IX sodium citrate solution whtriAi F« can b« aliminated by prior extraction with diethyl vther and then processing the sample for Hg extraction. e.Precision and accuracy: Replicate measurements were made in the concentration range of O.l-ll^/g of Hg as given in Table 1. It i- observed that results are lower by Table 1.Precision and Accuracy for the determination of Hg(II). Amoun t Amount Mean + S.D. Error t«k»n recovered (/jg) (7.) O.1113 O.1O72, O.1O84, O.lO7d O.1O78 + O.OOO5 - 3.1 1.113 1.11O, 1.104, 1.1O2 1.1O6 + O.OO4 - 0.63 11.13 tO.96 1O.94, 11.O3 1O.96 + O.Ol - 1.5O Table 2.Determination of Hg(II) by substoichiometric IDA. Amoun t Amount Mean + S.D. Error taken (ug ) recovered (/UQ) (A (7.) 5.2O 5.32., 4.79, 5.56, 5 .04 5.18 + O.29 - 0.38 1O.4 10.82, 11.50, 1O.24, 1O,.91 1O.86 + O.45 + 4.52 15.6 1A.92, 15.32, 16.29, 14..70 15.81 + O.86 + 1.3O 2O.8 21.94, 21.27, 22.83, 22.15 22. O5 + O.56 + 6.O1 J ^'.REFERENCES!. 1. S.M.Lin, C.H.Chiang, C.L.Tseng and M.H.Yang, Radiochem. Radioanal. Lett. 56, 261 (1V83). 2. H.Boddeley, B.J.Thomas, B.W.Thomas and V.Summers,Br.J.Rad iol. §6 449 (1983). 3. A.N.Krylova, V.N./hulenko and M.A.Malyarova, J.Anal. Chem. USSR, 4JL 53 (1986) ACKu^wLEDGFMENT:Financial assistance from IAEA, Vienna (contract no. 5649/RB/and UBC award to PDS under FIP Are gratefully acknowledged. AR - 39.2 SIMULTANEOUS DETERMINATION OF Sb AND Se BY RADIOCHEMICAL NEUTRON ACTIVATION ANALYSIS R.B. LANJEWAR, R. G. WEGINWAR. N. L. CHUTKE and A. N. GARG Department of Chemistry. Nagpur University, NAGPUR- 44O O1O CINDIA3 SUMMARYt A radiochemlcal neutron activation method has been developed for the simultaneous determination of Sb and Se in environmental samples. The method involves reactor irradiation, fusion, dissolution followed by solvent extraction for Sb and Se and y- counting on scintillation spectrometer. The method employs o-phenylenediamine for Se and KI + PAR for Sb as reagents In benzene. CKeywords: Antimony, Selenium, Solvent extraction. Neutron Activation Analysis, Environmental samples) I. INTRODUCTION: Radiochemical neutron activation analysis CRNAAD is the most accurate and precise technique for trace elemental analysis of environmental samples/1/. Sb and Se both are potentially toxic and carcinogenic in nature. These are found to induce diseases like muscular dyastrophy, pancreatic fibrosis. apetite loss. pneumoconiosis etc. Several radiochemical methods have been proposed for the determination of these elements with other toxic elements in various environmental and biological samples/2,3/. A simultaneous radiochemical NAA method has been developed for the determination of Sb and Se in fugitive cement dust particulate samples and environmental standards. II.EXPERIMENTAL: Fugitive cement dust particulates were collected from Mandhar Cement Factory, Raipur by Air Pollution Control Division of NEERI, Nagpur. About 56-60mg each of dry samples and standards were encapsulated in high purity quartz ampoules and irradiated at a thermal neutron flux of 1O19 n cm s* for 1 week in CIRUS reactor at Bombay. These were fused with Na2Oz and Sc2Oa with lOmg each of Sb and Se carriers in Ni crucible and contents dissolved in 2OmL. 6M HC1. To a lOml aliquot. 2mL of 2% citric acid solution was added for masking of Sb and then Se was complexed with O. lJi o-phenylenediamine in ethanol. It was extracted with lOmL benzene/4/. The aqueous phase was reduced to 2mL and 6mL of 6M H2SO4 was added. To it lmL each of 1M KI and O.IK aqueous solution of 4-C2-pyrldylazoD-resorcinol CPARD w^re added and Sb was extracted with lOmL of benzene. Aqueous and organic phases were counted on a well type 2" x 1.75" NaICT13 detector coupled with IK. NDO2 MCA and Centronics 1S4 graphic printer. The method was developed using Se and Sb tracers whence all the conditions were optimized and then employed on environmental samples. III. RESULTS AND DISCUSSION: Concentrations of Sb and Se in NIST SRMs Urban Particulate Matter C16485 and Coal Fly Ash C1633a3. NIES CRM Pond Sediment and USGS standard rock BCR-1 are given in Table 1. Concentrations in all the standards are in excellent agreement with the cerified values Cerror <5>O except for Pond Sediment where it differs significantly from the reference value. Similarly concentration of Se in Coal Fly Ash and BCR-1 are also in excellent agreement with literature values except AR - 40.1 for Urban Particulate where it is lower by 1854. R. S. D. In all cases are <1O>{ Therefore, the method seems to yield reasonably accurate and precise data. Cement dust particulates from 5 different locations, cement meal and cement slag samples all collected from the same factory were analysed. Concentration of Sb in all the cement samples is comparable with that of Coal Fly Ash but quite low compared to Urban Particulate Matter. However, concentration of Se is much lower Cby an order? compared to Coal Fly Ash and Urban Particulate, though it is comparable with USGS standard BCR-1. Therefore, the dust partieualte in cement factory is quite safe as far as these two toxic elements are concerned. Table 1. Concentrations of Sb and Se in environmental standards and cement dust particulates. Standard/Sample Sb Se C/ug/g? This work Lit. Value This work Li t. Val ue Urban Particulate 46. 1+2.0 C453 22. 2+1. O C27D Matter NIST SRM 1846 Coal Fly Ash 7.1+O. 7 C6. 8? 9. 57+O.4O CIO. 3) HIST SRM 1633a Pond Sediment 2. 9+O. 2 C2. CD NIES CRM No. 2 BCR-1 O. 70+O.OS CO. 693 O.O96+O. OOl CO. USGS Cement Dust 7.5-12.0 C©. 53 O.O98-O. 1O7 CO.1O2? (Five stations? Cerrent Meal 9. 2+1. 2 O.12+O.O8 Cement Slag 15.6+1.6 0.O8+O. O4 » Reference or noncertified values. + Mean values of five stations of factory. ACKNOWLEDGEMENTS: We are grateful to DAE and IAEA Vienna Ccontract No. 3649/RBO for financial assistance. IV.REFERENCES: 2.J.Tolgyessy, E.H.Klehr, Nuclear Environmental Chemical Analysis. Ellis Harwood Ltd. West Sussex C19873 pp. 185. a W C Cunningham, J.Radioanal. Nucl. Chem. Articles. 113, 423 C19873. S.C.Y.Wu, P.Y.Chen. M.H.Yang. J.Radioanal. Nucl. Chem. Articles, 112. 133 C1987?. 4. .1. Kalouskova, K.Drabek, L.Pavlik. F.Hodik. J. Radioanal. Nucl. Chem. Articles, 189. ES9 C19893. AR - 40.2 RADIOCHEMICAL NEITRON ACTIVATION ANALYSIS OF Fe, Zn, Co, Sb, Se AND P IN CANCEROUS BREAST TISSUE RAJIV G. WEGIt'WAR and A. N. GARG Department of Chemistry. Nagpur University. NAGPUR-400 010 CINDI A? SUMMARY: Cancerous and normal breast tissues of 15 patients have boen analysed for Fe.Zn, Co. Sb and Se by radiochemical neutron activation analysis-P was determined oy counting of ft . It has been observed that all elements except Fe are enhanced in cancerous tissue. Se shows maximum enhancement in cancerous tissue compared to normal tissue. (Keywords: Radiochemical NAA. Cancerous Breast Tissue, Solvent Extraction, ft- counting). I.INTRODUCTION: Many trace elements influence the permeability of cell memb-i anes by incorporating into the normal cell and exert direct or indirect action on the carcinogenic process or change enzymatic activity of normal cell and accelerate the growth of tumor /I/. Several workers have attempted to correlate the elemental concentrations of cancerous and normal tissues from the same human breast of an individual /2,3/. We have analysed normal and cancerous br>ast tissues of 15 patients along with NIST and IAEA standards for comparison for Fe, Zn. Co. Sb and Se by radiochemical neutron activation analysis. P was determined by ft counting of P by employing gas flow proportional counter and Al filter. II.EXPERIMENTAL: Normal and cancerous human breast tissues were obtained from the same Individuals by mastectomy at the Government Medical College. Nagpur. These were cut into pieces and thoroughly washec in hot water to dissolve fats. After complete drying, tissues were powdered and irradiated in a Cobalt-6O Gamma Chamber to a dose of 2. S Mrads. 1OO mg each of samples and standards were packed in quartz ampoules and irradiated for 1 week at a thermal neutron flux of _io" n cm"Z«i * in CIRUS reactor Bombay. Phosphor us was determined by ft counting of 92P employing a gas flow proportional counter and 28 mg cm Al filter. Samples with 1O mg each of carrier solution of Fe. Zn, Co, Sb and Se were dissolved in minimum amount of aquaregla and final solution was made in 6M HCl.In lOmL aliquot.Fe was extracted using ather and then 6^ aqueous solution of cupferron in chloroform at pM 4. To the aqueous phase 2mL of 5>S KCH was added to mask Co and then Zn was extracted using O.1M 2-thenoyitrif1uoroacetone in lsobutyl methyl ketune at pH 6. Finally Co was extracted using a- nitroso ft- naphthol in chloroform at pH 5. Sb and Se were determined by solvent extration procedure described elsewhere/4/. Al 1 quots of organic phase was counted on a well type 2" x 1.75" NalCTl) detector, coupled with ND 62 IK MCA. III. RESULTS AND DISCUSSION: Concentrations of Fe, Zn. Sb. Se and P in NBS SRM 1577a CBovine Liver) and IAEA CRM H-4 CAnimal Muscle) in Table 1 . are in good agreement with the certified values and within an error of < 1Q'/. except for Co. It is observed that concentrations ot all elements are enhanced in AR - 41.1 cancerous tissue except in Fe which is depressed by 5.7/4 CTable.23. Zn shows an enhancement in cancerous tissue by ZO% compared to that of normal tissue. It is followed by P C4O. I'/O . Sb C50.7JO and Co CB3.65O. The increase in mean concentration of Se is maximum Cby almost 2 fold} in cancerous normal tissue. It is apparent that the change in concentrations of these elements in cancerous tisssue is real and establishes a definite pattern in the carcinoma of the breast. ACKNOWLEDGEMENTS! Grateful thanks are due to DAE for financial support. We thank Dr.CMrs? V. Sagdeo, Govt. Medical Collage, Nagpur for providing samples. I V.REFERENCES '. S.L.Rizk and H.H.Sky-Peck. Cancer Res. 44. 539O C198OJ. 2. P. C. Map.gal and S.Kumar, Indian J.Phys. 58A. 335 C19845. 3. M. S. Bratakos, T.P.Vouterakos and P. V. Ioannou, Sci . Total Environ. 92. 207 C1990J. 4. P. B. Lanjewar. R. G. Weginwar , N. L. Chutke and A. N.Garg, This symposium volume. Table 1. Elemental concentrations in biological standards. Element Bovi ne Li ver Animal Muscle NIST SRM 1577a IAEA H-4 This work Lit.Value This work Lit .Value 1. 12+O. O3 1. 11 O.675+O. O12 O. 683 FeC 187+3.7 194 51.4+1. O 49. 1 128+2.5 123 81.7+1. 7 86. O O. 74+C. 05 0.71 O. 27+O. O3 O. 28 2. 71+O. 5O — — 3. O CoCng/g} 172+8 210 2.8+1 . 2 3. 1 Table 2. Range and mean values of trace elements in tissues. ELEMENT NORMAL CANCEROUS CHANGE C'sO Ranae Mean Ranae Mean FP Cpg/g} 20 .9-50.0 42. 2+11.3 18.7-61. 3 39. 3+13. O -5.7 Zn Cpg.^gJ 22 .4-52.3 40. O+8. O 29. O-61 . 1 47. 3+8.5 +20. O P CVO O. ll-O. 39 O. 3O+O. 1O O.3O-O.5O O. 45+O. 1O +4O. 1 Sb Cng.'gD 2. 50-7.38 4.36+1.35 2.6O-9.74 6. 57+1.94 +50.7 Co Cng/g} 1. 02-5.60 3.93+1.16 1.31-7.05 6. 43+1 . 46 +63.6 Se Cjjg/g3 O. 45-1.O8 O. 71 +O. EO O. 68-1. 75 1.44+O.31 +1O4 AR - 41.2 DETERMIHATION OF TRACE AMOUNTS OF INDIUM BY RADIOCHEMICAL DISPLACEMENT N. Rajesb, M.S. Subramanian Department of Chemistry Indian Institute of Technology Madras - 600 036 SUMMARY A rapid method has been developed for the determination of trace amounts of indium using radioactive zinc dithizonate as the reagent. The displacement of radioactive zinc from the dithizonate complex by Indium has been successfully utilised for the determination of 5-25 Ug Indium. [Key words : Indium, radiochemical displacement, radioactive zinc dithizonate) I. INTRODUCTION Dithszone has been employed for the solvent extraction of a large number of metal ions. A radiochemical displacement method for the estimation of palladium using radioactive zinc dithizonate anf< mercury dithizonate has been reported /1,21. The present method explores the possibility of determination of trace quantities of Indium by displacement of labelled zinc from zinc dithizonate complex. The labelled zinc dithizonate in carbontetrachlor ide is equil ibrated with Indium in the aqueous phase. Due to the nighcr stability of the Indium dithizonate complex zinc is displaced from the labelled zinc dithizonate from the organic layer to the aqueous layer. The activity of the released zinc is proportional to the concentration of Indium added under suitable conditions. II. EXPERIMENTAL Solutions and Reagents -u i. Zinc acetate solution 10 M ii. Indium(lll) 1000 |ig/ml iii. Acetate buffer 1M Preparation of radioreagent 50 ml of 10 zinc(ll) acetate solution was mixed with approximately 17 MBq of zinc-65 tracer and the mixture was transferred to a 250 ml separately funnel. 15 ml of pH 6.0 acetate buffer was added followed by 20 ml of 2 x 10 M dithizone in carbontetrachloride. The contents were equilibrated for 10 minutes and the red coloured organic extract was collected in a polyethylene bottle. Its specific activity was determined employing 1 ml of the extract. Instrumentation The gamma activity of released zinc was measured by a single channel analyser coupled to d 1 3/1" x 2" Nal/TI well type scintillation detector. AR - 42.1 Effect of pH 10 ug of Indium was mixed with 2 ml of acetate buffer of varying pH (3 to 6.0) and diluted to 10 ml. The mixture was equilibrated with 5 ml of radioreagent for 10 minutes. The gamma activity of 2 ml of the aqueous extract was measured. A blank was performed under each pH conditions and the blank corrected activity was calculated. Calibration Aliquots containing 5 to 25 ug Indium were transferred to a 125 ml separatory funnel, 2 ml of acetate buffer pH 4.3 was added and diluted to 10 ml. Then 5 ml of radioreagent was added and equilibrated for 10 minutes. The gamma activity of 2 ml of the aqueous extract was measured. A blank was performed under similar conditions and the blank corrected activity was calculated. A plot of the corrected activity against the amount of Indium added served as calibration. Mil. RESULTS AND DISCUSSION The optimum pH for the maximum displacement of zinc by 10 yg Indium was found to be '1-4.5. The optimum time of equilibration for maximum radiochemical displacement was found to be 10 minutes. The calibration graph was found to be linear from 5 to 25 u9 Indium. There are about twenty metal ions which react with dithizone forming dithizonates. However the extraction constant of Indium(lll) dithizonate is sufficiently higher than many of the metal dithizonates with the exception of copper, silver, bismuth, mercury and palladium /3/. The interfering effect of these cations together with cadmium, cobalt, nickel barium, calcium, strontium, iron, manganese, lead, thallium, are being investigated at 1 mg level. Anions like citrate, tartrate, sulphate, bromide, chloride, phosphate, fluoride, nitrate, oxalate are also being investigated at 1 mg level. IV. REFERENCES 1. V. Nirmala, M.S. Subramanian, J. Radioanal. Nucl. Chem. Lett., 128/1988/23. 2. N. Rajesh, M.S. Subramanian, Accepted for publication in the Journal of Radioanalytical and Nuclear Chemistry Letters. 3. E.B. Sandell, Colorimetric determination of traces of metals, Interscience Publishers, New York, Vol. 111/1959/152. AR - 42.2 MULTIELEMENTAL NEUTRON ACTIVATION ANALYSIS OF WATER SAMPLES AND ENVIRONMENTAL STANDARDS Miss M. B.Pawar, Miss M. N.Ambulkar and A. N. 6arg Department of Chemistry, Nagpur University. NAGPUR-44OO1OCINDIA5 SUMMARY-A multiel©mental instrumental neutron activation analysis CINAA3 method has been employed for th'j determination of 14 trace elements in water samples from different reservoirs in Nagpur and several environmental standards from NIST. USA and NIES. JAPAN. CKeywords- INAA, Environmental standards, water samples.3 I. INTRODUCTION- INAA has been employed widely for the determination of trace elements in water /I-3/. Awareness about the toxic effects of nonmetals and heavy metals in water has enhanced the need for their determination at trace to ultratrace level. INAA has the advantage that the multielemental analysis is possible in small size sample. Salbu et al /I/ reported a simple and rapid mul tiel erne; it al method for the determination of 4O elements in natural fresh water by thermal neutron irradiation followed by GeCLi3 gamma spectrometry. Naeem /2/ used INAA for Ca, Cl, Na, Mg, K, Co, I, Mn, Sm and Sb in water samples using INAA. Zimjewska et al x3/ employed NAA for the determination of Ag, Cd, Co. Cr . Cu, Fe. Hg, Mo. Se and Zn in water. Sharma et al /4/ have employed NAA and EDXRF techniques for the analyses of water samples from all over Patlala city and determined 28 minor and trace elements in drinking and ground water samples. We have employed INAA for the determination of the 14 trace elements in 4 water samples and environmental standards. II EXPERIMENTAL- Water samples were collected from four different sites around Nagpur University Campus and evaporated to dryness without boiling. . The residue was scratched out. 3O-5Omg samples and environmental standards were encapsulated in quartz ampoules and irradiated at a thermal neutron flux of lO12n cm~2s~* for 1 week in CIRUS reactor. After opening, the samples were counted on 13O cm coaxial HPGe detector and IK ND 62 MCA with Centronics 154 printer. III. RESULTS AND DISCUSSION- Concentrations of 14 minor and trace elements in water samples and the standards were 'calculated using USGS standard rock BCR-1 and are given In Table 1. Concentrations of most of the 14 elements in all the standards are in good agreementC wi thi n +, lOJi) wi^th those of certified values reported in literature. Therefore. it is assumed that elemental concentrations in water samples must be accurate and reliable.These concentrations are within reasonable range of those analysed by Sharma et al /4/ and also those recommended by ICMR / WHO.Water sample from Telankhedi has higher concentrations of Fe, Cr and Sb. Acknowledgements- Grateful thanks to IAEA Vienna for financial support under contract no. 5649/RB. IV.REFERENCES - 1.B. Salbu, E.Steinnes and A.C. Pappas.Anal.Chem.47. 1Ol1 Cl975). 2.A.Naeem.J.Radioanal. Nucl . Chem. Lett. 1 OS.79 Cl9873. 3. W. Zimjewska.H. Palkowska-Motorenko and H.Stokowska, J.Radioanal. Nucl . Chem. 116, 243 C^9873. 4. H. K. Sharma.B. Singh. V. K. Mittal and H.S. Sahota, J . Radi at. Appl . lnstrum.3,^89 C19893. AR - 43.1 Jaole 1 : Car*entration*of eluients in Water samples ana Envir El merits Coal fly Ash Uman Vehicle Exhaust Kinds "ei ^-nkh?ci University Tap *ater SRM, 1633a i-1 articulate (-articulate Sediment Lake Sewage Fit SW, 1948 NILS No. 8 N:ES,NO.2 Fe (*) 9.0O+O.01 3.80+0.2 0.54+0.02 6.53+0.07 CM V.-.22 0.07 0.10 (974) (3791) (-) (6.53) (234) (186) (108) (590) Zn mg/g) 0.665* 0.074* 400 - 50 320 (220) (0.476*) (0.104*) (343) (5C.4) (12.6) (36.7) Cr Ug/g) 181 +13 374 + 28 31.0+2.2 81.6+6.8 32.5 53.9 14.5 17.2 (196) (403) (2575) (75) (4.1) (14.5) (3.6) (2.0) Co Ug/g) 41.6+0.1 17.6+0.3 3.1 • C.I 27.7+0.7 2.4 2.7 1.4 1.4 <46") (Ta) (373) (27) (O.30) (0.72) (0.34) (0.16) Sc ^g/g; 35.8+4.6 5.8+0.6 0.75+0.05 28.2+0.6 0.46 0.58 0.04 0.20 (40) 77) (0755) (28) (58.0)» (156)« (9.3)« (22.9)* 56 (ng/g) 5.3+0.2 12.2+1.4 2.2+ C.2 2.2+0.1 0.72 1.90 1.6 0.33 (2.8) (45) (2.12) (2.0) (91)» (512)* (4O3)» (38)* HQ (tfo/g) 0.99+0.17 3.17+0.54 0.72+0.1? 1.4+0.5 i.Cl 2.25 1.80 0.54 (C7l6) (-) (-) (1.3) (130)» (610)* (450)« (60)* Se(i»3/g) 9.4.1.1 24.1*2.9 1.6+0.1 1.5+0.1 0.78 2.6 3.9 0.78 (10.3) (27) U~3) (-) (98.3)* (701 )• (983)* (90)* Cs Ug/g) 10.8-tl.l 3.6+0.2 0.43*0.04 4.1,0.1 2.45 0.32 0.18 0.21 (11) (3.0) (0.24( (310)* (90)» (45)» (25)* cu (yg/g) 3.5+0.4 0.7+0.1 0.08+0.01 1.3+0.1 0.08 0.06 0.03 0.04 (4) (o7e) (0.05) (-) (9.4)» (17.5)« (6.3)« (4.8)» Ih l^g/g) 30 r 1 13 • 1 0.31+0.03 4.3+ 0.4 7.5 0.84 2.6 0.86 (24.7) (7.4) (0.35) (-) (0.95) (0.23) (0.65) (0.09) Te (iig/g) 1.4 3.6 4.5 1.9 8.7 3.0 2.9 3.4 (-) (-) (-) (-) (1.09) (0.81) (0.73) (0.39) Ta (ug/g) 1.9 4.3 0.27 0.28 0.02 0.15 (-) (-) (-) (-) (-) {-') (0.005) (0.02) lr Values In parenthesis for standards art certified values and for water samples In tig/L or with • In ng/L< iKVfiSTi NATION UN THE EVFECT OF SOLVENTS Or.' THE Dli-FUSION OF IONS BY RADIOACTIVE TRACKH Ti.CriMQ.UE A. Das and S.N.Changdar Applied Kadioactivity Unit Bose Institute, 93/1, A.p.c.Road, Calcutta-700o09, India SUMMARY: The experimental investigation on the dit tusiori mecha- nism in electrolyte solutions by a radioactive tiacer technique is described. The sliding cell method developed in our laboratocy consists of equal radio-active and nonradioactive liquid columns and the radiation detector is placed in verti^iL jeorri' t ry over in>- dittusi.on column. The dirrusion of Thallium ions nas oeen measured in water and heavy water for <.'irietent concenl.rations <>t rhallous Sulphate and the results have ueen .jimlysed r t om fh<< point of" view ot ion-ion and ion-solvent internet ions. Key words: ^illusion coer t icient, Heavy water, islidiny eel I, i ad ioi sot ope,, water st ructu re. 1. INTRODUCTION : In recent years there have b-.&n specLdc:ulet; developments in the study or diffusion phenomena in liquids Doth theoretically an-, experimentally. There has been some ni->w add: t i oi.s£l J in the radioactive tracer technique and computer simulation techniques have added a new dimension in this tit-Id of investigation. Inelas- tic neutron scattering, pulsed EnR tt-cnnque.s, optical measurements also have accumulated data on diffusion phenomena in aqueous elec- trolyte solutions. But most of these investigations centered around ion-ion interaction with no emphasis on ion-sol vent inter- action. By usiny the sliding cell metnod and using both H u ami DyO as solvent we are trying to see Doth the ion-ion and iin- solvent effects. Instead of the conventional method of .Measuring diffusion coefficient by noting the change in radioactivity at a fixed plane in a liquid as a function of time, our technique measures the time dependence of what is effectively an weighted spatial average ot radioactivity. This not only allowed a simpli- fication of experimental arrangement but also pushed down the uncertainties associated witn experimental diffusion coefficients by almost an order of magnitude. This very consistent and accu- rate method enabled us to measure the concentration dependence of isotopic diffusion in two liquid systems. EXPERIMENTAL : A radioactive tracer technique n.-iS been developed tor the study ot diffusion in liquids. The technique ia based on sliding cell mechanism and the experimental geometry consists of radioactive and nonradioactive columns of equal lengths witn the radiation detector placed vertically over the diffusion column. ., for the above experimental conditions when Kt^- 0.5 (where fcy - (*. ^ ii is the diffusion coefficient and 2L iu the total length of tneM1/ diffusion column; all even order terms vanish in the Fourier solution of FicK's second law of diffusion tor unidirectional tlow: t •• - - ,->X »- AR - 44.1 and the solution is adequately represented by the first term which leads to No-Nt = A exp(~kt> where Nt is the count rate at time t and No is the count rate when the liquid column is completely mixed up; A is constant depending on the geometry of experimental arrangement. Thus O can be measured by noting the variation of (Mo-Nt; with time tOJJ- For the present set of measurements 204 Tl is the tracer and the solution of Thailous Sulphate in H20 and D20 is the liquid system. Thailous Sulphate solution has been used over a wide range of concentration 15 mcm/c.c. to 50 mgm/c.c) and the temperature was kept fixed at 30°C RESULTS AND DISCUSSION: The experimental investigation clearly shows a decrease in diffusion coefficient with increasing concen- tration both in D20 and H20 systems. The diffusion coefficient is higher in H2O medium compared to D2O at the same concentration (about 2% at 5 mgm/c.c. and 10% at 40 mgm/c.c. of Thallous Sulphate?. The diffusion" coefficient at lower concentrations obeys Nernst's limiting law but the coefficients at higher con- centrations can be explained only by Onsager's phenomenological coefficients. The difference of diffusion coefficients of 2O4Tl ion in H20 and D20 systems showed indications of fundamental changes in the structure of solvent upon the dissolution of a salt £3"J . Actually comparison of diffusion and viscosity data of D20 and H20 solutions provides insight into the effects of various ionic species on the structure of the solution* Since pure D20 and H20 are considered to have different structural characteristics the introduction of metal ions should affect the structures differently . The structural difference of D20 and water is expected to be reflected in diffusion measurements, toe are carrying out the experiments with other ions and detailed results with explanation will be published elsewhere. REFERENCES 1 P.Passiniemi, J. Soln. Cnem. 12, bOi [1983) 2 o.N. Changdar, J. Pure and Applied Phys. 11, «11 (1V/3) 3 D.E. Woessner et al, J. Chem. Phys. 4y, 371 (1968) AR - 44.2 RADIOMETRIC PROSPECTING IN THE APATITE MINES OF SITARAMAPURAM, VISAKHAPATNAM DISTRICT, ANOHRA PRADESH, INDIA T.V.S.R. Kshira Sagar, M.3. Ramakrishna, B. Nagamalles%cra Rao and R.T. Naidu Department of Geology, Andhra University, Visakhapatnam 530 003, Andhra Pradesh, India SUMMARY The apatite-magnetite- vermiculite deposits of Sitaramapuram, Visakha- patnam District, Andhra Pradesh, India, are associated with intrusive carbona- tite and syenite emplaced in the khondalites of Eastern Ghats. Radiometric prospecting has been carried out in Xhe apatite mines to delineate the radio- activity poor and rich zones. The high radioactivity observed in the syenite band (0.12 to 0.18 mf/h ) is attributed to the rare earth concentration. Highest radioactivity is observed in the calcites and apatites of carbonatites (0.18 mr/h ) and it increases with depth.. The investigation has brought to light the occurrence of monazite in the 4/2 apatite mines. Key words: Carbonatite, syenite, radiometric prospecting, rare-earth elements. I. INTRODUCTION A detailed study of the apatite-magnetite-vermiculite deposits of Sita- ramapuram (Lat.18°17'; Long.83°09') Visakhapatnam district, India; has recently proved their carbonatitic nature/1,2/. These deposits are associated with carbonatites and syenites (alkaiic rocks) which are plug like intrusions in the khondalite rocks of Eastern Ghat formations in an enechelon pattern. II. EXPERIMENTAL The radiom«»tric prospecting has been carried out in the different-apatite mines of this locality with a radiation survey meter M-12IC (ECIL make). The activity observed is also corrobarated in the laboratories of the Centre for Nuclear Techniques in the Andhra University campus. III. RESULTS AND DISCUSSION The results are tabulated (Table I). The radioactivity measurements indicate that the carbonatite is relatively 2.6 to 4.5 times more radioactive than the surrounding country nocks, namely khondalites in which the values range from 0.03 to 0.04 mr/h . The syenite band at Sitaramapuram contains relatively .high radioactive constituents and gives a value ranging 0.12 to 0.18 mr/h" . The highest values are observed in carbonatite calcites and they increase with depth (0.18 mg/h at 70 metres) in accordance with the relative increase of calcite with depth. This indicates the persistance of underground carbonatite magma in this locality. The sudden increase in activity with apatite occurrence is in accordance with the structural emplacement of carbona- tite. The increase of radioactivity in the syenite band denotes concentration of rare-earth elements. The concentration of radioactive elements in syenite and carbonatite encounters their co-magmatic nature. AR - 45.1 This study has brought to light the occurrence of monazite at the 4/2 apatite mine. xhe monazite is reddish brown in colour and anhedral under microscope; it has high relief and shows high order interference colours. V. REFERENCES 1. M.J. Ramakrishna, Geology, petrology and geochemistry of carbonatite in parts of Visakhapatnam. Ph.D. Thesis, Andhra University, Waltair (1983). 2. T.V.S.R. Kshira Sagar, M. J. Ramakrishna and J.S.R. Krishna Rao, Geology, petrology and geochemistry of carbonatite in parts of Visakhapatnam District, Andhra Pradesh, India. 28th International Geological Congress held at Washington D.C., Proceedings volume 2, pp. 235-236 (1989). TABLE 1 Measurements of radioactivity (mr/h Feldspar- Locat ion Carbonat i te Syeni te biot i te gneiss K-apat i te mine 0. 12 0.03 K-apatite mine (83 m) 0. 13 J-apatite mine 0. 16-0. 12 0,03 4^-apatite mine 0. 12- 0.08 0.03 C-apatite mine 0.07 0.03 (dumps) Budi zone 0.09 0.02 -0.03 Sitampeta 0. 18 0.04 Panduganipudi 0. 10 0.03 apat i te mine Pandugani pudi apat i te mine 0. 17 (30 m) Y. Sitarampuram - 0. 18 AR - 45.2 Multimode Activation Analysis of Niobium in Geological Materials N.R.Das, D.Basu ana S.N.Bhattacharyya Saha Institute of Nuclear Physics 1/AF, Bidhannagar, Calcutta - 700064 SUMMARY A novel method of multimode alpha activation analysis (CPPA) of traces of niobium in geological materiaxj has been established. The activation products, 96Tc,95Tc, 95lflTc. 94Tc and 92mNb produced by the (oc,n) , («.2n), ( INTRODUCTION In trace or ultratrace analysis of most of the elements, the nuclear techniques involving neutron, proton, charged particles, etc., were found to be very effective. In our earlier investigation , the method of alpha activation analysis has been fruitfully applied to determine traces of niobium in some specific geological materials using the nuclear reaction, 93 95 Nb(«,2n) Tc. However, it has been observed that the alpha beam, as it passes through the target matrix, interacted with niobium at its different °3 94 degraded energy levels resulting in the nuclear reactions, " Nb{«,3n) Tc, 9JNb(«,3n)94mTc, 93Nb(«,2n)95Tc, 93Nb(u,2n)95mTc, 93Nb(«,n)96Tc and 93Nb(«,«< n) Nb, and each of the corresponding irradiation products has got favourable radionuclidic properties such as half life, y-energies, etc. suitable for determination of niobium. Thus, in the present paper, attempts have been made to utilise all the reaction products for determining traces of niobium in the geological materials. * EXPERIMENTAL 95 Radiotracer, Nb(35d) which has been used as radioindicator for niobium was procured from BARC, Trombay, India. The specimen phosphate rocks were chemically processed for preconcentration of the traces of niobium through a chemical procedure developed earlier'. AR - 46.1 The experimental samples along with the standards in Al-holders were irradiated in vacuum with alpha particle beam of "40 MeV, of current intensity 1-3 /iA and a dose of " 15000 JJC . The activation products were analysed by y-ray spectrometry with a HPGe detector in conjunction with a ND65 8K MCA. DISCUSSION In activation of niobium in Al.O^ matrix with an alpha particle beam which is gradually degraded with its penetration through the target matrix under the experimental conditions,, the upper most zone being where (oc,3n) reaction cross section with highest energy beam dominates over other nuclear 94 reactions as indicated by the formation of the reaction products, Tc(293m) and Tc(52m). Another reaction, Nb(«,rr n) Nb,also takes place mostly at the surface zone with the projectile energy around 36 MeV. In the next deeper zone of interest wh-:re the projectile energy is degraded to about 26 MeV, the possibility for («,3n! reaction decreases and that of (a,2n) reaction 95 95 increases and the reaction products, Tc(20h) and Tc{61d) become most abundant in this region. Further inside the matrix as the projectile energy comes down around 18 MeV, the reaction («,n) predominates producing Tc(4.35d). In the present investigation, unlike the previous one , the radionuclidic property of all the five activation products have been considered independently for determination of traces of niobium in geological samples and the results derived from the activation products of wide ranging half lives more or less corroborated to each other. The interferences due to the detected contaminants, scandium and iron, were eliminated by taking proper precuationary measures. Thus, it is inferred that the technique as a whole can be efficiently applied to determine traces of niobium in the geological materials', in general, by judicious choice of the reaction product. REFERENCES 1. N.R.Das, D.Basu and S.N.Bhattacharyya, J. Appl. Radiat.Isot., Part A, 38(11), 939, (1987) 2. N.R.Das and S.N.Bhattacharyya, J. Radioanal. Nucl. Chem, Article, 68(12), 75, (1982) AR - 46.2 RARE EARTH POTENTIAL OF THE CARBONATITES OF VISAKHAPATNAM DISTRICT, ANDHRA PRADESH, INDIA AND THEIR GEOCHEM1CAL SIGNIFICANCE T.V.S.R. KSHIRA SAGAR', B. NAGAMALLESWARA RAO* R. PARf HASARATHY2, p.B. PAWASKAR AND P. ILA 1. Department of Geology, Andhra University, Visakhapatnam 530 003, India 2. Analytical Chemistry Division, BARC, Trombay, Bombay ^00 005, India 3. Cambridge University, Massachusetts, U.S.A. 02139 SUMMARY A number of apatites, calcites and one monazite sample of carbonatites from Sjtaramapuram, Boarra and Si valmgapuram of Visakhapatnam District, Andhra Pradesh, India, are analysed for their rare-earth element (REE) content besides thorium and uranium. The study of the tat bonatites, besides proving a potential source for REE, Th and U, provides adequate evidence for the carbo-natitic nature as also the alkaline environment with which the magma is associated. Key words: Carbonatite, rare earth element, magma, geochemistry, alkaline environment I. INTRODUCTION As a part of the bi oad based exploration for nuclear raw material resources, the carbonatites recently located at Si laraiimpuram (Ldt. 18" 17'; Long. 83°09'), Borra (Lat. 18° 17*; Long. 83°O3") and Si valingapuran: Uat. I8°12'; Long. 83°01') have been investigated for (heir mineral content. A number of radioactive apatites, caJcites, collected horn the three localities as well as one monazite located at Sitaramapuram have been analysed for their Th, U and REE. II. EXPERIMENTAL Some calcite and apatite samples as well as the monazite sample are analysed in the Analytical Chemistry Division of BARC, while few other calcites and apatites have been analysed in the Cambridge University, Massacnusetts, U.S.A. The results are tabulated (Table 1). III. RESULTS AND DISCUSSION The presence of minerals like rnonazite as also abundant proportions of REE in apatites and calcites prove that the carbonafites host useful nuclear raw material resources. The monazite analysis reveals, LaJd~ 21.9; Ce?0, 33.3; Y2O3 0.06; U-jOg 0.01 and ThO 2.8. An indepth study of tins region to understand its' full potential is therefore suggested. From the academic stand point the present study throws much light on the genesis of the deposits. The geochemical study relating to the nature, distribution and enrichment . pattern of REE substantiates the carbonatitic nature. The typical LREE enrich- • ment relative to HREE noticed both in calches and apatites not only pin points its carbanotitic nature, but also throws light on the alkaline environment with which the magma, is associated. The RUE content matches well with the alkalic carbonatite of Fleicher and Altschular. Even the analysis of monazite substantiates these findings. AR - 47.1 TABLE 1 REE, Th, U (ppm) in apatites and calcite; of carbonatites A B C D E 1 F G H I 3 2 La 3992 3900 3780 3610 5220 26. 00 50r0 5960 125 1.8 80 103. 53 Ce 9182 8780 9695 3880 9440 47. 50 79^0 9595 140 3•0 105 147. 00 Pr . 4. 80 Nd 3873 3595 1 7.20 285, 960 1.4 Srn 395 15 400 420 25.5 1 .80 250 260 13.6 0 . 17 10. 1 7. 7 Eu 106 185 102 1 18 170 0.20 70 68 2. 1 0 .032 1 .6 JO Gd 1 .6 1 Tb 20 66 IS 14.6 15.8 0. 10 12.6 12 10.5 0.9 0. 79 Dy — 0.50 -- • Yb 1 1 .5 11.2 0. 10 12.6 12 12.5 0 .04 1 .6 1 .68 tv) Lu 1 .6 1 .6 0. 10 1 .43 1.13 0.4 0.005 0.23 0.36 Th 700 622 40 55 0.58 43. 6 3 U 70 100 ------•------2. 70 A-E Apatites from Sitaramapuram F-G Caicites from Sitaramapuram H Calcite from Sivalingapuram 1-3 Caicites from Borra 1 Average of 62 samples of apatites from alkalic carbonatites (wt. %) after Fleisher Michael and Z.S. Altaschular, Geochim < Cosmochirn Acta, 33, 725-732 (1969). Average of 4 carbonatite caicites from Munnar, Kerala (after Santosh et al., 1987, 3. Geo.L. India, 29, 335-343 (1987). IN VIVO MEASUREMENT OF TOTAL BODY CHLORINE BY PROMPT GAMMA NEUTRON ACTIVATION ANALYSIS S. Mitra, L.D. Plank, G.S. Knight and G.L. Hill University Department of Surgery, Auckland Hospital, Auckland, New Zealand SUMMARY - Prompt gamma neutron activation analysis with o8Pu/Be sources has been used to measure tolal body chlorine (TBCI), in vivo, following the reaction "C^n/y^CI. The precision of the method has been determined from replicate scans of an anthropomorphic phantom and is found to be 4.9% (SD). The subject dose equivalent is less than 0.3 rnSv. The possibility of measuring extracellular water (ECW) from TBCI and total body water (TBW) from a combination of total body potassium (TBK) and TBCI is demonstrated. Results are presented for 32 normal volunteers. (Key Words: in vivo neutron activation analysis, chlorine.) I. INTRODUCTION A prompt gamma in vivo neutron activation analysis (IVNAA) facility for measurement of lotal body nitrogen (TBN) and hence protein (lotal body protein (TBP) = 6.25xTBN) in the critically ill has been in operation in the Department of Surgery, Auckland Hospital, since 1982 /1,2/. Body composition determination in terms of protein, water and fat is extremely useful in monitoring the nutritional state of patients and their response to nutritional therapy. The separation of the total water content of the human body into inlracellular water (ICW) and ECW is of particular interest in critical illness where these spaces may undergo rapid change. Measurement of these fluid compartments is thus important for determining appropriate electrolyte prescription. Since chlorine is found predominantly in the extracellular space a measurement of TBCI could serve as an index of ECW. TBCI has been measured in the past by a delayed gamma in vivo neutron activation analysis procedure using (he "CI(n,y)38CI reaction /3,5/. The patient is first irradiated and then moved to a whole body counter. Such a procedure is not logistically viable for seriously ill patients. Also the dose equivalents are high in order to obtain clinically useful precision values. The present paper deals with a prompt gamma technique for measuring TBCI and uses the same Nal(Tl) spectra used for determination of nitrogen. II. METHODS The Auckland IVNAA system consists of I wo collimatcd 281 (>Bq ^Pu/Be neutron sources (neutron output 2.2xl()7 neutrons s' per source) placed above and below the subject, who lies supine on a scanning couch. The prompt gamma photons emitted by (he subject are detected by two pairs of 12.5 cm x 15 cm Nal(T!) detectors placed one each on either side of the subject and with their axes parallel to the length of the subject. The detector outputs arc processed by conventional nucleonics. The subject is scanned through the neutron field and composite spectra are collected via a multiplexed 4096 channel MCA for 2000s livetimc. A typical spectrum is shown in Fig. 1. The peaks of interest in the spectrum for this study are hydrogen (2.223 MeV) and chlorine (5.6 and 6.1 MeV). These major chlorine peaks are however contaminated due to the emission at the same energies front fast neutron im.lf.slic scattering with oxygen present in the subject and in the radiation shielding. A pure component of Cl de-excitation in the complex prompt gamma ray spectrum was identified due to a transition at 8.57 MeV. Due to the poor spectral characteristics at [his energy, peak stripping was not attempted. Instead, the integral counts over the region 8.3 to 8.8 MeV were used. This region of interest (ROI) gave the best linearity of counts with change in the chlorine concentration: (he concentration of chlorine was varied from 2 to 16 times the normal physiological level (0.12% of body weighl/6/) and irradiated in a cubic acrylic box (35cm x 30cm x :10cm). The H peak at 2.223 MeV was peak stripped from the underlying continuum by a linear least squares algorithm after locating the spectral minimum on either side of the peak. Since no peak stripping was performed, background modification in the region H.3 to 8.8 MeV due to the emission from any higher energy gammas had to be ascertained before calibrating the system for chlorine. Of the major elements in the body, only nitrogen de-excites (15%) at higher energy with a prompt emission at 10.8 MeV AR - 48.1 due to the "N(n,y)15N reaction. The contribution to the Cl ROI was determined by irradiating box phantoms of various sizes containing nitrogen-only solutions. The ratio of the net count rates due to N in the nitrogen ROI (9.5 - 11.0 IvieV) to that in (he CI ROI has been found to be a constant for all depths of solution. Net count rates from a ROI were derived by .subtracting a water background obtained by irradiating an equivalent water phantom. The facility was calibrated using hyu'rogen as an internal standard and the mass of chlorine was determined from the observed ratios of chlorine to hydrogen counts /'// where MM is (he total mass of body hydrogen, q is a calibration factor determined from an anthropomorphic phantom containing known masses of chlorine and hydrogen. X,-,/XH is the net Cl/H counts ratio corrected for background and body habitus (i.e (he effects due l<> body sixc and shape on the absorption of neutrons and gamma rays). Using a five-compartment model of body composition, namely, water, protein, fat, mineials and glycogen where the mass of fat M,. is expressed as the difference between body mass M and (he other compartments, MH reduces to /8/ Mil = Mwfw t M,,{e + (M-MwMp M(!-M,,,)fp 4 Mtifti where Mw, Mh Mm, and M(J are the total body masses of water, protein, minerals and glycogen, respectively, and the coefficients f represent the fractional amounts of hydiogcn in (hese compartments. Mw is dclt-imined by Iriliiim dilution, Me by IVNAA and M by weighing the subject. The two small compartments (minerals and glycogen) are assumed to be related i:> Mw /8/. F.C'W was derived in 32 healthy volunteers using the relation /')/ ECW (1) - 0.9 TBCI (mmol)/plasma ( I (mmol/l) and ICW obtained from /V/ IC:W (1) = TBK (mmol)/l47, wi.ere TBK was measured by *'K whole body counting. III. RESULTS AND DISCUSSION Total body water obtained from the sum of ECW and ICW compares favourably with TBW as measured by tritium dilution (Fig.2). The regression solution for TBW(CI,K) against TBW(tritium) for the 32 subjects was TBW(CI.K) = l.(X)TBW(trilium) - 132 (r = 0.<»3, SEE = 2.7). An expected average plasma chloride concentration (l()() inM /•>/) was used in this study. The availability of individual plasma chloride measurements would be expected to further improve the correlation between the two estimates of TBW. I( is assumed (hat the plasma chloride concen' llion is representative of the chloride level in (he tola! ECW. The ratio of ECW to TBW (tritium) averaged 0.44 ±0.04 (SI)) for the 32 subjects, close to the expected 0.45 based on bromide dilution /<>/, suggesting that our derivation of ECW is reliable. Further work is required to validate the constants used to derive KCW and ICW from TBCI and TBK both in health and disease before this approach can replace tritium dilution for TBW in clinical situations. « IV. ACKNOWI.KDCEMKNT This woik was suppoiti il by I li - - Modiial Kcsi .ncli Council of New Zealand. V. IIEFERKNCKS 1. A.il. Beddoe, 11. Zuidmeer and til.. Hill, I'liys. Mod. Hiol., 29, 371 (I9K4). 2. J.F. Sulclilic, S. Mitia and til.. Mill, I'hys. Med. Biol., 35, 10H9 (MM). 3. N.S.J. Kennedy, R Ea.slell, M.A. Smith and P.Tothill, I'hys. Med. Biol., zX, 215 (1983). 4. A. Shurali, D. I'eurson, C.B. Oxl.y, B. Oldroyd, D.W. Kiupowic/, K. Brooks an AR - 48.2 60- Line of identity 10'. 50 103. 2.22 H ft I I j. 40 5.6 6.1 o 70 Cl + O m I 102- 00 9.5 11.0 I—N—• 30 Ui t 10 1 8.57 a 20 30 10 50 60 2.0 4.0 6.0 8.0 10.0 12.0 2C •"3W 'tritium), kg Energy (MeV) Fig. 1. A typical prompt gamma spectrum from one NalfTI) deteaor collected for 2000 s Fig. 2. Comparison of TBW measured by tritium dilution with that aenved from TBCI and livetime TBK. INSTRUMENTAL ACTIVATION ANALYSIS 0? MILLETS 3Y M-252 N3UTR0N SOURCE N.3. Rajurtcar (nee Adhyapak) and K.'-i. Kanade Department 6f~5hemi3tfy, University of Poona, Pane 411 007, India. 31H-1MARY : The concentrat i of mineral nutrients viz. Na, K» Kn and Cl has been determined in different varieties of Milleta : Jowir 3nd 3ajra by Instrumental Neutron Activation Analysis (IMAA) using Cf-252 source. The gamma ray spectra of the radio- isotopaa forme'! were analysed on multichannel aualyzer (MCA) coupled to IlPie letector. It is found that the elemental content varies from species to species and is in the order Cl>K>Na>Mn. KEY WORDS : INAA, Millets, Mineral nutrients. I. INTRODUCTION : The importance of trace elements in nutrition is now widely recognized and the list of elements that are indinpensible to the proper nutrition is increasing steadily, therefore the determination of trace element contents of food stuffs has a great importance. In the previous paper we have reported the mineral nutrient contents of Wheat, 3engal gram and Rice" , the present paper deals with the determination of Na, K, Mn and 01 contents of different varieties of millets : jowar and bajra which constitute the bulic of the Indian diet t?y Neutron Activation Analysis5 . II. 3XPSRLM3NTAL : The grains of different varieties of jowar and bajra were collected from seed testing division of Agriculture Department, Pune. All these samples were crushed, sieved to an average mesh size of 70, weighed and packed into air tight polythene capsules. The samples and the standards were irradiated under identical conditions for 24 hqurs in J£-2j2 spontaneous fission neutron source (flux ~10 n §'). J l 4 f 5C The activities of the radioisotopea formed vi^. " " Nas ' K,* mn and itCl were measured on K3A couoled to ;{P In case of various j owar arimolea, thoujh 3PV-357 variety shows highest chloride content it seems to be poorer in Mn and K contents than rest of the varieties. Further examination of Table 1 shows that MSFf-51, 03H-6 and G3H-9 show maximum concen- tration for Mn, Na and K respectively while Na and (31 Is found to be minimum in CSH-5 and 3HSr-6 respectively. The differences in concentration of a given element in different varieties of the same species is attributed to their genetic origin as well as to the influence of cultivation circumstances. The data obtained in the present work will provide information about the adequacy or otherwise of the concentration of the various elements in Indian diet. IV. RE?SrlSNJ35 1. N.5. Rajurkar (n6e Adhyapak), N.P. Shah and J. Purushottam, Int.J.Radiat.Appl.Instrum. Part A, 41, 579 (1990). 2. 3.P. Patil, N.5. Rajurkar and K.J. Kanade, presented at Int.3onf.Aot\n.Anal, and its Apol., 3eijin^, China (1990). 3. a. Hevesy and H. Levy, Mat.Phys.Medd.Dan.Vid.Selak, 14, 3 (1936). : INA.A of various elements in different varieties of jowar and bajra Variety Oonoentration of the element/% Mn/10" Na/10"4 K/10" 01/10' Jj>war 3PV-357 208 • 6 105 • 2 45.7 • 1 222 • 3 MSH-51 296 1 8 83 • 3 53.2 • 0.5 165 • 8 05H-6 275 • 4 106 t 4 51.0 • 2 115 * 6 05H-5 250 • 6 71 t 1 55.5 t 2 125 • 10 05FI-9 215 • 5 96 t 2 59-9 • 0.9 122-1 8 Bajra IOTP-8203 244 t 2 62..1 2 34.2 * 0.9 74* r> RHR-1 434 • 5 81 • 2 49.0 • 5 154 * 1 MR-123 331 • 6 152 * 8 62.7 t 1 102 • 1 HHR-2 173 * 5 85 * 1 42.0 t 1 115 • 5 3-207 206 • 2 114 t a 58.0 • 0.6 ri? * 3 OK-560 211 • 7 154 t 0.6 59.0 t 2 82 • 5 MBH-110 232 • 5 66 • 6 58.4 • 2 98 # 1 AR - 49.2 USE OF MAGNETISED CELLULOSE IN RADIOIMMUNOASSAY FOR TRIinOOTHYRONINE< l'»> . R.B.Thorat, N.Jyotsna Radiopharmaceuticals Operations, Board of Radiation and Isotope Technology, V.N.Purav Matg, Bombay- 400 094. SUMMARY : An essential requirement for reliable radioimmunoassays Is an efficient and practical method for separation of the bound and free ligand fractions. Solid phase separation methods which utilise antibody immobilised on magnetic particles are admirably suitable for this purpose, as a very clean separation is possible using simple inexpensive equipment. In this work an RIA procedure has been developed for triiodothyronine using antibody coupled to magnetised cellulose. Key words :Radioimmunoassay,Triiodothyronine,Magnetic eel lulose. I. INTRODUCTION: Radioimmunoassay (RIA) procedures are now widely used for invitro diagnosis of endocrine disorders and also for a variety of other chemical investigations/!/. The use of" magnetic traps to separate free and bound antigen would simplify the design of automated solid-phase RIA systems. In the magnetised particles solid-phase supports, a magnetic carrier for the antisera allows magnetic fields to be substituted for centrIfugation/2/. The high surface area available on the magnetic particles and their easy dispersion throughout a solution allow for rapid and complete capture of target antigen. The magnetic responsiveness of the particles allows for rapid, high efficiency washing to reduce non-specific binding, which often limits the sensitivity of serological assays. Th se features form the basis of extremely rapid and flexible assay for several hormones. In this work, iron oxide has been inoorporated Into finely divided cellulose for the preparation of magnetic solid-phase support for triiodothyronine (T3 ) antibodies and a solid-phase RIA standardised. II. EXPERIMENTAL: 20g of ferrous sulphate powder was dissovled in 500ml of double distilled water by slightly warming the solution. To this 15g of cellulose was added and the suspension stirred for lh. Ammonia solution was then gradually added to precipitate the iron hydroxide, filtered and washed several times with water and vinally with acetone. The cellulose cake was air dried to give a dark brown powder possessing the required magnetic properties. The coupling of Ti antibodies to the magnetised cellulose was carried out by the periodate oxidation method/3/. The washed immunoadsorbent was stored as a concentrated slurry in phosphate buffer (pH7.5) containing 0.5% bovine serum a 1bumin(BSA). m AR - 50.1 To carry out he RIA of T5, lOOul of standard i, solution/ human 12S serum sample was incubated with lOOul each of I-T3, hormone free serum, and the prepared immunoadsorbent for 3h at room temperature. After the incubation, the tubes were placed in a magnetic rack for IS min, when the eelluiose settled down and the supernatent could be decanted. The radioactivity bound to the pel let was counted. Fig. 1 shows a typical dose response curva. III. RESULTS AND DISCUSSION: The magnetised eelluiose prepared by the above method was rapidly attracted by a magnet. The immunoadsorbent prepared from this cellulose showed a very good uptake of antibody as well as other characteristics of an efficient separation medium. The performance of the assay was assessed by tests of reproducibi1ity and accuracy. The inter- and intra- assay variations were less than 10%, while the recovery of Tj from spiked samples was 96% to 100%, these values being comparable to that of other procedures. IV. REFERENCES: 1. Yalow R.S. and Berson S.A., Nature ,184, 1648 (1959). 2. J.L.Guerdon and S.Aurameas. Immunochemica1 technique part B. In methodism enzymology. Edited bsf John. V. Langone and Helen Van Vanakis (1981), pp471~482. 3. Sanderson C.J.anri I'ilsnn D.V. 1 mmuno1ogica1 methods, 20, 1061 ( 1971 ) . STANDARD CURVE 10 20 30 . STANDARD CONCENTRATION 09/01!) FIG. I AR - 50.2 PREPARATION AND CHARACTERIZATION OF RADIOLABELLED MYOGLOBIN A.Rathinaswamv Radiopharmaceuticals Operations, Board of Radiation and Isotope Technology, V.N.Purav Marg, Bombay- 400 094. SUMMARY: Labelled antigen is an important reagent in any immunoassay system besides antiserum and standards. The preparation and characterisation of such a product namely radialabel1ed myoglobin is reported in this work. The suitability of this product to measure the serum myoglobin levels by a RIA system in cardiac patients to detect the transient abnormal elevations of myoglobin levels in cases of myocardial infarction has been demonstrated. (Key (iiorris: Myoglobin, Radioimmunaassay, Myocardial infarction) I. INTRODUCTION: Great advances have been made in the development of immu.no diagnostic procedures far estimation of minute levels of hormones, proteins, tumor markers, drugs etc with wideranging applications in various disciplines of life sciences in general and clinical chemistry in particular. Myoglobin is a substance similar to the hemoglobin in the red blood cells that can store oxygen until needed by the mitochondria- Myoglobin can hold a reserve of oxygen when the partial pressure of oxygen is too low for hemoglobin to hold the gas. Thus when the partial pressure of oxygen is relatively low and the muscle needs some oxygen, the reserve oxygen in the myoglobin can be unloaded to the tissues. The hemoprotein leaks out as a result of tissue damage as in the case of a heart attack '.myocardial infarction) and because of its relatively small size (Molecular weight of myoglobin is '7,500 daltons), the molecule filters through the glomerulus and appears in the urine. The measurement of serum myoglobin by radioimmunoassay (RIA) could thus be a rapid and reliable means of excluding significant myocardial injury /I,2/. Levels of myoglobin detected by an immunoassay procedure has a high sensitivity of over 90*/. for 'acute myocardial infarction (AMI), eventhough its specificity is only 80-85'/. /1,2/. II. EXPERIMENTAL: Iodination of myoglobin was carried out using Chloramine-T as the oxidising agent. The labelled protein was purified by gel filtration over Sephadex S-25 column (25 X lcm) and 50mM phosphate buffer (pH7.5) as eluent. Tire labelled myoglobin thus collected was characterised for its radiochemical purity by paper e1ectrophoresis. The quantity of the oxidising agent and the reaction time were optimised to get maximum yield without affecting the quality of the labelled product measured in terms of immunoreactivity and standard curve (dose—response). Stability of the labelled myoglobin stored in iyophilised form was studied over a period of time. The labelling of myoglobin was carried out using Na131I also for comparison of the performance, with respect to assay parameters and stability. AR - 51.1 Various reagents required namely anciserum, standards, myoglabin free serum, assay buffer and separating agents, to develop an iffidiunoassay were prepared and optimized. With the optimized reagents the assay system WHS studied for various parameters like incubation time and temperature. A protocol was evaluated for the rapid measurement of myoglobin involving incubation for only 2h. III. RESULTS AND DISCUSSION: The purified labelled protein showed a radiochemical purity greater than 90'/, and was stable upto a period of 60 days when stored in lyophilised form (Table 1). The studies carried tut with t:51I laLcflled myoglobin indicated less stablility supporting the known higher stability of 1ZSI labelled proteins. The equilibrium assay optimized gave a sensitivity of lOng/ml at 90*/. intercept. 80 normal serum samples were analysed and the normal rancie was established as 0~35nq/ml. Table 1 STABILITY STUDIES OF MYOBLOBIN-*"! DAY '/, R C PURITY V, NSB '/. 7ER0 BINDING 0 93 4.0 50 2 92 4.5 49 5 89 3.5 46 10 90 3.6 45 15 87 5.5 46 30 89 4.0 45 40 B6 5 .0 48 50 84 5.4 49 60 SO 6.0 39 IV. REFERENCES: t. L.R.Withenspoon, S.fc.Shuler, I'l. M . Garc ; a and L . A . la J 1 mger . J.Nucl.Med. 20, 115 (1979). 2. A.Maddlson, A.Craig, S.Yusuf, R. l.apez and P.SI eight. Cl in. Chi m. Act a 106, 17 (1980,' AR - 51.2 NT - Nuclear Technology and Instrumentation Papers : NT -01 to NT -03 IRON-BEARING MINERALS IN SOME INDIAN COALS CHARACTERISED BY MOSSBAUEJ* SPECTROSCOPE L. C. Ram, P. 5. M. Tripathi and S. P. Mishra*, Central Fuel Research Institute, P.O. F.R.I. 828 108, Dt. DHANBAD (Bihar). SUMMARY Different iron mineral species, characterised by Fe Mossbauer Spectroscopy, in co=Us or different geological origin include pyrite (Lalmatia, Bapung, Dali), pyrrhotite (Bapung), goethite (Laimatia, Bapung), iron sulphate viz. jarosite (Tiponfi) and szomolnokite (Dali), magnetite (Lalmatia) and clay illite (Dali). (Key words : Coal; iron-bearing minerals; Mbssbauer Spsctroscopy) I. INTRODUCTION : Iron minerals in coals play significantly important role in governing the slagging and fouling behaviour of coals during combustion. In view of this, detailed and accurate characterisation of iron minerals in coals of different geological origin was undertaken. In continuation of our previous work , extended studies on iron mineral species present in some typical coals of North-Eastern region, Rajmahal, and Jam mil coalfields, charac- terised by Mossbauer Spectroscopy, are reported in this paper. IL EXPERIMENTAL : Powdered coal samples (200# mesh), sandwiched between two thin beryllium sheets in a sample holder, were used as absorbers for recording the Mossbauer spectra in transmission mode at room temperature (298 °K)v. Other experimental details are described in our previous commu- nication . The spectra were analysed using the well-known Mossbauer parameters viz. isomer shift, qudrupole splitting, and magnetic hyperfine field systematics. I.S. values are relative to metallic iron. HL RESULTS AND DISCUSSION : The Mossbauer spectra of the coals are depicted in Fig. 1 and the computed Mossbauer parameters together with the minerals identified are included in Table-1. Mossbauer spectra of Tipong and Bapung coals contain doublet and doublet mixed with sextet respectively, in which pyrite (major phase) and jarosite (minor phase) are common to both coals and non-stoic hio metric pyrrhotite (with less $ = 0.48, responsible for complexity of the spectrum ) and super- paramagnetic goethite (H ^390 KOe), both in minor phases, are present only in Bapung coal. As regards the presence of pyrrhotite in Bapung coal the formation of pyrrhotite during coalification is most unlikely, because . this coal is rich in pyrite ( >4 %) and pyrite can certainly not be incorporated during coalification under reducing conditions . The presence of pyrrhotite could be due to the reduction of pyrite during pulverisation of the coal sample. The presence "of jarosite and goethite indicates some weathering to have taken place, which could be responsible for the transformation, pyrite—> goethite/jarosite, during mining. Quite different from this, Dali coal contains only non-magnetic fractions like pyrite (major phase) and clay illite (minor phases). In sharp dissimilarity with other coals, Lalmatia coal evinces quite different and complex Mossbauer spectrum (superimposed doublets and sextets), * Department of Chemistry, t Banaras Hindu University, VARAN A3I 221 005 ( U. P. ) NT - 01.1 TABLE - 1 MQSSBAU ER PARA METERS AND IRON MINERALS IDENTIFIED IN TIP ON G, BAPUNG, LALMATIA AND DALI COALS Coal Mossbauer Parameters Mineral identified i.s. (6) q.s. (^)" h. m.f. (H) m m/s ±0.01. m m/s ± 0.01 K Oe TIPONG 0.33 0.65 - p y r i t e 0.41 1.15 - jarosite BAPUNG 0.33 0.67 - p y r i t e 0.40 1.10 - jarosite '0.48 0.19 281 p yrrh otite C.35 0.20 363 goethite LALMATIA 1.14 1.91 - sid erite 0.50 0.33 330 goethite 0.56 0.19 464 m agnetite DALI 0.30 0.62 - p y t i t e 1.26 2.65 - szomolnokite 1.10 2.50 - i 1 1 i t e the iron minerals identified being siderite (dominant phase) and the magnetic component magnetite and superpara magnetic goethite (M=33O K Oe) in minor phases. The presence of the latter two minerals is ascribed to some weathering/ oxidation. The asymetry observed in the absorption of pyrite in the Mossbauer spectra of'coals is attributed to the presence of jarosite, goethite and szpmolnokite REFERENCES P. S. M. Tripathi, L. C. Ram, S. K. Jha and S. K~ Rao, "Advances in Coal Chemistry", N. P. Vasilakos (Ed.), Theophrastus Publications, Athens, Greece, pp. 307-319 (1988). A Govaert, C. Dauwe, P. Plinke, E. De Grove and J. De Sitter, J. Phys.(Paris) (Colloq.), 37(C-6), pp. 825-827,(1976). L. F. Power and H. A. Fine, Min. Sci. Eng., 8, pp. 106-128, (1976). F. E. Huggins and G. P. Huffman, "Analytical Methods of Coal and Coal Products", VoLHJ, Clarance Karr,Jr.(Ed.), Academic Press, New York, pp. 371-423, (1978). NT - 01.2 TRACK LENGTHS OF 308Pb IN DIELECTRIC SOLIDS S. Ghosh, J.Raju and KJL. Dwivedi Depairtment of Chemistry, North-Eastern Hi:il University, Shillong-793 003. SUMMARY Track lengths of 17.1MeV/u 208Pb ions in nine different dielectric materials have been determined and compared witji the theoretical data. (Key Words- 17.1 HeV/u, 20BPb, Maximum etchable Track length, Track Detector) I. INTRODUCTION One of th<: most essential data required for any study based on the Nuclear Track technique is a knowledge of the maximum etchable track length /1-6/. With an increasing variety of track detectors now available, it becomes imperative to know the maximum etchable track length of any ion in different solids and the variation in range with different chemical compositions and density. With this aim we have measured the maximum etchable track lengths of 208Pb ion at an energy of 17.1 MeV/u in nine detectors and corap&red them with two sets of theoretical data obtained from the computer codes (i) 'DEDXT' 111 and II EXPERIMENTAL fti. Irradiation- Nine different detectors -eight plastics and one crystalline detector -were irradiated by a well-collimated beam of 208Pb ions having an energy of 17.1 MeV/u. The irradiations were done at the X0- channel of UNILAC, Darmstadt with an J en flux of -^104/cm2 at an angle of 45° to the detector surface. The chemical composition and other details of the detectors are given in Table 1. b) Chemical Etching- The suitable etching conditions for the detectors ore listed in Table 1. Tne detectors were etched till tips were rounded and the time for complete etching was determined in this way. c) Measurement of Track Parameters- All measurements were done with an optical microscope ( Leitz, 'Laborlux r'(. The track diameters were measured as the minor axis of the ellipse formed by the track face while the track length* ware measured from tip to tip. From measurements of the projected track lengths the maximum et.chable trick lengths were derived tilling the formula given by Dwivedi and Mukherji (5) . d)Experimental Errors- The accuracy in the measurement of track length waa ±l«12um. The standard deviations obtained from track length distribution curves are given in Table 1. XXX RESULTS AND DISCUSSION the maximum etchable track lengts of 208Pb in different detectors were measured and the values tabulated in Table 1. Fig.l shows a typical track distribution curve for 208Pb ion* in CR-39. The standard deviations are found to vary between 2.2-4.5 um. These detectors can be used for various purposes like windows and backings for nuclear physics experiments, for preparation of microfilters and for particle identification and for all these purposes a knowledge of the maximum etchable track length is essential. NT - 02.1 Table 1 also gives a list of the Pb in CR-39 theoretical values obtained from the two computer codes 'DEDXT' •nd 'TRIM', while the 'DEDXT' valuer were found to be quite comparable, the 'TRIM' values were found, to be consistently higher by more than 10%, varying with the type of detector used. Before anything can be said conclusively about the theoretical values more experimental data are required for verification. Further work in this direction is in progress. Fig.l Track length iym) Table 1 SSKTD aod Deader Colour lo/.ll Thror* Dcorr I.1SJ in N»on/55»C/120 sin*. 153.40 x >.* C,K,O, TTl«I*l'iM 1.300 333.02 1 J.I J.i.S 357.0 C3o"l2°S Yellow in M«aK/naC/311 JJS.O 311.0 Cu"l4°3 Colourl... IK K.on/M-c/100 »lo.. 111.1) > 1.1 Cll"l»°) D1.J1 t i.) 310.0 3t«.O 317.0 337.0 ::>.o ><•.a Ccllglot* Hltr»t» C0iovrl.il 3D H.oK/ii'C/HO 333.30 t 4.3 Hlc« i;i. .01 3.3 IV REFERENCES 1. R.L. Fleischer, P.B.Price and R.M. Walker, Nuclear Tracks in Solids (Principles and Applications), University of California Press, Berkeley, London (1975). 2. E.V. Benton and R.P. Henke, Nucl. Instrum. and Meths., 5£, 241 (1968). 3. A. Murakamani, Nucl. Instrum. and Meths., 111. 567 (1973). 4. J.Tripier, G. Remy, J.Ralarosy, M. Debeauvais, R. Stein and D. Huss, Nucl. Instrum. and Keths., 1^5, 29 (1974). 5. K.K. Dwivedi and S.Mukherji, Nucl. Instrum. and Meths., 161, 317 (1979). 6. A.Saxena, K.K. Dwivedi, E. Reichwein and G. Fiedler. Nucl. Instrum. and Meths., B_36, 276 (1989). 7. K.K. Dwivedi, Nucl. Tracks Radiat. Meas., lj[, 345 (1988). 8. J.P 3iersack and U.G Haggmark, Nucl. Instru.-n. and Meth., 174, 257 (1980) NT - 02.2 PERFORMANCE STUDY OF MODIFIED GAS FLOU PROPORTIOAL DETECTOR A.G.Rafi Ahmed, M. S. B^Singh,R.V.Subba Rao, A. Ravi .ankai- and G . R . Bal .lsubi nmanian Reprocessing programme, IGCAR ,Ka]pakkam-6()3 10 2. A tnultiwire gas flow detector way fabricated with certain modi fi cat ions in design at our centre.This necessity arose due to the fact that loop type detector often gets distortion in geometry resulting in extreme variations in electric field which effects avalanche process.The end insulation on either side which is indeed significant, has been provided with perfex.Head amplifier ECIL make type HA 502 B haw been skillfully utilised, and at present study with Faat preamplifier to produce still better reaults is undertaken.So far,investigations on mingle as well as double anode systems have been accowplish«d.Though the efficiency is comparable to convent i otial loop type aysst em, the reproduciblity,piateau length ami slope are far better than the old one.A brief information is as below. Single wire Anode Double wire Anode fVolts) CVolts ) Operating Volatage 1040 J160 Plateau Length 80 120 Slope 1 .8'i 1 . 2 6 \ (per «(J Volts) (per 100 Volts) Ef f i ci ency 3 6.08'. 16.03*. Reproduci billty Good Very Good A sketch of the detec:lor fabricated by us is gi veri in fig. 1 CONCLUSION The performance ot double wire anode system ot this detector ia found to b« better than single wir« ones and much better than the loop tsydtein.A detailed study u^ing various nuclides is in progress. NT - 03.1 l.KNOB FOR CLAMP 6.HARDENED PIN 2.CLAMP 7.VERTICAL POST CLAMP 3.VERTICAL POST 8.GAS MULTIFILICATION CHAMBER 4.NOZZLE FOR GAS FLOW 9.PLANCHET HOLDER 5.SPECIAL SCREW 10.TURNING KNOB Sketch of modified gas flow detector (fig.l) NT - 03.2 SSC - Solution and Solid State Chemistry of Actinides Papers : SSC - 01 to SSC -37 EXTRACTION OF U(VI) ,Pu(IV) ,An(III) ,Pn»(III) , Zr(IV) ,Ru(III ) AMD Pd(II) FROM NITRIC ACID MEDIUM BY MIXTURE OF OCTYL(PHENYl,)-N,N-DIISOBUTYLCARBAMOYL METHYL PHOSPHINE OXIDE AND TRIBUTYL PHOSPHATE IN DODECANE J.N.Mathur, M.S.Murali, P.R.Natarajan Radiochemistry Division and L.P.Badheka,A.Banerji Bio-organic Division Bhabha Atomic Research Centre Bombay-400085. India. SUMMARY: Extraction of actinides(U(VI),Pu{IV)and Am(III)) lanthanide Pm(IIl) and other cations of relevance in the PUREX process high level acidic waste(HLAW) solutions has been carried out with a mixture of octyl(phenyl)-N,N-diisobuty1carbanoy1 methyl phosphine oxide(CMPO) and tributyl phosphate(TBP) in dodecane. The effects of CMPO,TBP and nitric acid concentrations on the extraction of these metal ions have been studied.An ideal experimental condition for the removal and separation of the actinides and Pm has been worked out. (Key words: Extraction, CMPO, Actinides, Pm, fission products.) I INTRODUCTION After the successful deployment of the PUREX process for the separation of plutonium and uranium from the irradiated nuclear fuels, one of the most difficult tasks before the chemists all over the world was to remove the long lived actinides(Am ,Cm etc.,) from the HLAW solutions. The reagent mixture of CMPO and TBP in aliphatic diluent dodecane has many advantages over other reagents. Horwitz et al.(l,2) and Mincher(3) have studied the extraction of actinides by mixtures of CMPO and TBP from nitric acid solutions. In the present work besides the extraction of actinides, the extraction of Pm(III),Zr(IV), Fe(II I),Ru(III) and Pd(II) from nitric acid solutions by mixture of CMPO and TBP in dodecane has been studied. II EXPERIMENTAL CMPO has been prepared by a modified three-step procedure described by Horwitz et al.(4). It was purified by mercuric nitrate/potassium cyanide method and also by successive treatments with macroporous cation(H ) and anion (OH ) form _of the resins(4). The purity of 'CMPO was checked by IH and H, P NMR spectroscopy and by determining D^,,, at pH 2.0(HNO3). Low D^m at pH 2.0 (<0.03) with 0.2M CMP0+1.2M TBP in dodecane confirmed that the purified CMPO is free fisPm ^he-acidic impurities. The radioactive tracer solutions of Am, ' U, Pu and Pm were purified and .prepared by wellknown procedures.The tracers Zr, Fe, and Ru were procured from BRIT of DAE(India). Pd was prepared by irradiating ammonium t^trachloro palladate in the Apsara_research reactor of BARC. The radioactive tracer solutions of U, ' Pu^and Q5 Pm^re assayed b^Q 1 iquid^seint i 1 1 at ion counting and Am, Zr, fiu. Fe and Pd(as Ag, which is in secular equilibrium with Pd) were counted with a well-type scintillation counter using Nal(Tl) detector. SSC -01.1 Ill RESULTS AND DISCUSSION The extraction of U(VI),Pu(IV),Am(III),Pm(III),Zr(IV), Fe(III),Ru(III) and Pd(XI) was done at 3.OM HNO3 with a mixture of 0.2M CMPO and 1.2 M TBP in dodecane. The order of the distribution ratio (D)(table 1). could be explained on the basis of „ effective ionic charge in some cases' Pu (+4.O),UO2 (+3.2)and Am or Pm (+3.0). The D values for the above mentioned metal ions were also determined at varying concentrations of TBP (0.6 M -1.4M) while keeping the HNO3(3.O M) and CMPO (0.2 M) concentrations constant. It has been observed that the change in the TBP concentration does not affect the D values of the metal ions,thereby confirming that TBP is not one of the moeities in the species extracted into the organic phase. The CMPO variation experiments (0.01-0.2 M) at fixed HNO3(3.O M) and TBP (1.2 M) concentrations have shown that the D values increase with increasing CMPO concentration for all the metal ions studied. This indicates the species extracted with varying number of CMPO molecules attached to each metal ion under different CMPO concentrations. From these studies it could be concluded that the actinides and Pm(III) could be efficiently extracted from HLAW solutions (which has acidity between 2-3 M HN03) by 0.2 M CMPO + 1.2 M TBP in dodecane leaving behind most of the fission products present therein. IV REFERENCES 1. E.P.Horwitz, D.G.Kalina, H.Diamond, G.P.Vandegrift and W.W.Schultz,Solv.Extr.Ion Exch.3,75 (1985) 2. W.W.Schultz and E.P.Horwitz, Sep . Sci . Technol , 2JJ, 1191 (1988) 3. B.J.Mincher, Solv.Extr.Ion Exch. 7,645 (1989) 4. Ralph C.Gatrone, I. .Kaplan and E.P.Horwitz, Solv.Extr.Ion Exch.5,1075 (1987) TABLE 1 Distribution ratios of metal ions at various nitric acid concentrations. [TBPJ= 1.2 M, [CMPO] = 0.2 M Metal HNO3 ion PH 2.0 0.2M 0. 5M 1.OM 2 .OM 3.0M 4.5M 6.0M 2+ uo2 - 77. 1 115 .7 145 . 1 - 240.0 265.7 Pu* + - 238.7 1045 .3 1647 .8 - 99 .0 3500.2 — Am 0 .03 2:07 7 .74 14 .91 23 . 12 22 .5 22.25 19. 1 Pm3+ - 1.63 5 .72 11 .06 17 . 19 19 .88 20.49 21. 5 ZrA + - - - 0 .83 0 . 91 0 .92 0.99 1. 52 Fe3+ 0 .002 0.008 0 .02 0 .04 0 . 14 0 .37 1.37 7. 54 Ru 0 . 14 0. 25 0 .27 0 .25 0 . 21 0 .20 0. 15 0. 08 Pd2+ 0 .28 0.25 0 . 17 0 . 10 0 .04 0 .02 0.01 0. 004 SSC - O'l .2 STUDIES ON THE EXTRACTION OF SOME ACTINIDES FROM ACIDIC MEDIUM BY DIHEXYL- N,N-DIETHYL CARBOMYL METHYLENE PHOSPHONATE (DHDECMP) K. V. Luhithakshan , P. S. Nair, K. Raghuramarir A. V. Jadhav and H. C. Jain Fuel chemistry Division* Bhabha Atomic Research Centre, Trnmbay, Bombay - 85. SUMMARY studies on the extraction Am(IIl>, Cm< III) and U(VI) from HNO^HCl and H2SO4 solutions by dihexyl~N,N-diethyl carboroyl methylene phosphunate< DHDECMP) was carried out and it was found that at higher acidities the extraction was quantitative. Th« effect of diluent on the «xtraction of Am activity of nuclids in the aqueous phase/ml All the egu'l ibrat ion w«re done in duplicate. SSC - 02.1 IV RESULTS AND DISCUSSION The effect of diluent on the extraction of Am(TTI> from 2H HNQ, by 30*% DHDFCMP was studied. Tt was observed that, only in the case of n—dodecrane second organic phs^e was fm nwrt. The diluent UKP><1 and the corresponding 0 values are given in the Table I.The data shows that there ie a decrease in D value from hexane to cat b->n tetrachlor ide. The distribution ratio of Cm(ITT> from ?M UNO,aw ;, function r»f DHDROMP was studied. The log-Ing plot of D vs DHDECMP revealed that th;N Cm S£>«c-ies «xtrarte Diluent Cyclohexane Benzene Toluene Xyl^ne Harbontetra- chloride DA.n< TIT) 3.6 1.43 1.12 0,75 0.45 Table-II KK t r «»<:.• t ion of metal ions into benzene by DHDECMP as a function of acid Distribution ratio HC1 b KxS04 M/I. VT> M/[. M(m; Hindu,; 0.1 0.04 0.035 0.09 0.1 0.0039 0.0067 0.06 .05 — 4. 48 l.O 0.0015 0. 003 — 0. 1 0. 05 8. 80 2.0 — O.OO4 O. 143 2.O 0.00038 0.003 1 0.085 2.0 1. 04 35 .2 3.O 0.0036 — 3. O 1.59 36 .O 4.0 - - 1,86 4.0 - 38 .6 5.0 0.018 0.042 4.0 0.004 0.0072 0.012 5.0 2. 3 -- 6.0 — O.370 11.12 — 6.0 -- -— 7.0 1.267 -- . — 6.0 0.53 1.O64 0. O47 7.0 -- 43 8.5 4.707 7.95 27.44 8.0 1.44 6.79 0.43 9.0 5. 76 -— 10. 7 17. 5 57 . 7 ». Extraotanl concent rat ion 30% b« Exttactant concentration f>%. V. REFERENCES 1. T. H. Siddal Tr. U.S. paten 3, 243< 1966) T. H. Siddal JouJ. of lnorg. Nucl. Chem., 25,883< 1963) and 26. 199K 1964) E.y. Hoitiz, W. W. Hcliuiz, Symposium on Radiochemistry and Rad!ation chemistry.. Kalpakkam , .Ian 4-7.1989. A. V. Jadhav, V. K. Goyal, 8. N. Pat tana ik. P.S. Bankaran. S K. Patil, Joul of Raadioanal. and Nucl. Chem., Ar ticle 82/2. 229< 1984). K. V. Lohithakshan. P.S.Nair, K. Haghuraman, A. V. Jadhav and H. C. Jain,. Symposium on Radiochemietty and Radiation chemistry.Naypur, 199O . SSC - 02.2 EXTRACTION OF PLUTONIUM FROM PHOSPHATE CONTAINING NITRIC ACID SOLUTIONS USING DHDECMP AS EXTRACTANT V.B.Sagar,S.M.Pawar,A-R.Joshi, U.M.Kasar * and C.K.Sivaramakriahnan Fuel Chemistry Division Bhabha Atomic Research Centre, Bombay -400 085, India SUMMARY Distribution data -for the extraction o-f Pu(IV) by DHDECMP (Di-hewyl,N-N-diethylcarb-moylmethylphosphonate) in xylene -from aqueous nitric acid and its mixtures with sulphuric acid and phosphoric acid were obtained to explore the feasibility o-f recovery of Pu(IV) -from analytical waste generated in the laboratory. Based on the data obtai ned, coridi t. i ons -for the recovery of plutonium Are suggested. (Key words: Plutonium, Phosphate waste, DHDECMP) I INTRODUCTION During the routine analysis of uranium in plutonium bearing fuel samples by the Davies ?< Gray method , large volumes of phosphate wastes are generated from which Pu has to be recovered. A method using DBPECMP II EXPERIMENTAL Inn exchange purified plutonium was adjusted to Pu(IV) and used for the extraction studies. DHDECMP was obtained from Columbia Organic Chemicals Co. Inc. Camden B.C. USA. Since the purity of this compound was not known, it was purified by hydrochloric acid hydrolysis method Distribution ratios of Pu(Iv') were determined from aqueous media of desired composition using 107. DHDECMP-xyl ene <»s extractant. Details of the experimental procedure have been reported earlier . SSC - 03.1 Ill RESULTS AND DISCUSSION Distribution ratios IV REFERENCES 1. W.DAVIES, W.GRAY, Talanta, 11 (1964) 1203. 2. V.B.SAGAR,S.r1.PAWAR,M.S.0AK and C. K. SI VARAMAKR I SHNAN Paper CT-24,DAE Symp. an Radiachem. & Radiation Cheii. Bombay,Feb. 22-26, 1988 3. W.W.SCHULZ and L. D. Me ISSAAC,Rpt. ARH-A--263, 1977. TABLE-1 : Distribution ratio (D), for Pu(IV) Aq. Phase: 2M UNO^ + Varying cone, of H-,SO4 CH2SO4.lr1 0.25 0.5 1.0 1.5 2.0 I) 145 4B 14 7 4 TABLE-2 Distribution ratio (D), for Pu(IV) Aq. Phases 2M HNOX + 1M -?S04 + Varying cone, of Phosphor i c aci d 0.25 0.5 1.5 2.0 D 3.5 1. 35 0.15 8.08 TABLE-3 Distribution ratio l.HNO3JM B. 5M 1. 0 2. 0 3. 0 4. 0 5. 0 D 0. 12 0. 18 0. 26 0. 3 0. 3 0. 3 SSC •- 03.2 Studies on the extraction of piutonium(IV) from aqueous oxalic at; i d — n i i r i <; acid solution.-. by Aliquat—336 in presence »l t rival KIII cations Al" , Fe [.(,'.Pius, M.M. Clifiryu ? u , and 0 . K . Si varamak r i shnan Fuel (Mit'inisl ry Division, HARC, Bombay, INDIA /«00 0Mr> .SUMMARY The extraction behaviour of Pu(lV) have been invest i^rttud using a liquid ariion exchanger- Aliquat-336 in xylene from on.ilic acid-nitric acid mfd i urn in presence of large excess of <: Table Variation of distribution ratio of Pu(IV) with aluouniua/f erric nitrate in 0.05 M oxalic acid-1 N nitric acid solutions Organic phase: 20% Aliquat-336 in xylene S.No. [Al^/Fe3'] Distribution ratio [M] With aluminium nitrate With ferric nitrate 1. 0.0 0.9 0.9 2. 0.05 3.3 140 3. 0.075 4. 0. 1 5.3 405 5. 0.2 8.7 805 6. 0.4 37 1340 210 2650 SSC - 04.2 MOrELLIFG OP EQUILIBRIUM DATA FOR THE LIOUIB-UOUID EXTRACTION OP METALS UHAJf1VE-L2m>:- • TO?Ofi'II?O/TB? SYSTEM K.H. Eareendrant ?uahpara.1a. P.V. RaTlndraa » and J.V. Abraham Health Physics nviaion lhabha Atonic Research Centre Bombay 400 085, India • Uranium Extraction Division • Analytical Chemistry Division SUMMABT Equilibrium data have been obtained for the extraction of Uranium (vi) Troa phosphoric acid by di (2-ethylhexyl) phosphoric acid (T2EI:PA) in the presence of neutral orranophosphorous comnounds. The data were corelated using Semi-eapirlcal models assuming analogy betveen pxtrnction and aBsorption of a solute by a solid fron PJL3 phuae. The fits are good. The d-sta are useful in predicting orgaric phas^ loedinp: for a given aqueous uranium concentration* (Key words : Vrcn^un, Equilibrium '-ta. Modelling, Orpranopnosphorous Coitpounds) I. IJITrOT/UCTIOTT tfrar-ium (vi ) is extracted from nhosnh'-.ric iicid using D2«T?A (VT)- TOPO fror. kerosene nedium/l^. The reversible reaction is assumed to be t U0 2 + + 9 ^ 2 (H X) + TOPO, x -i U05 lij. ,X9n • 2H 2 (B) h(o) III. EESULTS AND DISCU3SI0N Equilibriua data wei*- obtained by varying aqueous to organic phase ratio and equilibrium isother^a were plotted. ..2 SSC - 05.1 The data were then fitted using the seml-eanirical Lftngmuir and Preundlich isotherms whioh are In fonn i Y m and Y «• respectively. There 9.. end"O2 are the parameters of the models and x and y are the fiaueous and organic phase concentrations. As can be seen froon the Fig. 1, the fits are good Indicating the assumed analogy between extraction ieothem and Langrauir end Preundlich adsorption Isotheras Is fairly oorroct. "he values of the parameters of the two models ere also given for a typical eyetoo (Table-1) which can be used to predict the organic phase loading for the given aqueous uranium concentratio*. 1. F.J. Hurst and *'.J. Crouse, in 'Actinide Recovery from r&ete and Low Grade Sources', Eds. J.D. Kavrettl and V;.T;. SchulE, Uarwood Academic Press, New York, 1982 2. D.A. Bills, Ind. and F.ng. Ohem. t 52, 251-252, I960 3. J.A. Golding, S.N. Tlievsar end Pushoeraja, presented in Second V'orld Conpre^s of Chemical Engineering, Oct.4-9, Canada, 1981 TA^LE-1 Experimental and Predicted Tallies )- raniA-Tcpo DZE 1PA Y < red.) Y 'Pred.) [111. lr ?rendlioh 0.5 0.90 0.9184 0.9632 0.91 1.39 1. '5893 1.3797 1.42 1.76 1.7917 1.8019 r 1.64 1.97 1.9251 1.9645 Oo. 1.79 2.01 2.0059 2.0735 Langmuir fit 91 -3.7076, 82 - 0.6585 Freundllc 'fcit e -1.4600 9n - 0.60 0 FOFU CONL AU PHA'j£ I g/ij SSC - 05.2 SYNERGIC EXTRACTION OF PLUTONIUM(IV) FROM AQUEOUS HC1 BY MIXTURES OF HDEHP AND TOPO D.G. Phal and V.V. Ramakrishna Fuel Chemistry Division Bhabha Atomic Research Centre, Trombay, Bombay 400 085 SUMMARY Synergism was observed in the extraction of Pu(IV) by mixtures of di-2, ethylhexyl phosphoric acid (HDEHP) and tri-noctyl phosptiine oxide (TOPO) taken in the diluents n-dodecane, toluene or chloioform when the aqueous medium employed was hydrochloric acid in place of sulphuric acid where antagonism was reported earlier. The composition of the species responsible for synergisro and the constants for the equilibria involved are given. [Key Words : Solvent Extraction, Synergism, Antagonism, Diluent effect, Species Composition, Equilibria) I. INTRODUCTION Tn the extraction of Pu(IV) from aqueous H2SO4 into mixtures of HDEHP, the monomeric and dimeric f01 ins of which may be represented as HY and H2Y2 rebpectively, and TOPO antagonism was observed . With HNO3 as the aqueous medium, however, synergism was exhibited' . The Py(lV) species extracted into.HDEHP were al shown to be PuY2(HY2)2 *d Pu(N03)2(HY2)2) when the aqueous media were sulphuric and nitric acids, respectively. Antagonism was probably due to the interaction between HDEHP and TOPO^ ' and the lack of any adduct formation between TOPO and PuY2(HY2>2 due '° steric hindrance caused by the 2-ethylhexyl groups crowding the Pu(IV) ion. Probably due to the reduction of such a crowding in Pu(NO3)2(HY2)2 adduct formation with TOPO could take place resulting in synergism. If this is so, synergism should be observed with hydrochloric afi4 as aqueous medium where the species extracted were suggested to be PuClY(KY2)2 anc* PuCl2(HY2)2* The results obtained on this investigation from aqueous HC1 are reported here. II. EXPERIMENTAL Distribution ratio (D) measurements of Pu(IV) weKGg carried out using a plutonium tracer mainly constituted of Pu. The alpha activity of both the aqueouF and organic phases, was assayed by using liquid scintillation counting technique. III. RESULTS AND DISCUSSION Synergism was observed in the extraction of Pu(IV) from aqueous HC1 by mixtures of HDEHF and l'O.'O taken in any of the diluents n-dodecane, toluene or chloroform. Representative data are given in Table 1. If D^g is the distribution ratio of Pu(IV) with a mixture of HDEHP and TOPO and, DA and Dg are those with the reagents HDEHP and TOPO respectively, when they are used s''.: - O6.1 alone but at the same concentrations as that in the mixture, it follows that : D AB » DA + DB ± ...(1) A + A. D indicates synergism and -AD antagonism. Also, the variation of +AD as a function of the variables involved reveal the composition of the species responsible for bynergism. The log ( ,A-D) - log [TOPO] and log (AD) - log [H2Y2] plots gave slopes close to unity and the log (AD) - log [H ] plots gave slopes of minus unity, in ell the three diluents studied, suggesting that the equilibrium responsible for synergism is 4+ - AB Pu H ...(2) 3Ci + H2Y2 + TOPO ^==.=*» PuCl3(HY2) .TOPO <0) (0) (0) The average values of log KA5 =4.85, 3.81 and 1.87 were obtained for the diluents n-dodecane, toluene and chloroform, respectively. This species might have been formed in the organic phase according to PuCl2(HY2)2 + TOPO.HC1 ^===* PuCl3(HY2) TOPO + H2Y2 ...(3) (0) (0) (0) (0) The I*ck of direct adduct formation between PuCl2(HY2)2 and TOPO suggests that apart from steric factors other structural aspects seem to be involved in forming a stable species, IV REFERENCES 1. D.G. Phal, S. .'Caiman and V.V. Ramakrishna, Symp. Radiochem. and Radiation Chem. Nagpur, 1990, Paper AL-30 and Bombay, 1988, Paper CT-21 (1A), 18 (IB) and 20 (1C). 2. C.F. Baes, Jr. J, Tnorg. Nucl. Chem., 24, 707 (1962) 3. D.G. Phal and V.V. Ramakrishna, Ann. Convention of Chemists Bodhgaya, 1990. Table 1 - Distribution ratio data on the extraction of Pu(IV) Aqueous medium = 2 M HC1; Diluent = n-dodecane [H Y J - 1.0 x 10-3 2 2 M DA 0. 113 [TOPO] M x DAB AD 2.0 0.638 0. 119 0.519 1.09 0. 145 0.945 6.0 1.58 0.205 1.38 8.0 2.30 0.345 1.96 10.0 3.02 0.418 2.60 SSC - 06.2 SOLVENT EXTRACTION AND SPECTROPHOTOMETRIC STUDIES OF URANIUM (VI) WITH HBXAMETHYLENIMINE - CARBGDITH1DATB, Narendra S. Dongre, Manjirl N. Pathare, G.S. Bhat, Dr. A.D.Sawant. Inorganic and Nuclear Chemistry Laboratory. The Institute of Science, h 15, Madam Cama Hoad, Bombay-400 032. : '••'••••• + SUMMARY':' Hexavalent Uranium in Uranyl ion U0 a forms 1:2 complex with Hexamethylenimine carbodithioate of its Sodium Salt, (NaHMICdt) and was extracted into chloroform from aqueous solution of pH 5.0 Xmax was found at 330 nm and Beer-Lambert's law is obeyed in concentration range of 0.5-50 ppm. 'Key Words: Solvent Extraction of Uranium - Carbodithioate.) -INTRODUCTION: The behaviour of ligand having seven membered non-planer • «>terocyclic ring system linked to carbodithioate group is expected to give 'iranyl complex with much of steric hindrance. Recently1 it has been used for *he extraction and AAS determination of some of the toxic trace metals. Some of the complexes of NaHMICdt have been synthesized and characterised2. For the first time, we have used this ligand for the synthetic and analytical applications to the Actinides. U02 (N03 )2 gives yellow colour complex of the composition [UOa (HMICdt)a], which can be extracted over 99% in chloroform at pH 5.0. The method is applicable for the separation of Uranium and Thorium and Spectrophotometric determination of Uranium at trace level from a variety of samples. EXPERIMENTAL : Apparatus: Spectral measurements were made on Shimadzu U.V. - Visible spectrophotometer and ELICO Digital pH meter (Model LI-120) was used for the pH measurements. Reagents: The stock solution of 0.02 M Uranium was used by Uranyl nitrate UOa (NO3 )a .SHiO (BDH-England) in double distilled water and was standardised using standard method3. HC1 and NH^ OH were used for pH adjustments. 4 NaHMICdt.2Ha0 was prepared by the reported method and 0.01 M solution was prepared in double distilled water. All other chemicals used were of AR grade. Recommended procedure for extraction: The suitable aliquote (l ml) of 0.02 M Uranium solution was taken ir.to 50 ml beaker to which was added 10 ml of 0.01 M reagent solution and pH of the mixture was adjusted to 5.0. The mixture was transferred into 125 ml separating funnel and equilibrated \vith 15+10 ml of chloroform for 2 minutes. The chloroform layer was separated for spectral measurements. RESULTS AND DISCUSSIONS Absorption curve: The absorption curve was measured on spectrophotometer and Xmax was found to be at 268 nm. However for all other studies measurements were taken at 340 nm. SSC - 07.1 Effect of pH: Keeping other factors constant, extractions were carried out between pH 1.0-9.0 and was found maximum at pH 5.0. Beer's Law, sensitivity and composition of the extracted species: Beer's law was found to be obeyed in the concentration range of 0.5-50 ppm at 340 ..TI. The metal-ligand ratio in the extracted species was determined by Jobs' and Mole ratio method as 1:2 and the complex can be formulated as [UO4 (HMICdt)a ]. Elemental studies were carried out. Fairly a large number of transition elements were found to be not interfering. Hence the said method is being applied for the estimation of Ucanium at trace levels and its separation from Thorium. REFERENCES : /I/ A.K.Singh; B.Puri; R.K.Rawlley Microchemical Journal 3T_, 221-224(1988). 12/ A.K.Singh; B.Puri; R.K.Rawlley. Ind.J.Chemistry 28-A, 59-62(1989) /3/ A.I.Vogel, "Text Book of Quantitative Inorganic Analysis". Longman-Green and Co.Ltd., London(1961) /4/ A.K.Singh, S.Sharma. Microchem. Journal. 35, 365-368(1987). ********** SSC - 07.2 LIQUID-LIQUID EXTRACTION OF CALIFORNIUM(IIl) FROM ALKALINE PYROPHOSPHATE AND TRIPOLYPHOSPHATE SOLUTIONS V, Chakravortty , S.A. Perevalov and Yu. M. Kulyako V.I, Vernadsky Institute of Geochemistry and Analytical Chemistry, USSR Academy of Sciences, Moscow (USSR; SUMMARY — Studies on extraction of cali,fornium(III) from alka- line solutions contg. pyrophosphate and tripolyphosphate or carbonate by dioctyi pyrocatechol (DOP) in octane have been made. This element is also quantitatively extracted from pyrophosphate media between pH 4 & 5 by di-(2-ethylhexyl) phosphoric acid (HDEHPA), and back-extracted with 1 M H SO4. About 300 Mg of californium(III) has been recovered. (Key wordsi Californium(III), Dioctyi pyrceatechol, Dl-(2-ethyl hexyl) phosphoric acid). I. INTRODUCTION — There are not many satisfactory extrac- tants for extraction of actinides from alkaline solutions and in presence of complexing agents. The oxidation of californium (III) /l-3/ in alkaline complexing media was attempted and the problem of extraction of Cf^+ from such media was encountered. The present report deals with extraction of californium(III) from slightly acidic and alkaline solutions containing pyro- phosphate and tripolyphosphate complexants in presence of various chemical oxidants by di-(2-ethyl hexyl) phosphoric acid (HDEHPA) and dioctyi pyrocatechol for regeneration of californium-249 after various redox experiments due to its limited availability. II. EXPERIMENTAL — The radioelemfent 249Cf was used for solvent extraction of californium and its radiocher.ical purity was ascertained from its gamma-spectra. Sodium pyrophosphate and potassium tripolyphosphate were at a concentration of 0.1 M in aqueous solution. Purified commercial HDEHPA and dioctyi pyrocatechol were the extractants in n-heptane and n~octane, respectively. The distribution studies were carried out using glass-stoppered tubes at an organic to aqueous phase ratio of lsl and with continuous monitoring of reversibility of distri- bution. The procedure for determination of distribution ratio, D, from the 1 -activities of *^*Ct has been reported /4/. III. RESULTS AND DISCUSSION — Dioctyi pyrocatechol in octane extracts californium(III) from solutions of NaOH and Na2C0, (5-6 M) containing pyrophosphate and tripolyphosphate in presence s w tn of different oxidants like Xeo|- and 2°8 * considerable ease. complexants keep elements in tne alkaline solutions in the soluble form and Cf^+ can be extracted from alkaline and carbona- te solutions in the form of ion-associates, saturated chelates Department of Chemistry, Utkal University, Vani Vihar, Bhubaneswar-751004, (India). SSC - 08.1 and hydrated compounds. The anion part contains either the hydroxo-complexes of Cf3"*" or compounds with the complexants. Extraction of Cf from aq. pyrophosphate and tripolyphosphate media by HUEHPA (1.0 M) in heptane has been studied as a function of equilibrium pH (0-5). Extraction from tripoly- pohsphate media is less than that from pyrophosphate media at any aqueous acidity. Even 1 M H_S0. can back-extract californium(III). Extraction chromotography experiments were undertaken in order to recover 300 Mg caJifomium-249. IV. REFERENCES 1. V. Ya. Frenkel, Yu.M. Kulyako, V.M. Chistyakov, I.A. Lebedev, B.F. Myasoedov, G.A. Timofeev, E.A. Erin, J. Radioanal. Nucl. Chem.„ Lett., 104, 191 (1986). 2. B.F. Myasoedov, I.A. Lebedev, P.L. Khiznyak. G.A. Timofeev, V.Ya. Frenkel, J. Less Common Metals, 122, 189 (1986). 3. V.N. Koayakov, G.A. Timofeev, E.A. Erin, V.V. Kopytov. V.I. Andreev, Radiokhimiya, 19, 82 (!"77); Sov. Radlochem., 19. 66 (1977). 4. V. Chakravortty, S.A. Perevalov and Yu.M. Kulyako, J. Radioanal. Nucl. Chern. , Lett. 136, 85 (1989). SSC - 08.2 STUDIES ON NEW DIOXOURANIUM(VI) COMPLEXES HAVING EIGHT COORDINATION NUMBER WITH BIOLOGICAL IMPORTANT LIGANDS M.A. Pujar, K.Siddappa and A. Malleswar Reddy Department of Chemistry, Gulbarga University, Gulbarga-585106 (Karnataka). SUMMARY - Dioxouramum(VI) forms 1:1 and 1:2 (Metal: Ligand) complexes with some Schiff bases derived from biologically important ligands coumarine derivative. The complexes have been characterized by physical methods. The studies suggest that the complexes have eight coordination number around the metal ion moiety. Key Words: Dioxouranium(VI) complexes, Schiff bases, complexation, I. INTRODUCTION - A search through the literature has revealed that dioxou- ranium(VI) complexes of Schiff bases of biologically important coumarine derivatives (L) have not been reported so far, even though other Schiff bases complexes reported in recent years. |l-3|. Therefore a.e part of prog- ramme for the synthesis and characterization of unusual coordination actinide complexes, we report here ,_s±x complexes of dioxouranium(VI) derived from Schiff bases of coumarine (L ). = C6H4C1 II. EXPERIMENTAL - The 6-for;nyl-7-hydroxy-5-methoxy coumarine was synthe- sized in our laboratory and desired Schiff bases were obtained when it was condensed with appropriate substituted anilines in 1:1 ratio in acetic acid. The structures of these compounds were confirmed by analysis, IR and NMR spectral data. The complexes were synthesized by adding appropriate amount of Schiff bases to ethanolic solution of uranyl nitrate. The reaction mixture was refluxed for 3 hours. The complexes separated directly after condensing or after bringing pH7 were collected, purified, dried in vacuo over 6ilicagel. The complexes purity was checked by TLC. They were analysed for metal carbon and hydrogen. Conductance, IR, NMR and electronic spectral measure- ments were taken for the coinplexes |4|. III. RESULTS AND DISCUSSION - The complexes obtained are lister! in Tablejl together with their analytical data. The molar conductance values of 10 M DMF solutions of, 1:1 complexes have been found to be in the range 65-75 ohm-1 cm2 mole . These complexes can be regarded as 1:1 electrolytes. The others are iound to non-electrolytes having 1:2 composition. The electronic spectra of ligands are characterized by three absorption bands in the JV region. The band at 358-375 nm is assigned to charge trans- fer within the solute molecule. The rest 'nands are due to v\ — "H* and 7T re* transitions and the spectra of li.-se bases are different from those of their solutions with complexes. This can be considered as an evidence SSC - 09.1 for complex formation 121. The tentative assignment of some of the important bands of ligands and their metal complexes are recorded in Table-1. It is evident that significant changes in the IR spectra of the ligands are obser- ved on complexation. Infrared spectral data for 90H, -?C=N, S>OCH3 f i> CO conclude that the coordination of the Schiff bases with the metal ion through the azomethine nitrogen, oxygen of methoxy group and hydroxyl of ligand takea part in complex formation J3|. This is further confirmed by observing •>) M-0 and 9M-N bands which were observed in the far TP region. The symmetric stretching band (vi) of uranyl group in all the comple- xes is observed in the expected range suggesting slight distortion of the liniarity of the uranyl group. All the complexes are found to be diamagnetic indicating that all of them are spin paired. Using conductivity, electronic, IR and NMR spectral data the formula of the complexes under study can be represented as follows in which eight coordination number is maintained in all the complexes. I 1 [UO2(L )2JNO3 and lUO^L )^ Table-1: Analytical, conductivity and IR data for Schiff bases and their complexes with dioxouranium(VI). Formula % Aalysis Found/ Molar IR M.P. (Calculated]1 conduc- 9C=N C H M tance ., 1 L 69.35 4.21 1630 1280 171 T (69.15) (4.36) - X [UO?(L ) JNO 44.00 2.90 29.31 64 1620 1390 300 £ C j (44.25^ (2.81) (29.51) t1 [LO (L ) j 41.50 3.11 31.82 09 1620 1295 300 £ c. (47.44) (3.02) (31.46) T1 T1 I, 71.05 4.80 - - 1635 1285 200 Tt (71.28) (4.69) n [UO?(L )JNO, 46.04 3.27 28.68 72 1625 1298 300 C CD (45.47) (3.15) IT (28.42) 48.75 3.50 30.18 12 1625 1300 300 c c. (48.64) (3.37) (30.40) L111 62.30 3.60 - - 1635 1290 185 (62.21) (3.51) TU1 T T [UO7(L ) JNO 44.12 2.62 27.51 64 1620 1310 300 (43.96) (2.58) (27.30) (UO (L ) J 41.21 2.60 29.19 08 1625 1305 300 (41.21) (2.58) (29.09) Acknowledgement: The authors are greatful to Prof • S.P.Hi remath, Chairman of the Department of Chemistry, for the encouragement and facilities. IV. REFERENCE 1. A.O. Baghalf and N.S. Ganj., Poly. 6, 205 (1987). 2. P.Roy and P.C.Roy. Inorg. Chem. Acta., 129, 265(1987). 3. M.A. Pujar and B.R. Pirgonde, Symp. on Radiochem. Nagpur, Al-10, (1990). 4. M.A. Pujar and D.N. Satyanarayana, Trans. Chem., 13, 423(1988). SSC - 09.2 STUDIES OF THORIUM SAI.ICYLATES BY SOLVATION WITH TRIS-(2-BTHYL HEXYL) PHOSPHATE G.S.Desai and V.M.Shlnde Analytical Laboratory, The Institute of Science, 15, Madam Cama Road, Bombay 400 032,India SUMMARY A method has been proposed for the extraction of thorium (IV) from salicylate media using tris-(2-ethyl hexyl) phosp.iate dissolved in toluene as tin extractant. The optimum conditions were evaluated from critical study of j1'1. salicylate concentration, extractant concentration, period of equilibration UM. diluent. The method permits the separation of thorium from the assor.i- elements and is applicable to the analysis of thorium in synthetic mixtures. (Key Words : Thorium, sodium salicylate. tris-(2-ethyl hexyi ) phosphate:). I. INTRODUCTION Various organophosphorus compounds such as tributyl phosphate (TH1 : trioctylphosphine oxide (VOPO) and di-(2-ethyl hexyl) phosphoric acid ha--• been used for the extraction of thorium from different media. However i" systematic study has been carried out for extraction and separation of thori". with tris-(2-ethyl hexyl) phosphate.The isotope of thorium, Th232 is a source of fuel for atomic power. Thorium is also widely used in magnesium alloys. gas mantles, electronic products, refractories and polishing. So the investigation was undertaken to evaluate a rapid, simple and selective solvent extraction procedure for quantitative extraction and separation of thorium from salicylate media. II. EXPERIMENTAL Stock solution of thorium(IV) was prepared by dissolving 0.634 g of thorium nitrate in 250 nil of distilled water containing 2 nil of concentrated nitric acid.Solution was standardized by known method /!/. A 4% m/V solution of tris-(2-ethyl hexyl) phosphate in toluene was us<» III. RESULTS AND DISCUSSION The extraction of thorium(lV) was studied at various pll values (2.0-8.0). sodium salicylale concentration (0.001-0.025M) and TEHP concentration (0.25- 5.0%) It was found that 4% solution of tris-(2-ethyl hexyl) phosphate In toluene extracts thorium quantitatively from 0.005M sodium salicylate solution at pH 2.75-3.5. The optimum shaking period was found to be 15 seconds. Thorium(IV) was backstripped from the organic phase with 2M hydrochloric acid and was estimated spectrophotometrically using thoron 1 at 540 nm. SSC - 10.1 The nature of the extracted species was established using log-log plots. The lug-log plot of distribution ratio versus salicylate concentration (at fixed pH and TEHP concentration) or versus TEMP concentration (at fixed pH and salicylate concentration) yielded a molar ratio of 1:2 with respect to both extractant and salieylate. Hence the extracted species was thought to be a neutral complex of .probable composition Th (H sal)., 2TKHP, where TEMP is tris-{2-ethyl hexyi) phosphate, it causes the extraction of thorium by solvating the salicylate salt. The suitability of several solvents such as benzene, toluene, xylene, chloroform, carbon tetrachloride and hexane for the extraction of thorium using proposed method was investigated and it was found that 4% ni/V solution of TKHP dissolved in either toluene or xylene allows quantitative extraction of thorium.The extractions were incomplete with other solvents. Varying amounts of foreign ions were added to a fixed amount of thorium (45ug) to study interference in lite recommonded procedure. The tolerance limit was set at the amount required to cause ±1% error in thorium recovery. There was no interference from 2000 ug each of Pbtll). Mn(Il), U(V1), Cd(II). As(Ill), Mgdii, Hg(II) and nitrate; 1000 ug each of Cr(VI). Ni(II), Zn(II). Co(ll), Ba(ll), Au(IIl), Pd(II), Hf(iV), Ag(l), Te(JV), thiourea, sulphate, chloride, -ji-yjcrbi'.: arid and phosphate and ZOO ug each of Bi(III). V(V). Sb(lll), Aldil), Zr(IVJ but oxalate, citrate, tartrate and EDTA interfered seriously. The proposed method was applied to the synthetic mixtures containing different metal ions with Uiorium(lV). The recovery of thorium! IV) was found to be greater than 99.0%. The method is selective and permits a rt;pid separation and determination of inicrogram amount of thorium. The average recovery of thorium was 99.0*1.0% . Kach determination took t> total of 20-25 minutes. IV. REFERENCES 1. A. I. Vogel, 'Textbook of (Juantitati ve Inorganic Analysis'. 3rd Ed., Longmans, London, page 442 (19f>2). 2. Z. Marczenko, 'Spectrophotometric Determination of fileicntits' , Ellis Horwood, Chichester. page 5.'19 (197H). SSC - 10.2 SOME OBSERVATIONS ON OXIDATION STATE ADJUSTMENT OF PLUTONIUM IN HYDROCHLORIC ACID MEDIUM U.M.Kasar, S.M.Pawar, V.B.Safiar, A.R.Joshi and C.K.Sivaramakrishnan. Fuel Chemistry Division, Bhabha Atomic Research Centre, Bombay-400085, India. SUMMARY Effect of H2O2 on Pu(III) and Pu(IV) in HC1 medium has beer investigated. It was observed that up to 6 M HC1 most of plutoniun is present as Pu(III) in presence of hVjOo even after heating at 60°C. Above 8M, pluconium is quantitatively oxidised to Pu(IV) on heating It was also observed that Pu(III) could be oxidised quantitatively at 3 M HC1 with or without H202when >0.1 M HNO3 is present. (Key words: Plutonium, Oxidation states, H2O2) I- INTRODUCTION Large amount of waste containing plutonium in HC1 medium,though not very common, is occasionally generated from the nuclear fuei fabrication scrap* '. Significant quantities of Plutonium from 3-4M HC1 had to be recovered and converted to PuOo for recycling in fuel fabrication. Plutonium oxide is normally obtained through Pu(IV)- oxalate precipitation from 3-4 M 9NQ3 and its subsequent conversion to oxide by thermal decomposition* '. Change of medium from HC1 to HNO3 to recycle plutoniurn from HC1 medium would involvo additional steps generating large volumes of aqueous waste. Direct precipitation of Pu{IV)-oxalate from HC1 medium would therefore be desirable. Information available on precipitation of Pu(IV)-oxalate from HC1 medium is meager and hence preliminary experiments were carried out using Th(IV) as a stand-in for Pu(IV). These experiments resulted in the final product ThOo with desired physical characteristics. Therefore it was decided to precipitate the Pu(IV) oxalate directly from HC1 medium to obtain PuO?. i-.-jst of the piutonium in the above HC1 solution was in trivalent state and it was necessary to convert it to Pu(JV) prior to oxalate precipitation. As H20£ does not introduce any impurity,it is normally used to adjust plutonium to Pu(IV) during oxalate precipitation from HNO3 modi urn. Information on adjustment of plutonium to Pu(IV) by Ho02 in HC1 medium is very limited* '. Therefore some experiments were carried out to arrive at the conditions to adjust plutonium to Pu(IV) by H2O2 in HC1 medium. II. Stock solutions containing purified Pu in the range of 12-13 Qig/ml in HC1 of various concentrations were prepared. Plutonium in these stock solutions was adjusted to Fu(IV) electmlytical ly. One ml aliquots from each of these stuck solu,:jns were taken in separate tubes. About 15 ul of 30% H2O2 was added to each tube and mixture was heated at 60 + 1 °C for an hour with intermittent stirring. 100 ul from the resulting solution was then taken and percentage of Pu(III) in it was determined by controlled potential coulometry. Similar experiments were also carried out at 75 + 1 °C. A few experiments were carried out with plutonium solution ssc -11.1 containing mixture of Pu(III) and Pu(IV) before addition of H2O9. Determination of Pu(III): Pu(III) in the aliquot was determined by oxidation to Pu(IV) at 0.7 V vs SCE in 1 M H2SO4. Then all Pu in the aliquot was reduced to Pu(III) at 0.3 V vs SCE and total Pu was determined by its subsequent oxidation to Pu(IV). III.RESULTS AND DISCUSSION. Since t^Og interferes in the determination of Pu(III), it is essential to establish the conditions to destroy excess H9O0 prior to Plutonium determination. It was confirmed by blank runs that excess H2O9 is destroyed completely at all acid concentrations studied by heating at 60°C for an hour. Results of the experiments on the effect of H2O2 on oxidation states of Pu in HC1 medium are summarized in table-I. It can be seen from the results that the addition of HgC^ to Pu solution in .^ 6M HC1, all Pu was converted to Pu(III), while in case of >„ 8M HC1, it was converted to Pu(IV). It was also observed that initial oxidation state of Pu prior to addition of HgOg or heating at higher temperature does not make any change on final oxidation state of Pu. The results suggest that in HC1 medium, adjustment of Pu to Pu(IV) by H2O2 is possible only above 8 M HC1. However, as oxalate t>rcoipitation has to be carried out from 3-4M HC1, the solution needs dilution which v/ould result in additional waste generation. In view of this, quantitative oxidation of Pu(III) to Pu(IV) at 3 M HC1 by adding small amount of HMO3 was explored. It was observed that Pu(III) could be quantitatively oxidised in 3 M HC1 in the presence of >/ 0. 1 M HNO3 at room temperature within an hour even without addition of H2O2. IV. PKFERENCES 1. M.M.Charyulu et al. Int. Symp. on Thermochem. & Chem. Processing, Nov. 1989, IGCAR, Kalpakkam, India. 2. J.M.Cleveland, "The chemistry of plutonium", p:527 (1970) 3. P.V.Balakrishnan et al, J.Inorg. Nucl. Chem. 23. 759(1964). Table- >. S.NO. Medium Pu Cone. Temperature % of Pu(III) mg/ml 0 C Initial Final 1. 3.0 M HC1 12 60+1 <0.2 98.4 2. - do _ 12 75 + 1 <0.2 99.3 3. - do _ 6 60+1 77.3 98.2 4. 4.0 M HC1 12 60+1 <0.2 96.5 5. - do _ 12 75 + 1 <0.2 95.5 6. 6.5 M HC1 12 60+1 <0.2 87.0 7. - do _ 12 75 + 1 <0.2 82.6 8. 7.0 M HC1 12 60 ± 1 <0.2 28.8 9. - do 6 60+1 77.3 27.0 10 8.0 M HC1 6 60 ± 1 77.3 <0.2 11. 9.0 M HC1 6 60 ± 1 77.3 <0.2 SSC - 11.2 STUDIES ON EXTRACTION OF Pu F.R.Kuikar, M.S.Nagar and M.S. Subramanian Radiochemistry Division Bhabha Atomic Research Centre Tr. otnbay, Bombay-400085, India. SUMMARY The extraction behaviour of U(VI) and Fa< IV) with gamma-irradiated dibutyl derivatives of hexanamide ( DBHA>, octanamide I. INTRODUCTION Ishiara et.al. have studied the effect of extractant degradation on the extraction of U,Th.Pu and fission products with irradiated TBP and irradiated amines while radiolysis of TBP was studied in detail by Merklin in dodecane to confirm the synergistic effect of radiolytic products at higher doses/1,2.3/. In our lab. a detailed study has been carried out on the extraction behaviour of U, Pu.Zr and Ru with irradiated sulphoxides/4/. These studies have revealed that degradation products at even higher doses ate not deleterioue to the extraction process a& compared to TBP and hence sulphoxides were considered as potential extractants for fuel reprocessing even a.t higher doses. Recently, Gasparini and Musi teas have suggested dialkylamides SSC - 12.1 td determine the d isttitn.it ion ratio. Pu was maintained in the tetravalent state by the addition of 0.0b M sodium nitrite and 10 mg. of ammonium vanadate. Equal volumes (0.5 ml.) containing the tracer in 3.5 M nitric acid and preequi1ibrated dodecane phases of the irradiated amides were equilibrated for one hour at 25'-'U. Suitable aliquots of the two pi,^es were assayed for alpha activity by scintillation counting. Amide contents wei.e determined by nmv-sqiieous potent ioniptr i c titration in £>c«=>tin anhydride using perchloric acid in dioxane ae titrant after w^t-hing twice with 0.5 M nitric acid and with warm water to remove radiolytic degradation products which otherwise interfere in the titration /&/. III.RESULTS.AND DISCUSSIOM Tables I. shows that the Kd values of Pu< IV) decreases rapidly upto a dose of 35 M rads. increasing thereafter indicating synergistic effects of radiolytic products in extraction. The Kd values of U(V1) on the other hand, decrease gradually with dose indicating that radiolytic products formed do not have any undesirable effect on the extract ion. IV.REFERENCES 1. T. Ishiara and T.Tsujino, J. At.Ener. Soc. Japan, 2_, 659 < I960), J. NUCTI. Sci. Tech., 3, 144, 320 < 1966;5, 104 < 1963); 2, 463 (1965). D. I..Gief'er t» J.F.Merklin, Nucl. Inst. Methods, 128, 609 (1975>. J. P. Holland, J.F.Merklin and J.Razvi. Nucl. Inst. Methods, 153, 589 ( 1978). S. A. Pai. J.P.Shukla and M.S.Subramanian, Radiochem. & Radioanal. Let ters. 53< 1), 5 ( 1982); 74( 1/. 31 < 1982). 5. G. iM. Caspar ini & G. Gross i, Solv.Bxt.6i Ion Kxch. ,4(6), 1233 ( 1986). 6. C.Musikas, Inorg. Chim. Acta, 14, 197 (1987). 7. T. 11. LHddall, M. O.Fulda and G.S.Nichols, USAKC. DP-541 (196 1) 8. C.Wimer, Analytical Chem. , 30_, 77 (1958). TABLE-1. Kd VALUES FOR THE EXTRACTION OF Pu( TV) 6, U(VI) WITH AMIDES. O. 5 K DBOA 0. 5 M DBHA DOSE M.rad. Amide% KdPu(IV) KdU(VI) Amide% KrlPu( IV) KdU(VI) 0 98. 4 5. 1 3. 1 98. 2 5. 2 2.7 10. 2 96. 5 4.6 2.6 96. 2 4.6 2. 3 30. 3 94.4 3.6 2. 2 92. 2 3. 1 1.8 42. 8 92. 4 3.9 2.0 90.0 3. 4 1.7 72. 1 85.6 5. 5 1. 7 84. 6 4.9 1. 5 92. 1 86. 4 6. 2 1. 7 79. ? 4.8 1. 3 124. 8 94.0 8. 0 1.6 75. 3 5. 5 1.0 174.0 94.6 7.0 1.3 , 72.1 5. 3 0.8 SSC - 12.2 INVESTIGATION OF SOME DIAMIDES FOR THE EXTRACTION OF URANIUM(VI) AND PLUTONIUM D. R. Prabhu,G. R. Mahajan,J. P. Shukla,G. H. Hair and H.S.Subramanian Radiochemistry Division,Bhabha Atomic Research Centre. Trombay,Bombay 400085. SUMMARY A number of diamides were synthesised and investigated as extractants for the removal of actinide ions from wastes generated in reprocessing plants. Among these tetrabutyl malonamide( TBMA) in r dodecane was found to be most useful for the extraction cf U(VI> and Pu(IV) from dilute nitric acid solutions. Both the actinides were found to be extracted as disolvate?. Key words: Diamides.. n-dodecane, extract ion» U( VI) ,Pu< IV). I.INTRODUCTION Diamides have been .'ound to be useful for the extraction and recovery of actinides in waste btreams from reprocessing plants /Ir2/. All the work reported in literature deals with extraction study of diamides in aromatic diluents. The presnt work was undertaken to investigate the extraction of U( VI) and Pu( IV) using some diamides synthesised in the laboratory in aliphatic diluents such as n-dodecane. The diamides synthesised and characterised for the present work were tetra hexyl malonamide II.EXPERIMENTAL Malonyl chloride used for the synthesis of the dimides was obtained from M/s Merck, FRG. The diamines were obtained from M/s Fluka,Swiss. Methylene chlo/ride used as the medium of reaction was of laboratory reagent grade. PJutoniurn in solution was adjusted to tetiavalent etats by treatment with 0.01M NaN0t and using 0.005M NH4VO3 as the holding oxidant. U-233 tracer was used for U III.RESULTS AND DISCUSSION Among the listed diamides* TEMA could not be separated out because of its solubility in water. TLMA was obtained as a solid material having low soublHty in n-dodecane, the medium chosen for the extraction study. TIMA was obtained as a colourless thick 1iguid,readi1y soluble in n-dodecane. However on equilibrating 0.2M of the amide in dodecane with HNO^ of concentration as low as O.^M, third phase was found to be formed in the system. TBMA SSC - 13.1 Isolated was found to De pure (C ~ 68. 8%< 69. 9%> ; H= 11. 47%< 11. 75%) ). The IR spectrum of the compound gave the characteristic band ol: the ami da group. The diamide was found to be easily soluble in n-dodecane. With a solution of 0.4M, no third phase formation WEB noted when equilibrated Kith IN HNOg. Therefore the distribution of LK VI) and Pu( IV) was studied from an aqueous phase of 1M nitric acid. The distribution ratios IV.REFERENCES l.C. Mueikas, Inorganica Chimica Acta, 1-50, 197 (1987). 2.C.MuEikas and M. Germain. Solvent Extr. Symp. Proceedings, Vol.4, Moscow ,1988, p.124. 3.G.Thiollet and C.Musikas, Solvent Extr. Ion Ex., 7, 813 (1989) 2.0 o.o- - - 1 25 -1.5 0.94 0.28 Log [TBMA] FIG.-l. LogD0/LogDRu vs. Log[TBMA] Ru SSC - 13.2 SYNERGISTIC EXTRACTION OF AMERICIUM ( III) WITH 1-PHENYL-3-METHYL-4-BENZOYL PYRAZOLONE-5 AND DIALKYLAMIDES R. Veeraraghavan & M.S.Subramanian Radiochemistry Division Bhabha Atomic Research Centre Trombay, Bombay-400 085, INDIA SUMMARY The synergistic extraction of Am I.INTRODUCTION. Dialkylamides (A) have been suggested as alternative extractants to tributyl phosphate SSC - U.1 xylene; but two synergietc adducts of Am Variation of D with concentration of HPMBP and amide 1. HPMBP alone. 2. HPMBP +0.8 mM DBHA 3. DBHA +0.02 M HPMBP 4. DBOA + 0.05 M HPMBP 5. DBDA + 0.05 M HPMBP a a o -0.5 - REFERENCES. 1. G. M. Gasparini and G.GroBsi, Solv. Ext. & Ion Exch. ±<6). 1233 ( 1986). 2. C. MUBikas.J. C. Morrisseau, P.Hoel. B.Guillaume in "Actinide Extractants for the Nuclear Industry of the Future",Vol.1, Chem.E.Symp.Ser. , No. 103. 3. R. Veeracaghavan, S. A. Pai, M. S. Subramanian, J. Radioanal.Si Bucl. Chem.. 14_K2>, 339 (1990). 4. R. Veecacaghavan, S. A. Pai. M. S. Subramanian, ibid., (to be publAshed). 5. W.H.Baldwin, C. E. Higgins, J. M. Schmitt, USAEC Report ORNL-3679, 56 S S C - . 2 SOME NOVEL DIALKYLAKTDE ADDHCTS OF PLUTONYL PYRA7.0I.0NATES ¥'. B. Ruikar, M. S. Nagar and M. S. Subr a man i an Radiochemistry Division Bhabha Atomic Research Centre Trombay, Bombay-400085. SUMMARY Six solid complexes of Pu(VI) with 4-benzoyl(PMBP), 4-acety] ( PMAP> substituted pyrazolones with dibutyl derivatives of hex2namide( DBHA) octananiide< DBOA) and decanamide I-INTRODUCTION The aoyl pyr^zolonee,viz.,PMBP and PMAP are good chelate extractants for actinides from acid media. Okafor has investigated solid trivalent lanthanide chelates with PMBP and PMAP, while Umetani et.al. have recently used PMTFP with TOPO and CMPO for synergistic extraction of lanthanides t±>2I. N-N'dialky]amides have interesting solvent extraction properties useful for reprocessing (3—5). Little data is available in the literature on solid plutonyl complexes with substituted pycazolones and amid^:5. Their possible applications in laser chemistry, gas chcomatographic separation and as shift reaqents in NMR spectroscopy led us to investigate them. II- EXPERIMENTAL 4-substituted pyrazolones were prepared by a modified Jensen's method anB recrystal1ised twice from n-hexane /6/. Melting points, carbon, hydrogen and nitrogen contents were determined by usual methods in a glove box 111. The residual PuO_after C-H combustion' gave the plutouium content which was also determined rad iomet r ical ly. DBHA, DBOA and DBDA were synthesised by reacting dibutylarnine with the respective acyl chloride /4/. IR spectra were recorded on a PU-y512 spectrophotometer in the range 4000-200 cm"' using nujol mulls between CeI discs. A Beckmann DU-7'spectrophotometer was used for absorption spectial studies in benzene. Plutonyl complexes with SSC - 15.1 pyr 37.01 ones and amides were prepared by solvent extract ion of 10 ml. of 1 mM plutonyl nitrate solution in 0.3 M nitr ic acid with 2 mM of pytaiolonRs SSC - 15.2 LIQUID-LIQUID EXTRACTION OF AMERICIUM (III) BY BIS (2-ETHYLHEXYL) SULFOXIDE FROM AQUEOUS NITRATE MEDIA J. P. Shukla and C. S. Kedari * Radiochemistry Division, Bhabha Atomic Research Centre, Trombay, Bombay 400085. SUMMARY Extraction of Am(III) was studied from acidic nitrate modia into n-dodecane by bis (2-ethylhexyI) sulfoxide (BESO). Ca'NO3)2 was used as the salting-out agent. Effect of several wator-miscible polar organic solvents was examined for their possible synergistic effects on its extraction. Slope analysis revealed a predominant formation of the trisolvated organic phase complex, Am(NO3)3.3BESO. (Key words:americium, sulfoxide, extraction, synergi3m) I. INTRODUCTION Bis (2-ethylhexyl)sulfoxide has been by now established to be a novel extractant for uranium and plutonium/1,2/. No attempt seems to have been done to explore its application for americium extraction. The present work deals with the extraction of ameri^ium from aqueous acidic nitrate medium in presence of Ca(NO3)2 as the salting-out agent. The effect of some polar aqueous miscible additives such as acetone and acetonitrile on its extraction from different nitric acid concentrations was also examined. To reveal the structure of complex formed in the organic phase, distribution ratios were measured as function of BESO concentration, employing n-dodecane as the diluent. II. EXPERIMENTAL BESO obtained from Fairfield Chem. Co., USA was used as received. n-Dodecane (AR, Fluka) was washed with 1 M NaOH, 1 M HNO3 and finally with distilled water. Am-241 tracer was used throughout the distribution experiments. 1.0 ml of the aqueous phase with tracer containing 3.0 M Ca(NO3>2 and desired concentration of nitric acid was equilibrated with 1.0 ml of pre-equilibrated BESO/n-dodecane solution for nearly 30 mins. on a mechanical shaker. After settling for half an hour and centrifuging, suitable aliquots from both the phases were withdrawn and assayed radiometrically. Values of distribution ratios (D^m) were then calculated as the ratio of Am-241 activity in the organic phase to that in an equal volume of the aqueous phase. III. RESULTS AND DISCUSSION Extraction of Am(III) was negligibly small even from concentrated nitric acid solutions (tested upto 8 M HNO3 ). This necessitated the use of a salting-out agent such as Ca(NO3>2- As the acidity Increases from 10 M (pH<-4) to 0.8 M, the D^m values decreased rapidly after passing through a maximum at low acidity (Table-I). Probably this is caused owing to the competitive formation of * Fuel Reprocessing Division SSC - 16.1 BESO.HNO3. From a plot of log D^ vs log BESO, a slope of nearly +3 was obtained indicating the predominant formation of the trisolvated organic phase complex of the type, Am(NO3)3.3BESO; the extraction mechanism thus-can be represented as follows. 3+ Am aq + 3 N03 ag + 3 BESOorg = Am(NO3)3.3BESOorg Preliminary experiments indicated that the addition of some polar organic additives generally exerted a greater pronounced effect on aroericium extraction. Similarly, an increase in aqueous phase acidity also exerted a drastic reduction in itn recovery (Table-I) on adding 30% (v/v) acetonitrile and 40% (v/v) acetone to the acidic nitrate solutions. Americium could be easily stripped better than 95% in single equal-volume contact with 1-3 M nitric acid solutions from the loaded BESO/n-dodecane phase. IV. REFERENCES 1. Bruce Moyer, W. J. McDowell and G. N. Case, Int. Solvent Extr. Conf. Proc. p.144, Denver (1983) 2. J. P. Shukla and C. S. Kedari, DAE Symp. on Radiocheir , Radiation Chem. IGCAR, Kalpakkam, Jan 4-7 (1989). Values of D^mfor the extraction of Am(HI) by BESO/n-dodecane from aqueous acidic nitrate solutions Salt medium :- 3.0 M Ca (NO3)2 Organic phase:- 0.4 M BESO/ n-dodecane. n HNO3 D;Ira 1•'Am ^ presence of polar additives M 40% Acetone 30% Acetonitrile pH= 4 1.15 17.13 8.52 0.03 0.93 17.35 8.74 0.30 0.12 2.88 0.35 0.40 0.05 0.79 0.11 0.80 0.01 0.10 0.01 SSC - 16.2 EXTRACTION OF Am(lII) WITH DINONYLNAPHTHALENE SULPHONIC ACID IN THE PRESENCE OF KRYPTOFIX 22 P.K. Mohapatra and V.K. Manchanda Radiochemistry Division, BARC, Bombay 400 085. SUMMARY: Mechanism of extraction has been investigated for the solvent extraction system Am(III)/diaza-18-crown-6(K22)/Dinonyl naphthalenesulphonic acid (DNNS)/Toluene. It appears that the micellar mechanism is prevalent only in a limited DNNS concentration range which is a function of K22 concentration. (Key Words : Americium(III), Dinonylnaphthalenesulphonic acid. Extraction, Diaze-18-crown 6, Complexation) I. INTRODUCTION: Due to the large hydration energies of high oxidation state cations, their interaction with crown ethers have been found to be too weak to be measured in aqueous phase /I/. Recent studies in non aqueous media heve revealed that lanthanides as well as uranyl ion form distinctly stronger complexes with diaza crown ethers than those with analogous crown ethers /2,3/. It was thought of interest to investigate the aqueous complexation behaviour of a typical actinide ion, Am(III) with K22 and compare the results with those of 18-crown-6 (18C6). In our earlier work, picrate was used as an organophilic anion to extract the cationic macrocyclic complex of Am(III) /I/. During the present studies, it was observed that Am(III) does not extract towards the organic phase even in the presence of K22. It is apparently due to the fact that the protonated form of K22 itself extracts towards the organic phase as ion-pair with picrate anion. An alternate cation exchanger DNNS has been used in this work. Preliminary studies carried out to understand the extraction system Am(111)/K22/DNNS/toluene have been reported in the present communication. II. EXPERIMENTAL: K22 (MCB, USA) and DNNS (R.T. Vanderbilt Co., USA) were used without further purification. DNNS solutions were made in toluene (A.R.) and were pre-equi1ibrated with 4 M NaCl for 24 hours before use. All other experimental details were similar to those mentioned in reference /I/. III. RESULTS AND DISCUSSION: DNNS is known to extract metal ions by micellar mechanism. As a consequence, dependancy of metal ion extraction on DNNS concentration is independent of the charge on metal ion. It was observed during the present work that the extraction of Am(III) at pH 2.0 follows expected micelle mechanism in the DNNS concentration range 6xl0~ M to 1x10 M. Fig. 1 suggests that the extraction of Am(III) in the presence of K22 follows micellar mechanism only in a limited DNNS concentration range which is a function of macrocyclic ligand concentration. Ensor et. al, during their investigations of the system Ln(III)/15C5/DDNS/Toluene have suggested the possibility of extraction of mixed species of the type LnL Nan_3 (DDNS)n (where L = 15C5) /A/. These authors did not offer any explanation for the decrease of distribution coefficient values with increase of 15C5 concentration beyond 1x10 M. To get a better picture of the nature of extraction in the SSC - 17.1 present work, variation of D^m wae also studied as -a function of K22 concentration. Slope of -1 observed in Log D^m - Log [K22] plots^ at pH 2.0 with [DNNS] = 3 x 10 M as well as (DNNS] 6x10 confirmed the formation of 1:1 species in the aqueous phase. Possibility of the extraction of these complex species towards organic phase by the formation of ion pairs with DNNS rather than micelle formation is not ruled out. Similarly, there is also a_possibi1ity of protonated K22 itself forming ion pairs with DNNS . It is required to know the aggregation number of DNNS as well as cone. of K22 at equilibrium to discount or substantiate any of these possibiIites. IV. ACKNOWLEDGEMENTS: Authors wish to thank Dr. P.R. Natarajan, Head, Radiochemistry Division for his keen interest in this work. V. REFERENCES: 1. P.K. Mohapatra and V.K. Manchanda, Presented at Radiochem. and Radiation Chem. Symposium, Nagpur (India), AL-05 (1990). 2. J. Lagrange, J.P. Metabazoulou, P. Fux and P. Lagrange, Polyhedron, 8, 2251 (1989). 3. W. Szczepaniak, B. Juskowiak and W. Ciszewska, Inorg. Chim. Actat 147, 261 (1988). <\. D.D. Ensor, G.R. hcdonald and C.G. Pippin, Anal. Chem. 58, 1814 (1986). z z Q UJ DC UJ I U. CM (M CM CM ~ ^ ro U. CM CM CM O 5C U. U. in r- 2 2 5 O O ro w z CO z Ul w HO U 'o o Q o Z Z X x x i WI T u. ul 2 O UJ < O <, o • a: or 1 I ^s\ 1 1 ro o o ro ro i ro o 6 , i E I s8 0 2 t i u. SSC - 17.2 EXTRACTION BEHAVIOUR OF PLUTONIUM< IV) AND NEPTON1DM( IV) WITH 3- PHENYL-4-BENZ0YL -5-IS0XAZ0L0NE V. K. Hanchanda and P. K. Mohspatra, Radiochemistry Division, B. A. R. C. , Bombay-400085. SUMMARY: Extraction ana stripping behaviour of Pu( IV) and Np< IV) employing 3- phenyl -4-bt>nzoyl - "=>- isoxazolone (HPBI) has been investigated in tolune medium. HPBI was found to be a much better extractant than thenoyltrif 1uoroacetone ( HTTA) for these actnide ions. < Key words: Thenoyl tr if 1 vioroacet one, 3- Phenyl -4-benzoyl -5- isoxazolone. Extraction, Plutonium II. EXPERIMENTAL: HPBI was prepared by the benzoylation of 3- phenyl -5-isoxazolone (Aldiicli) following the method of Korte and Storika 111. Plutonium experiments were carried out with a3Tu (principal isotope) and neptunium experiments with ~3;vNp. Extraction experiments were performed by attaining the equilibrium from backward direction. Assay of the two phases was carried out by scintillation counting for -a*Pu , c^-proport ional counting for "^Np, and «<-spectrometry where mixtures of nuclid.es ( "Tu, "^Pu and "'-"Np) were present. III.RESULTS AND DISCUSSION: Delocalisation of the negative charge in the isoxazolone group is L aspons ible for th<2 exceptionally low pKa value (1.12) of HPBI/2/. As a consequence, this reagent hap been found to be an excellent excractant for several bivalent metal irjnu/3/. Fiij. 1 shi,u'i the exlract. ion behaviour of Pu< IV) with HPBI as well as HTTA. It is observed that under identical reagent concentration (0.05M), whereas quantitative extraction of Pu< IV) at 5M HNO^ is possible with HPBI, there wag negligible extraction with HTTA. Back exl i n.-f ion behc.vie.ui of Pu in Pu( IV) -HPBI -Tolune system with several stripping agents has been summarised in Table ]. Hydroquinone IV. ACKNOWLEDGEMENTS: Authors wish to thank Dr. P. R. Natarajan, Head, Radiochemistry Division for his keen interest in this work. V.REFERENCES: 1. F. Korte and K. Storika, Chem. Her.. 4. 1956 (1961). 2. A. Quiilco. G. Speionl. C. B. Lyell and R. L. McKee, "The Chemistry of Heterocyclic Cumpound6"/ Wiley Inttreeiunck, p. 1430 ( 1962). 3. A. Jyothi and G. N. Rao. Chemica Scripta, 27. 36? < 196?) 4. A. Ramaswamy (. private communi cat ion >. Table 1: Backward extraction behaviour of Pu S. No. Stripping agent %age stripping 1. 0. 1M NHaOH. HC1 ( 1M 2.5 2. 6. 1H HQ ( 1M HNU^) 55.0 3. 0.2M HxCsO.,t- 0.2M ( 1H 95.5 4. 0. ]M HCi < 1M HOI ) 99. 5 100 80 IM 2M 3M 4M 5M CHNO,]/M FtG-I % EXTRACTION OFPu(IV) WITH 0 G5M HP8I AND 0 05M HTTA FROM HNO3 MEDIA SSC - 18.2 THERMODYNAMICS OF SYNERGISTIC EXTRACTION OF HEXAVALENT PLUTONIUM WITH 1-PHENYL,3-METHYL,4-BENZOYL PYRAZ0L0NE-5 S. A. Pai Hadiochemistry Division Bhabha Atomic Research Centre Bombay-400 085/ India. SUMMARY Extraction of plutonyl ion with HPMBP in cumbination with vari&\)!5 sulphoxides viz. d i-isoamyK DIASO), di-n-hexyl (Key ' words." Thermodynamics^ Extraction, Plutonyl Ion, 1-phenyl , 3-methyl, 4 -benzoyl \>yrazolone-5< HPMBP) , Sulphox ides. > I.INTRODUCTION Aroyl pyrazclones have found their use as better chelatants in place of conventional ^-diketones like HTTA, for mere than a decade, as they are capable of exrtacting metals even frnm acidic media/1-3/. Earlier we have used these pyraiolones as chelatants for the synergistic extraction ?•? well as thermodynamics of extraction of uranyl ion, in combination with variuos neutral donors/4-7/. The present wotl describeB similar extraction studies with plutonyl ion with HPHBP/sulphoxides. II. EXPERIMENTAL HPMBP and sulphoxides were synthesized as described ear 1ier/5/. All the other reagents used were of analar grade. Pu-239 was the riajnr constituent in the plutonium used and was purified by ion exchange and double peroxide precipitation. Valency of plutonium in 1H nitric acid was adjuBted to VI with excess Ce SSC - 19.1 From the values of log KB at various temperatures, the thotmodynamic constants have beer, evaluated by the usual tamper at vire ro -efficient method and are given in Table- 1. Table- 1 Thnr modynamic Data of PuOjrt /HFMBP/Sulphox ides, DONOR Kh* log KB AF AH AS K cals/nole K caIs/mole e.u. ().?.} 5.36+0.02 -7.44+0.03 -10.64+1.0 -10.6+3.3 1)1 iVO 0.37 5.2 1+O.O3 -7.23+0.04 -6.20+0.7 +3.4+2-3 DSSO d.44 b. 39±0. CI2 -7.49+0.03 -12.30±0.3 -15.4 + 1.0 no8o 0.4^ r>. 40+0. 03 -7.50+0.04 -9.62+0.4 -7.0+1.3 ' Kh refers to relative basicities of sulphoxides/5/. III. RKSni/PS AND DISCUSSION From the data of "\ible-l it can be observed that the increase in log Ks with basicity is not observed with DHSO which may perhaps be due to the spacial steric hindrance exerted on the incoming eulphoxide molecule by the plutonyi chelate. High negative AH and low values of AS for DSS0 onwards suggest that the organic phase reaction to be of addition and not of substitution foe water molecules. This may be explained on the basis of ccatic part of ant ropy change as described by Choppin etal/B/. IV.REFERENCES 1. B. F. Hyasoedov, N. E. Kochetduva and M. K. Chmut ova, Zh. Ana 1 . Kh im. 27,678< 1978) . '^.W.E-iacher and c:. Keller, J. lnor g. Nuc J. (-hern. 35, 2945< 1973) . 3. G. A. Pr. ibi 1 ova, M. K. Chmutova & B. F. Myasoedov, Rad iokhimiya, 4, 5?. 1< 198 1) . 4.S.A.Pai and M. B. Subcaman i an, J.Rsdioanal. Nucl. Chew. 69, 42'3< 1985) . b.S. A. Fa i and M. S. Subr amaru an, J . Had i Dana 1 . NnoJ. Chem. ±Q£.> 40J( 1987). 6. S. K. Mundra, S. A. Pa i and M. B. Hubt am=in i an, J. Radioanal. Nucl. Clmm. Ar t in] efi, 116, 2O3< 1987) . 7. S. K. Mundra, S. A. Pa i and M. B. SubL ainan i an, Lanthanide and Aotintde Research.2,323( 1988). B. K. [.. Nash.G. R. Choppin, J. inorg.Nucl . Chem. 39.13 1< 197 1 >. SSC - 19.2 THE THERMODYNAMICS OF THE EXTRACTION OF PLUTONIOK '.VI) BY DI< 2-ETHYL HEXYL) SULPHOXIDE FROM NITRIC ACID MEDIUM G.R. Mahajan, D. R.Prabhu, J. P. Shukla and G.M.Nair Radiochemistry Divieion,Bhabha Atomic Research Centre, Trombay,Bombay 400085. SUMMARY The distribution ratios of Pu(VI) were obtained at different temperatures in the range of 25-50°C for varying concentrations of nitric acid and 0.2M di(2-ethyl hexyl) sulphoxide Key words: Plutonium,Sulphoxide,Extract ion, Thermodynamic functions I.INTRODUCTION Potential applications of di(2-ethyl hexyl) sulphoxide (DEHSO) ••* an extractant for U(VI) and Pu(IV) were investigated earlier 111. The effect of temperature on the the above extraction reaction* was Investigated subsequently 11,21. The present work describes th« results of our study concerning the extraction of Pu(VI) with 0.2M DEHSO and evaluates the effect of temperature on its extraction as a function of aqueous HNOa concentrations. II. EXPERIMENTAL DEHSO obtained from M/s. Fairfield, USA was used as received. n-Dodecane < AR Fluka) was washed with 1M NaOH. 1M HNO3 and finally with distilled water.Pu( VI) was obtained by oxidising with AgO and NaBrOs<0.01M) was used as holding oxidant. Equal volumes (0.5ml) of agueous phase containing HNO9 of the required concentration and tracer Pu(VI) and preequi1ibrated 0.2M DEHSO in n-dodecane were equilibrated for 1 hour in a 'hermoBtated water bath adjusted to the desired temperature ( 25—50*»C> to within +0. 1°C. The phases were then allowed to settle and aliquots from both phasns were drawn for radio assay by liquid scintillation counting.The distribution ratio ( DM) was calculated as the ratio of a31*Pu activity in the organic phase to that in an equal volume of the aqueous phase. III. RESULTS AND DISCUSSION From a plot of log DM VS log[DKHSQ) in 2M HNOa medium a slope of nearly 2 was obtained indicating that the species extracted to the organic phase to be PuO» where D**'M is the distribution ratio corrected for the aqueous nitrate complex ing of Pu IV.REFERENCES 1. M. S. Mural i,D. R. Prabhu, G. R. Mahajan.J. P. Shukla.G. M. Nair and P. R. Natarajan,Paper AL-3L, Pre-print Volume^Radiochemistry and Radiation Chemistry Symposium, Nagpur, 1990. 2. G. R. Mahajan, M.S. Mural i,D.R. Prabhu, J. P. Shukla,G.M. Nair and P. R. Natarajan,Thermochimica Acta (in press). 3. M. S. Sa jan, V. V. Ramakr ishna and S. K. Pati1,Thermochimica Acta, 47.277C1981). Table 1^ The enthalpy values for the extraction of Pu( VI) Into 0.2M DEHSO/n-dodecane from aqueous nitric acid media n M O.5 1.0 2.0 3.0 4.0 5,0 98 ± -4. 44 ± -4. 96 ± -3. 69 ± -3, 97 dt -4. 36 dt 0. 42 0.48 0.39 0.31 0.40 0.47 Mean value of AH = -4.23+ 0.45 K Cals. SSC - 20.2 SYNTHESIS AND CHARACTERIZATION OF THQRIUfi(lV) & DIOXOURANIuTl(Vl) COMPLEXES OF 4-(m-Pl£THGXY BENZYLIDENEjAMINOANTIPYRINE R.K. AGARUAL*, PRASHANT OUTT and 3AI PRAKASH Department of Chemistry, Lajpat Rai (Poat-Graduate) Collage, Sahibabad-201 005 (Ghaziabad) Keywords: Thorium/uranium/Schiff base/complexes/lR/Thermal I. SuWIARY: Thoriucn(IV) and dioxouranium(Vl) complexes of a Schiff base derived from m-methoxy benzaldshyde and 4-amino- antipyrine have been synthesized uith the composition ThX..2L (X - Cl, Br, N03 or NCS) , ThX4.3L (X - I or C1O4) , UG2X2.2L (X - Cl, Br, I, N03, NCS or CH3C00) or U02(Cl04)2 .3L. All the complexes are weekly diamagnetic. The IR studies indicate that ligand was coordinated to metal ion through azomethine nitrogen and carbonyl oxygen. The probable coordination number of Th(iv) is 6, 8 or 12 and of U(VI) is B or 10 depend- ing on the nature of anions. Thermal properties have also been discussed. II. INTRODUCTION: The Schiff base complexes of actinidas are well documented in the literature. But less attention has bean paid to biologically important Schiff base complsxes of actinides. In the present uork the Schiff base isolated from 4-aminoantipyrine has been investigated as a coordinating agtnt for thoriuro(iv) and dioxouranium(vi) salts. III. EXPERIMENTAL: Schiff base was isolated in the solid rtate by refluxing 4-aminoantipyrine and m-methoxy benzal- dehyde in 1:1 molar ratio in nethanol for £a. 3 hrs. On cooling a yellou solid mass is collected. Th(lV) and Uu2(l/l) complexes were isolated by interaction of Schiff base in required molar ratio. The solvents used either iaopropanol or acetone. IV: RESULTS AND DISCUSSION: The elemental analyses and rolacular weight determination suggest the composition of SSC - 21.1 the isolated complexes as: ThX4.2L (X « Cl, Br, NO- or NCS), ThX4.3L (X - I or ClO4) , U02X2.2|_ (X » Cl, Br, I, NO-j, NCS or CH3C00) or U02(ClG4)2.3L. All the complexes are weakly diamag- netic. The molar conductance suggest 1:2 electrolytic natur* of ThI4.3L and UG2(Cl04) 2.3L and 1:4 of Th(Cl04) 4«3L. The. infrared spectra concludes that the ligand is a neutral bidentate ligand coordinating through azoroethine nitrogen and carbonyl oxygen. In far infrared region (400-300 cm*" ) " have also been identified. The covalently bonded nitrates bshave as bidentate, thiocyanates are N-bonded and perchlorate ions are ionic in nature. The probable coordination number of Th(li/) is 6, 8 or 12 and of U(l/l) is 8 or 10 depending on the nature of anions. The TG and OT-cu.rvas of the complexes indicate the absence of uater molecule. Finally in all cases ThO2 o? ^3^8 is obtained as end product. V. ACKNOULEOGENENT: One of the authors (RKA) is thankful to U.G.C., Neu Delhi for financial support. SSC - 21 .2 COORDINATION COMPOUNDS OF URANIUM S.G. Deshpande and S.C. Jain Chemistry Division, Bhabha Atomic Research Centre Bombay 400 085 SUMMARY - Adducts of the type (UCI4.L or UC14.2L and UO2C12.L or UO2C12.2L) were obtained on reacting UC14 with 'P' containing ligands [ L=l,l,l tris (diphenyl phosphineomethyl) ethane, 1,1,1- tris(diphenyl phosphino) methane; tris (2'-cynoethyl) phosphine, triphenyl phosphine, triphenyl phosphine oxide (TPPO)] in CC14 and THF medium respectively. Complexes formed were characterised by elemental analysis, IR and determination of molar conductance and magnetic moment. Oxidation of U(IV) to ++ UO2 (VI) in THF medium appears to be due to abstraction of oxygen from THF. T. . INTRODUCTION Ho evidence is obtained for the formation of tj(iv) tertiary phosphine arHition compounds in ethanol, THF and actonitrile/1/. In view of the above,UC14 was reacted with various phosphine ligand in THF and CC1« media to study the cause of oxidation of U(IV) to dioxouranium(VI) TI. EXPERIMENTAL All the chemicals and ligands (Fullea or Ma son Colemen and Bell) were AR grade. Solvents used were dried and purified by reported method /2/. UCl^ was prepared by known method /3/. Method for preparation of the complexes is same as reported earlier /4/. IR spectra were recorded in nujol (4000-200 cm" ) and molar conductance and magnetic moments were determined as reported earlier /5/. III. RESULTS AND DISCUSSION Adducts from THF were dark yellow, diamagnetic and showed a strong absorption around 920 cm"1 which was assigned to asymetric streching frequency ( V3) of uranyl moiety /4/. Complexes obtained from CCl4 were dark_green, paramagnetic and did not show any absorption around 920 cm . No shift in the absorption band due to vi(P = 0) was observed in case of U(IV)-TPPO complex. Gas chromatograms of THF before and after the reaction indicated the participation of THF in the chemical reaction. Complexes obtained from both solvents were found to be non ionic in acetonitrile. Results obtained indicate that oxidation of U(IV) ++ to UO 2 (VI) in THF appears to be due to abstraction of oxygen from THF. Authors sincerely thank Dr J,P. Hittal and Oc A..J. Singh foe their keen interest in the work. SSC - 22.1 REFERENCES 1. Gans P & Smith B C, J Chem. Soc 4, 172 (1964) 2. Reldick I A & Bagner W B Organic Solvents physical properties and methods of purification, Interscience publications (1970) 3. Khan IAS Ahuja H S, Inorg. Synthesis Series, Interscience Publications, 21 (1982) 4. Deshpande S G, Vaidya M A & Jain S C, Proc of Indian Academy of Science 101, No.3, 211 (1989) 5. Vaidya M A, Deshapnde S G, Jain V K and Jain S C, Inorg. Chim. Acta, 143, 123 (1988) 6. MaUy D K & Ghosh H N , Indian J. Chem 28(A), 180(1989) SSC - 22.2 DISTRIBUTION COEFFICIENTS OF ZIRCONIUM, URANIUM AND PLUTONIUM ON DOWEX 1X4 RESIN IN HC1 AND HC1 * CH3OU MEDIA K.L. Ramakunar, V.A. Raman, M.K. Saxena, V.L. Sant and H.C. JaLn Fuel Chemistry Division, B.A.R.C., Bombay 400 085, India SUMMARY Distribution coefficients (D values) for Zr, (J and Pu have been determined on anion exchange DOWEX 1x4 resin in HC1 and HCH-CH3OH media at different acid concentrations. While the D values for U and Pu show the similar trend of variation in both the media, those for Zr exhibit completely opposite trend indicating the decreasing tendency in anion complex formation ability in mixed solvent media. (KEY WORDS : Zirconium, Uranium, Plutonium, Distribution coefficients, anion exchange resin. Mixed solvent mediaj I.INTRODUCTION Metal alloys of U/Pu-Zr-Al and U/Fu-Zr are being considered as potential nuclear fuais in our nuclear energy programme. Precise determination of concentration of Zr in these alloys it> necessary trom the homogeneity point of view. The first step iit this direction is the separation and purification of Zr from U/Pu and Ai. A simple anion exchange procedure tor the separation an II.EXPERIMENTAL DOWEX qlx4 (200-400 meshj resin was purified as described earlier/2/. "Zr, 233y and ^8pu have been used as tracers for determining the distribution coefficients. batch equilibration method was employed to determine the D values. III-RJEtiULlS AMP DISCUSSION The D values for Zr, U \nd Pu obtained in both aqueous UlCi) and mixed solvent media (HUi + CH^oH) of ditterent compositions art; given in Table 1. it is seen that the D values lur U and Pu arc in general higher in mixed solvent medium than in aqueous medium. The D values show the similar trend in both the media. In the cass or Zr, on the other hand, the trend in the b values in mixed solvent medium is quite opposite to that observed in aqueous medium. They are higher at lower acid concentrations. Kxperiments were repeated in other solvent compositions (9b % (JH^OH+HCi and 70 %CH-jUhH-HCl) also and similar behaviour has been observed. The results are given in Table 2. lt seems as it the anion compiexing ability ot Zr in the mixed media containing higher acid concentrations is not as high as it is in lower acid concentrations, but In pure aqueous media exactly opposite trend is seen. lt is known that chloride complexes of Zr are the weakeat/3/ and ZrCi^is formed and is stable only at higher acid SSC - 23.1 cortoentra ti^ns. At lower acidities Zr may be predominan tly t-xisLing as Zru ^. This may be the reason why in pure auid medium, the u values tor Zr increases as the acid strength increases for more and more ZrCJ. ^"complex is lu Lined. But in m xed solvent media it appears as Ll there- may be Some o lier competitive reactions at higher acici compositions resulting in the decrease in the formation ol amon complex. One such redc ioii may be the formation ot neutral,,speoies ZrCl^ . 2(Jti^UH competi } ng with the formation ot ZrClo° The neutral species may be preferentially retained in the solvent phase resulting in the decrease in the D values. However, further experiments need be carried out to understand this interesting behaviour ol Zr. 1V. The authors are grateful to Dr. D.D. Sood, Head,Fuel Chemistry Division, tor his keen interest in this work. They also thank bhri V.V. kamakrishna, Fuei Chemistry Division, tor useful discussions. V.HKFERENCES 1. K..L. Kamakumar et.nl., Uhli 6ymp. Hadiochem. Kadiat. Chem: , Nagpur, L'eb . 5-8 , iyyQ , Preprint Volume,Laper no. , HA - 4 2. K..L. Kamakuuiar et al. , Separation oci . 1'cclinoi. li, 147J(J9b J. Treatise on Anal . Chem. , t at t I1,V.J1.L> tl.M. Koitali, F.J. Elving and E.B. bandeii Kds. ( inter Science New Yurk (I'd61) TABLE J, : Distribution coei:l icients lor Zr,U and t'u in HCl and HCl+CH;jOH media ture aqueous medium Mixed solvent medium Acid cone. D value Solvent D value Zr U lu composition £i U U. 3M HCl 2 1 U.3M HCl 263 214 Ub l.UM HCl 3 2 l.UM HCl 2y .64U 146 3.UM HCl - 7 2 3.UM HOi 12 20 bb 6.UM HCl 8 1U26 6.0M HCl 14 424b J 14y 8.0M HCl 9u 17uo b.uM HCl i» 6ub5 Ib7d 0 . UM HCl 24Cu lfcsuo 1UU 1U.UM HCl It TAJjLg Z Distribution ratios for Zr in mixed solvent media D liol Vent D composition value composition value yb/iCHjUll 0 . 3M HOI tibV U. 311 HCi 34 lb l.UM HU1 1 .UM HCi L2JU 3.0M HCi 3.uM HCi 9 b y 5.UM HCi 5.UM HCi •ihif H.^M 11C1 •fb U.uM HOI 2C1 1U.UM HCi bb 1U.UM HCi 1U4 SSC - 23.2 ON THE CONVERSION OF PLUTONIUM NITRATE TO PLUTONIUM OXI0E VIA Pu (IV) OXALATE PRECIPITATION ROUTE G.R.Dharmpurikar, K.Kumaraguru, K.M.Michael .Y.D.Shukla , G.M.Dhabolkar, N.Ramamoorthy ,C.V.Narayanan and S.C.Kapoor. Fuel Reprocessing Division,BARC.Trombay.Bombay- 400 085. SUMMARY A continuous precipitation cum washing setup along with SS filtration setup was designed, developed and tested for suitability in large scale conversion of Pu(NO3)4 to PuO2 via Pu(IV) oxalate route . (Key Words : Plutonium nitrate, Pu(IV)oxalate, PuOg, Continuous precipitation). I. INTRODUCTION Pu oxalate precipitation process has been widely employed in the conversion of Pu nitrate solution to Plutonium oxide because of its amenability to continuous operation and relatively good decontamination with respect to impurities. A precipitation cum washing column was tested for continuous precipitation of Pu solution as Pu(IV) oxalate at elevated temperatures.A SS filtration setup having p'orous sintered SS disc of 15 micron average pore opening was fabricated and successfully used for direct calcination of Pu oxalate.All the aspects regarding the optimum capacity of the column,impurity pickup,fr&quency of regeneration of the filter medium etc have been studied. II. EXPERIMENTAL A gla&s column of around 1000 mm long ,90 mm od and 80 mm id was used for the column (Fig 1).Oxalic acid and hot>Pu nitrate solutions were continuously fed to the column at desired rates through metering pump.A hot waterbath was used for heating the feed.The oxalate slurry was continuously washed in the column itself.The washed slurry was pumped to the SS filter boat having a connection to the vacuum system at the bottom. Various flow rates and temperature ranges were studied to arrive at optimum condition for routine applications.Retention ability,feasibility of direct ignition, impurity pick up and maximum amount of Pu that could be filtered and ignited without regeneration were studied by using the frit. During the precipitation, both by cold and hot conditions, the particle sizes of the oxalates and oxides were determined. III. RESULTS AND DISCUSSION Details regarding precipitation of Pu as Pu(IV) oxalate are already reported /I/.Its adaptation into a continuous process has also been reported /2/.Tests showed that the column could be operated at a maximum feed rate of 16 1/hr upto an elevated temperature of 60°C.Filtration was slow in cold condition of precipitation (23°C) and the precipitate was found to be finely SSC - 24.1 dispersed. The frit could filter large quantities of oxetlate precipitate in hot condition without regeneration /3/. The retention of Pu in the frit when it was used upto the saturation level was only around 0.15% of the total amount filtered, similarly in the next campaign large amount of hot Pu oxalate precipitate could be filtered and ignited when the same frit was used for several runs.No impurity pick \x? from the frit was noticed.Thus better throughput rate could be achieved by thia system thereby reducing processing period and radiation exposures to the operating personnel. IV. ACKNOWLEDGEMENTS. Authors wish to thank Shri.A.N.Prasad,Director,Fuel Reprocessing and Nuclear Waste Management Group, Shri.M.K.Rao, Head, Fuel Reprocessing Division and Dr.R.K.Dhumwad, Head Laboratory Section for their keen interest in the work. V . REFERENCES 1. J.F.Facch Jr.and K.M.Harmon, Precipitation of Pu(IV) as oxalate. Plant Process Unit.Separation Technology Sub Section, Hanford Atomic Products Operation, Richland, Washington. H.W. 31186,(1954). 2. Pierre Richard Continuous Process of Precipitation of Plutonium Salts ORNL-TR-1720 , (1967). 3. H.A.Dayem et.'al, Coordinated Safeguards for Materials Management in a Nitrate to Oxide Conversion Facility . LA-7011, (1978). THERMOCOUPLE DXALIC " ACID HOT WATER PIG7/- BATH OXALATE *-- SUPERNATENT" WASH ACID [Pj ^SLURRY DOT I r»"+ TD VACUUM Flg-1. CDNT1NUUUS PRECIPITATION, WASHING CUM FILTRATION SYSTEM SSC - 24.2 PREPARATION OF Puff V. VENL'GOPAL, K.N. HOY, V.S. IYER, Z. SINGH, N.K. SHUKLA, R. PRASAD and D.D. SUOD Fuel Chemistry Division, BARC, Tromboy, Bombay 400 085 SUMMARY : This paper describes the preparation ofPuN by reacting nitrogen with plutonium metal chips in the temperature range 750 to 850 K. The reaction was carried out in a closed steel vessel under static nitrogen pressure. The course of nitriding was followed by measuring the drop in nitrogen pressure. Complete conversion to PuN could not be achieved. The rate of reaction was a strong function of the surface characteristics of the metal chips. 1. INTRODUCTION Pure PuN is required for determination of its thermodynamic properties. Normally PuN is prepared by arc-melting of metal under nitrogen atmosphere, carbothermic reduction under flowing nitrogen and by hydriding, dehydriding and ~»itriding of metal (1J. It was decided to try preparation of PuN by reacting nitrogen with the metal chips under pressure. Nitrogen reacts with plutonium metal according to the reaction. Pu(g) + 1/2 N2(g) - PuN(g) As it is a gas solid reaction, the surface area of the metal is an important factor governing the reaction rate. 2. BXPfiBIMBsTTAL A stainless steel vessel was attached to the bottom of an inert atmosphere box using water cooled flange. The top flange to the vessel had provision for introducing a thermowell, a pressure gauge, gas inlet and outlet. All the connections were mode using swags-1ok type fittings. Metal charge was loaded into a tungsten crucible kept in a graphite crucible and lowered into the vessel. The entire assembly was flushed with high purity nitrogen and pressurised to 4 x 10 Pa to test leaktightness. A calibration experiment was performed by pressurising the vessel to 2 x 10 Pa at 298 K and then heating the vessel to 850 K at a heating rat* of 20 K/min. The variation of pressure with temperature and time was monitored. When the vessel was cooled to room temperature, nitrogen pressure was again noted. The pressure returned to 2 x 10 Pa confirming the leaktighntess of the vessel. After charging with plutonium and pressurising by nitrogen to 2 x 10 Pa at 298 K, the vessel was heated at the name rate (20 K/min) till it reached the set isothermal temperature. It was observed that around 20% of reaction was complete before reaching the set temperature. The reaction rate wae then followed with respect to time. SSC - 25.1 3. RESULTS The fraction reacted ' (a) 'was' ' calculated at a given isothermal temperature by monitoring the pressure drop with time. The reaction was carried out -at 745, 748 and 779 K. A typical a versus time plot at 779 K is shown in figure 1. The data are found to fit into a parabolic rate equation. The rate of reaction is linear upto 40% of reaction and then levels off implying that no significant reaction could further be carried out. The initial reaction with a linear rate may be due to the chemisorption of nitrogen on the surface of the metal forming a product layer of PuN. The rate of the reaction was thus significant till the product layer is formed. Once the product layer attains sufficient thickness, then it retards the diffusion of nitrogen to the inner core. In the present study it has been found that the reaction could be carried out upto 60% of completion. The ruaction rate has been found to be very much dependsnt on the nature of metal chips. Fire chips of metal were found to be react with nitrogen. Coarse chips do not react at al I. The product retained its geometry when taken out from the vessel. The product was found to be very reactive. When it was taken out for characterising by X-ray diffraction, it reacted with oxygen impurity in the box immediately. Hence the product could not be characterised. CONCLUSION The present study indicates that plutonium metal chips cannot be used for the preparation of PuN. The reaction does not go to completion even with frne metal chips. The product is reactive and hence it can not be used tor high temperature thermodynamics research. REFERENCES 1. H. Matzke, Advanced l.MFBR, Fuels, Elsievier, Amsterdam, (19H6) pp.335. SSC - 25.2 FRACTION REACTED,Ot SSC - 25.3 EPR INVESTIGATION OF "»!ls* (Sf1) H. D. Sastry, N. fC. Porwal and Mithl<=sf R"ad i ochem i s t r y Division, Bhabha Atomic Research Centre, Ttomhay, Bombay-400 085, INDIA. (SUMMARY) We present the first observation of thn hyperf ine ptrueture : due to * '"U (1 = 5/2) in its pentavalpnt form stabilized in LiNbOa. The expected systematics of the hyperf sue constant ^^"A, in octahedral U0A^~, ( U0=F)*•- and ( U04F2)'" shows that with substitution of oxygen by less covalent 1 igand F~, the Z"A decreases contrary to the general expectation. This is understood intfirniR of crystal field admixtures of upper'/ states into the ground state on increased distortion of the octahedral complex. w (Key words : Hyper fine constant "=»U. EPR, LiNbOa:U ~) I.TNTR0DUC10N Among all the oxidation states of uranium, the pentavalent ion, U0"", with outer 5f * electronic configuration is interesting for EPR investigation. Under n«arly octahedral oxygen coordination its RPR is expected sf easily s^r^essiiile temperature. Lewis et.al./i/ have shown tha' V.^* gets stabilized in nearly w octahedral Nb * sites in I.iNbOa, i.ui doping with 0a0o, and its EPR is observable even at room tempertur.". In the present work we have ZMS carried out similar investigations with Ua0a (nearly of 100% radiochemical purity) doped LiNbO3 powder. These investigations are primarily directed at getting the i nf i. Tabis: Systematics of hyperfine constant of jn oxyfluro complexes. Remarks Complex ( in 10—* cm-*) < in -" cm"1) 108 129 exper imente)y observed*present wock) 99. 5 122. 2 calculated form data of ref./2/. Trans. 77.9 127.65 ! It can be seen that fluoride ion substitution decrease A substantially. This looks rather surprising in view of the greater ionic nature of F~ compared to 0—. This effect, however can be explained by the admixture of higher lying 7 states, due to lowering of symmetry in to the ground state which increases the electron density in axial (0-U-F) direction. REFERENCES 1. W.B.Lewis, H. G. Hecht and M. P. Eatsman, Inorg. Chem. , 12_, 1634 < 1973). 2. P. B. Dora in. C A. Hutch iuson and E.Wong. Phys. Rev.. 225.' 1307 (1957). 3.I.Orsu and V.Lupei, Bui 1.Mag.Res.,b, 162 <1984). SSC - 26.2 OPTIMUM CONDITIONS FOR THR QUANTITATIVE CONVERSION OF PuO2 TO PuF,4 BY HYDKOh'I.UORI NATION K.N. ROY, V. VF.NUGOPAI., Z.SINGH, V.S. 1YF.R, N.K. SHUKLA, V.N.VAIDYA, R. PfiASAU AND D.I). SOOD Fuel CheiniHt ry Division, B.A.R.C., Tromliny , Bombay — 400 Ofl.1) (Key words : PuO^, PuOF2l PuFz, SUMMARY : This paper describes the optimum conditions like temperature, bed depth, HF ("low [file necessary for t hie quantitative conversion of PuO-2 to PuF/,. In the present. process an overall HF utilisation of '•.">% has been achieved which was three times larger than the utilisation reported in the literature. An intermediate product PuOF^ is proposed. I INTRODUCTION Plutonium halides and PuO2 are used as starting material for the preparation of metallic: pint on i urn. Until recently PuF^ was almost universally used as starting material for the preparation of metal , but. of late direct, oxide reduction (DOR) process has also gained importance. Plutonium tetraf1uoride preparation has been studied by a number of workers . It can be prepared by She hydro 11 uori nat i on of PuO-j or peroxide or oxalate in the presence of oxygen. The presence of oxygen is necessary to avoid the reduction of PuF^ to PuF'} due to hydrogen formed by the reaction of HF on the metallic walls of the container, ths gas piping and the reaction vessel. The liyrlrof luorination is carried out in a static bed. As the hydrofI norinaI ion is a solid-gas reaction, large surface area and thin layer of oxide are desirable for the quan1itative conversion. The process parameters are controlled so that HF utilisation is maximum and HF disposal is less cimberscme. II. EXPERIMENTAL The essential parts of the experimental set up are: (i) argon purification assembly, (ii) HF gas manifold, (iii) reaction vessel, (iv) off-gas disposal. Argon carrier jfas was purified by passing through towers containing molecular sieves and heated uranium metal (770 K). HF manifold consists of nickel cylinder for storing HF and differential manometer for monitoring the flow rate of HF. The reaction vessel is fabricated from nickel. It has a flanged opening at one end and the other end was attached to a nickel filter to avoid the carry over of powder. Nickel boats were lined with platinum to minimise contamination of tbe product. Metal diaphragm tnonel valves and flare fittings were used wherever necessary. The entire gas line carrying HF was heated to 400 K. The unreacted HF was continuously monitored by reacting with standard NuOH solution. in a well ventilated fume hood. After loading the oxide into reaction vessel, and. closing the flange, the entire system was pressurised to 2 x 10~ Pa in argon to check leak tightness. The vessel was heated and the flow of HF + O2 started. Completion of reaction was inferred when flow S S C - 2 7.1 rate of HF in the effluent was same as that in the inlet. III. RESULTS AND DISCUSSION The fol1 owing parameters were optimised during the present investigation: (i) temperature of hydrof1uorination, (ii) flow rate of HF, (iii) oxide bed depth. The oxide powder (7-8 m2/&) was observed to react rapidly with HF at 423 K and 50% reaction was over at this temperature. Reaction rate was then observed to pick up only when temperature was increased to 573 K. However, complete hydrofIuorination was not found possible at this temperature and about 10% of this reaction could be completed only by raising the temperature to 800 K. HF flow rate in the range of 75-125 mi/'min was tried. At 75 ml/min the HF utilisation was better but hydrof1uorination time was large at 125 ml/min substantial quantities of HF were released to the effluent. The flow rate of 100 ml/min was found to be optimum. The depth of PuC>2 bed was very from 10 to 20 mm. When bulk of the hydrof1uorination reaction was carried out at temperature below 573 K, oxide bed depth of 20 mm could be hydrof1uorinated completely in 6-7 hrs. Completion of 50% reaction at 423 K indicated that the reaction could be proceeding as follows: Pu02 + 2 HF > Pu0F2 +H20 PuOF2 + 2 HF > PuF4 + H20 The product after 50% hydrofluorination was pale green as compared to khaki green oxide. X-ray diffraction did not reveal any lines of PuFz, and only weak lines of PuO2 were indicated. This indicates that an amorphous PuOF2 is formed. Based on the results, a stepwioe hydrofluorination method as per the temperature scheme as given above is suggested. IV. CONCLUSION The present study has established a procedure for hydrofluorination of PuO2 at lower temperatures. This procedure leads to higher HF utilisation, less waste disposal problems and gives a non caked product. Much larger bed depths of oxide can be hydrof1uorinated because of proper control of temperature and this leads to smaller and simpler equipment. REFERENCE 1. M. Taube (Ed.), Plutonium, a general survey, Verlag Chemie GmbH, (1974). 2. A.N. Volskii and Ya. M. Sterlin, The metallurgy of plutonium, Israel Program for Scientific Translations, Jerusalem (1970). SSC - 27.2 5PECTR0PH0T0METRIC DETERMINATION OF OXYGEN! TO URANIUM RATIO IH URANIUM DIOHIDE BASED OM DISSOLUTION IN SULPHURIC ACID B.Narasiaiha Murty, R.B. Yadav, C.K.Ranainurty and S. Syamsundar Cpntrol Laboratory.Nuc1 ear Fuel Complex.Hyderabad, India t.SUMMARY: The present method describes a spectrophotonetrlc determination of 0/(J ratios in urar.iutt dioxide fuel pellets and uranium oxide powders. Th I I. INTRODUCTION: Uranium dioxide is one of the major nuclear fuel materials being used in nuclear power reactors. The 0/U ratio of Uranium dioxide is an important parameter in the characterisation of the sinterabi1ity of the powder and it also affects the performance of the fuel under reactor conditions.Hence, the determination of 0/U ratio is essential as a quality control measure. The spectrophotometric methods for the determination of 0/U ratio of UOi . . reportod so far fi-32 are based on dissolution in phosphoric acid medium which suffers froa many disadvantages like high viscosity, longer dissolution time. poor sensitivity and elaborate treatment procedure [4 5 which makes the spectrophotometric determimation procedure highly inconvenient for routine use. The present method is based on the dissolution of the sample In 1M sulphuric acid containing a few drops of HF and subsequent measurement of absorbances at specific wavelengths for U(IV) and U(V1J from which the 0/U ratio was calculated. .MJLt JOJLERIHENTAL: Reagents used < sulphuric acid and HF ) were of analytical grade. Absortoance neasurements were carried out on r Shimauzu Model. UV-24S UV-visibie spectrophotometer. J-V-.PROCEDURE : The UO. pellet is taken intc a cylindrical container of the Spfcx Mixer Mill and the aid in that is displaced with a jet of argon and the- lid is tightened. Grinding is dc.iv, m.ly for 5-10 sees so as to get powder of the required form and to avoid air oxidation. The powder uo obtained is used tor dissolution in S0-60 ml of 1M H,SO, containing a few drops of HF uiu'er argon atmosphere on a water bath. The dissolution is carried iut for about. 1.5 hours. In the case of 00^ powders however, t ie dissolution is carried out for about half an hour only without purging argon du-ing dissolution. These solutions after cooling to room temperature are used fur abr.orbance measurement s. V. RESULTS AND DISCUSSION: The molar extinction coof t lciiintc of U(1V) at 630 and b3b urn and that of U VI.CALCULATIONS; The 0/U ratio can be calculated HOT pel lets/powders using the following general equation. 0/U = 2 • (A/B)x Ja/b) where A.B.a and b are the solar absorptivity,.observed absorbanoa values of U(!V> and U TABLE 1 Coaparision of 0/U ratios in U0»•. and Ua0.. 1 S.No. ! Saapla 0/U ratio value for UQ,hU30«» ! Code No. S spec t r ophotoae try voluaetry 1. ' M 370 ! 2.077 2.078 2. M 365 i 2.058 2.056 3. M 30 2. 120 2.118 4. n 31 2. TABLE 2 Coaparision of 0/U ratio of UO? fuel pellets. ! Oxygen to Uraniua ratio for UO; pellets i i i S.No. 13M phoshoric acid aediua ! 1H BU!phuilc acid aediua ! < 544 na, 290 na for U(IV >'f 535 nm . 285 nm for UIIV) i and U(VI) respectively.) '. and U(IV) respectively.) I 1. 2.010 2.009 ' : /,• 2.007 2.008 J 3. 2.006 2.005 ! A. 2.008 2.006 1 5. 2.008 2.007 t VII.REFERENCESt l.Takeishl. H.; «uto» H.; Aoy»«l» «-t Adaohl, T.t Izawa, K.t Yoahida. Z.j Kawaaura, H.3 Kihara, S. Anal. Chaa. 58.458.(1986). 2.Kuhn, E. } Bauagart >1, G.; Schnieder, H. Fressnlus 'Z. Anal. Chaa. 264.1H93. (19731. 3.Ahaed. M.K.; Sreenivasan, N.L. Anal. Chea. 58.2479,<1986). A.Tolk, A.; Lingerak, W.A. Proceedings of panel on Analytical Cheaistry of Nuclear Fuel; SAEA: Vienna,(1972)5 5T1/PUB/337 pp 51-58. SSC - 28.2 CO CO n oo 250 aoc 700 WAVELENGTH Figure 1. Absorption spectra of U ( IV ) and U ( VI ) in 1M sulphuric acid medium. (a) U (IV), H.6 mg/m L (b) U (VI) 0.32 mg/ml CRYSTAL STRUCTURE OF K4Pu(SO4}4.2H2O K.D.Singh Mudher and S.C.Jayadevan Fuel Chemistry Division, Bhabha Atoinic Research Centre, Bombay 400 0«5 India. SUMMARY - The structure of K4Pu(SO4)4.2H2O was determined by single crystal methods. The crystals are monoclinic, space group P2^/c with four formula units in a cell of dimension a=12.392 A, c b = 11.119 V c ~ 13.546 \\,, & = 111.8111.8°,, ppQbQbgg =3 3 . . lOm lOmgg . ir.T ir T, ,P caPcaii 3.12mg.m . The structure consists o_f cent-"ot;miiieciic dimeric 8 anioris [ (SO4 I ,Pu (SO^l 2Pu ( SO4) 3) " where each Pu ion is coordinated to 5 O atoms from three bidentate sulphate and two tridentate sulphate groups. Neither of the water molecules are coordinated to Pu ions. < Key Words: Crystal structure, K^Pu(SO^»^.2H2O, monoclinic, dimeric ions, coordination geometry ) I. INTRODUCTION Plutonium disulphate tetrahydrate, Pu(SO4)2-4H2O is a recognised chemical standard of plutonium(1). In spite of the a- activity and the four molecules of water of hydration, it remains stable on storage. Four oxygen atoms from four water molecules and four oxygen atoms from four sulphate groups form a square antiprism around the plutonium ions (2) . The search for an alternate standard without any water of hydration has led to the identification of K4Pu(SO^)4(3). While the rigorous evaluations on the stability, stoichiometry and purity is being conducted in our laboratory, it was found necessary to undertake a crystal structure determination on the parent compound, K4Pu(SO4)*.2^0,from which it is obtained by dehydration. The results of this study are reported in this paper, II. EXPERIMENTAL Crystals of K4Pu(S04)4.2H,0 were obtained by the slow evaporation of a solution containing calculated amounts of K^SO4 and Pu(SO«)^.$H ,0 in 1M H-^SOj . Selected crystals were enclosed in quartz tubes (ff.3nun ), the open end of which was sealed with araltdite and mounted on the goniometer. Intensity data were collected on films by the equi-inclin*tion nultipie-flira Waissenberg method using Cu Ka radiation on e crystal of dimension 0.6x0.25x0.25 mm. The intensities of 1300 reflections were estimated visually using a standard scale and were corrected for Lorentz and polarisation and absorption effects. The crystal data derived from the photographs and refined by least squares method are: a=L2.392 A, b-ll.l_19 A, c=13.546 A, P =111.8°, lOmg.min , pcai = 3.12 mg.mm" , Z=4, Space group= P2A/c. III. RESULTS AND DISCUSSION The structure was solved by the heavy atom method and refined using full matrix Least squares program (XFLS) for atomic positions, anisotropie temperature factor for Pu, K and S, isotropic temperature factor for all 0 atoms to a R-factor of SSC - 29.1 0.12. Further refinement work is in progress. The structure consists of potassium ions and centrosymmetric SO anions [ ( SO4 ) 3P11 ( SO^ ) 2Pu * 4 ' 3 J similar to those found in the structure of K^Ce(SO.)4.2H2O(4). The structure of the dimeric ions is shown in the Figure. Each Pu atom is coordinated to nina oxygen acorns at an average distance of 2.37 A. Six O atoms from three terminal bidentate sulphate ions and three O atoms from two bridging sulphate ions (each tridentate) are coordinated to the Pu ion3 . The coordination polyhedra is neither a tricapped prism nor a monocapped square antiprism. Since none of the water molecules are coordinated to Pu ion, anhydrous K^Pu(Su^)^ compound could be easily prepared without affecting the coordination geometry of the Pu ions. The structure evaluation thus is in support of the selection of anhydrous Ky,PC4Pu(SOu (SO,,4) )4 A as a chemical assay standard for plutonium. REFERENCES 1. C.E. Piecri and A.W. rel, Report, NBL-204 (1963). 2. N.C. Jayadevan, K.D.. Mudher, Zeit. Kristal.161,7 (1982). 3. K.D. Singh Mudher, K.. aandekar, K. Krishnan, N.L. Misra, N.C. Jayadevan and D.D. ood, BARC Report, 1496 (1990). 4. N.C. Jayadevan, K.D. Singh Mudher, National Seminar on Crystallography, Bombay (1989). Figure - Structure of dimeric anions in K^Pu ( SO^ ) SSC - 29.2 OXIDATION BEHAVIOUR OF UC AND (Uo,9Ceo.i)C WICKOSPHERKS 8.K.NUKKRJEE, G.A.RAMA RAO, J.V.DBHADRAYA, V.N.VAIDYA and D.D.SOOD Chemistry Division, B.A.R.C. Bonbay-^00083. SUMMARY: Oxidation studies on UC and (Ug.9 Ceg j)C microEpheres were carried out under isothermal and non-isothermal heating conditions using a thermobflIance. The partial pressure of oxygen in the reaction environment was maintained at lO.lkPn. The effect of heating rate on the mechanism and kinetic parameters of the reaction was studied. (Key words: Uranium carbide, oxidation, kinetics, microspheres) I. INTRODUCTION Several reports are available on the kinetics of oxidation of UC "d (U,Pu)C powders and pellets of varying densities under different • '"rtial pressures of oxygen/1-4/. There is p.o general agreement in 'e results with respect to mechanism and kinetic parameters. This -»id be due to the strong dependence of the reaction on the '•nditions of heating, nature of sample, sample size and partial pressure of oxygen in the reacting environment. In the present nork :;e effect of heating rate on the mechanism and kinetic parameters of he reaction was studied. During this study dense UC/(UQ gCeg j)C icrospheres prepared by internal gelation process were used. The •!ata from non-isothermal experiments were analysed, by calculations -uggested by ?,sako/5/, using a computer program suggested by Ravi ndran/6/. U. EXPERIMENTAL Samples oi 5)C e'td (UQ # tjCerj. i )C in the form of mi crospheres were prepared by heating UO3 and UO3+CeO2 gel containing 3.45 to 3.5 mole of carbon per mole of mr,tal, in vacuum/7/. The miorospheres were 97%TD, containing 4.7-4.8%C and 6.1%O2. The thermogravimetric experiments were tarried out on SINKU-RIKO thermal analyser. Amount of sample taken for each experiment was 12-13mg. The partial pressure of oxygen '. n the reaction environment was maintained at lOkPi. Isothermal experiments were carried out between 573-613 K. During non isothermal experiments heating rates were varied between 1-6 K/in i n. The reaction products were analysed by chemical methods and XRD. III. RBSU1.TS AND DISCUSSION The weight changes recorded from the thermogram and the XRD data of the intermediates and the final products indicate that the oxidation of carbide proceeds to t.5ie formation of U3OQ through the intermediate dioxide. The overall reactions for UC and (UQ_gCeQ^j)C oxidation are represented by the equations 1 and 2 respectively. 3UC + 7O2 > U308 + 3CO2 (1) 3 u Ce c ( 0.9 0.i> + 6.902 > 0.9U3O8 + 0.3CeO2 + 3C02 (2) Results from non-isothermal experiments are given in table 1. During non isothermal study, as can be seen from table 1, reaction rate controlling process, activation energy(Ea) and preexponentia 1 factors were found to vary with rate of heating. At lower heating rate(lK/min) reaction is governed by diffusion controlled process, g(u) = -ln(l-a) _i(3) where a is the fraction reacted, with Ea of 125 kJmoie . At intermediate heating rates(3-5 K/min) Ea reduced to around 80 SSC - 30.1 k.hnole-1 with n rate equation of phase boundary controlled as given i n eqiiri t i on 4 . l/3 *(") = ( 1- (1-u) ) At higher heating r«tf'6 K/min) the rate controlling process is imclent ion find its growth fitting in eq.5, with a reduced activation energy of '4 5 kJmole g(a)= -1 ri( 1-a ) (5) The change in rate controlling process with the change in heating rate could be explained on the basis of following points. At lower heating rates slow evolution of product gas resulted in the iormation of a protect ive product layer- on the reactant . Diffusion of CO2 through this product layer is slow and rate controlling. In the i.'iiscs of intermediate heating rates the evolution of product gas is compnrat i ve I y faster and t ha product layer' formed on the surface peels off once <« critical thickness is reached thereby fitting in to he equation of shrinking core mode 1 (eq.4 ) . In the case of higher tea' i :ig rates, based on the rate equation *>, the reaction appears to >e (!<:i:u i" i ng through out I he bulk of the sample. This is further supported by delayed formation of "3OQ as indicated by XHI) studies of he sample? taken at various stages of the reaction. Comparison of he results at S.No. 3 and 6 of Table 1 points towards the delayed formal ion of U3OQ in case of (UQ,qCeg 1)C oxidation which is governed by IV. KEFKRKNCKS 1. H.M.Del!, V.J. Wheeler and E..). Me Ever, Trans, Faraday Soc, 62, 3591 ( 1966) 2. K . A. Peakal 1 and .1 . E . Acit i 1.1 , J " )ss Common Metels, 4,426 (1962). 3. C.Moreau and J . Phi 1 ppoi't , Con •• Rend., 258, 4079 (1964). 4. K.Naito, N . Kainegashi ra , 'F.Koni, •,»nd S.Takeda, J. Niicl. Sci. and Techno I., 13(5), 260 (1976). 5. J.Zsako, J.Phys.Chem., 72, 2406 (1968). 6. P. V . Havindran, Thermooh imi ca Ada, 39, 135 (1980). 7. S.K.Mukerjee, J.V.Dehadraya, Y.H.flamfinkar, V.N.Vaidya and D.D.Sood, Convention of Chemists, Calcutta (1988). TABLE 1 Oxidation of UC and (U() 9<-'e() j)C microspheres under non-isothermal heating conditions. Partial pressure of oxygen in flowing argon+oxygen 10. I kPa. Sample size = 12-13mg. No Sample Heat i ng start End Rate Ea s. 1 Z 1 Hate Temp Temp contro1. kJmo1e 1 Sec ' k/misi k process 1 UC 1 483 628 Eq.3 125 1x10® 2 UC 3 488 658 Eq.4 75 1x10^ 3 UC *'; 495 667 Eq.4 80 4xio;r 4 IJC 5 498 688 Eq.4 85 6xl0j 5 tie 6 513 735 fiq.5 4 5 1x10 6 • (IJO.9CeO. 1)C 4 473 685 Eq.6 96 4x10 SSC - 30.2 X-RAY, TftKdftAt. AttO ItitHAKKV SiVBIRS OK RURJ&tun AND CAKSIUM URANYI. OXALATR HYDRATES K.L. CHAWI.A, N.D. DAHALE AND N.C. JAYADFVAN Fuel Chemistry Division, B.A.R.C, Trombay, Bombay 400 085 C SUMMARY: M2UO2( 2°4)2•xH2O compounds (M=Rb and Cs) have been prepared and characterised by chemical, thermal, »-ray and infrared studies. The compounds belong to orthorhombic system. Infrared and thermal studies show that the compounds decompose to rannourflnates through the formation of UO2 and Alkali metal cflrbonate. (Key words: Alkali metal uranyl oxalates, X-ray diffraction, thermal analysis) I. INTRODUCTION Al though the preparation of all the alkali metal urarnyl oxalates with UO2 to C^O^ ratio of.1:2 are known, structural studies on U only K2UO2(C20^)2.3H2O ' ' are reported. It is shown that the uranyl group has a planar pentagon of oxygen atorns from^ two bidentate oxalate groups and one water molecule. Earlier^""' we have reported the thermal decomposition studies of lithium and sodium Baits. We report here the results of X-ray, thermal and IR studies on the rubidium and caesium selts. II. EXPERIMENTAL Saturated solutions of rubidium or caesium oxalate and uranyl oxalate were (nixed in equimolar proportions. The solutions were heated to boiling and allowed to cool slowly. The yellow crystals separated wer<- washed with ice cold water and alcohol and dried in air. Results of the chemical analyeis for uranium and oxalate and the thermal analysis for water of hydration confirmed the composition as Rb2UO2 (C2O4 ) 2 • 2H2O and Cs2UO2(C2O4 ) 2 • 2H2O . The X-ray diffraction patterns were recorded on a Sienen's diffractometer using CuKa radiation. The infrared absorption spectra *er* recorded on a Perkin-Elmer Model 180 spectrophotoineter with the samples dispersed as KBr discs. Simultaneous TO and DTA were recorded on a ULVAC Thermal Analyser. III. RESULTS AND DISCUSSION The TG and DTA curves obtained show that rubidium and caesium compounds loose both the weter molecules in a single atep between 100-180°(J which is characterised by endothermic DT'A peaks ot 140°C for the rubidium compound and at 120°C with a shoulder at 130°C for thtj caesium compound. It is observed that all the alkali metal uranyl oxalates follow the saine thermal decomposition scheme as shown below: 2 > U02 + M2CO3 — > M2UO4 (M - alkali metal) SSC - 31.1 The X-ray diffraction patterns of Rb2UO2(C2Oz,) 2 • 2H2O and 032002(0204)2-21120 could be indexed on or thorhombi c cells of dimensions a = 10.96 A°, b = 17.18 A° . c =» 7.67 A° and a = 9.31AC, b = 16.00 A°, c = 12.11 A° respectively, containing 4 molecular units in both cases. The thermal decomposition scheme discussed above is further confirmed by the IR spectra of these compounds heated to about 350°C in argon atmosphere. Spectral bands appear in the range 1100-1055 cm due to carbonate and uranyi bands in the range 950-920 cm disappear as expected. The absorption bands in the infrared spectra of these oxalate complexes are listed in Table J. All have absorption bands from the water molecule and the coordinated oxalate groups. However, it is difficult to arrive at the structure by interpreting these absorption bands. The infrared spectrum of K2UO2(C2O^)2.3H2O has been discussed in terms of simple pentacoordinated arrangement involving two hidentate oxalate groups and a water molecule. The similarity of the IR bands with those of the potassium salt listed in the table indicate similar structure. IV. REFERENCES 1. N.C. Jayadevan and K.D. Singh Mudher (unpublished work). 2. R.N. Shchelokov and V.E. Karasev, Ruse. J. Inorg. Chem., 19, 776 (1974). 3. K.I.. Chawla/ N.D. Dahale and N.C. Jayadevan, Proc. 7th National Symp. on thermal analysis, Srinagar, (la89) p. 171. TABLE 1 Infrared spectra of alkali metal uranyi oxalate hydrates (Cm ) * Assi gnment K Rb CB V OH 3630, 33O. 3410, 3630 va8 CO 1727, 1718, 1655 1715, 1685, 1655 1715, 1655 1/sy CO J455, 1420, 1288, 1450, 1405, 1380, 1435, 1405,1380 1270 1295, 1250 1280, 1250 927 925 920 VaB O-U-0 Vsy O-U-0 840 835 830 6 o-c-o 787,740 785 . 785 • Ref (2) SSC - 31 .2 OXIDATION STATE OF URANIUM IN V^Oj FHOM .SOi.IT) STATE REACTIONS N. r"._ Jayadeyan ar.d M. Keskar, Fuel Chemistry Division Bhanhn Atomic Heiiearch Centre Trombny, B»nihfly-1(H)()2fl. SUMMARY: U3O7, an intermediate oxjde formed in the oxidation of UO2 to "3O8, when reacted with (NH/j^SO^ at 2f»O°C is found to form a mivtwr*? of ammonium uranium double sulphates. Chemical,X- ray and thermal analysis combined with UV and IR spent roscopi c measurmeiit s flire interpreted to show tfiat U3O7 is a mixed valence oxide containing U{"V) and ui(VI) in the rfttio of 2:1. (Key words: Uranium oxides, double sulphates, oxidation states, solid state reac!ions.) I. INTRODUCTION: The low temperature oxidation of UO2 gives an oxide of approximate compos 11ion of UO^ ^3 which are col lectivly referred as U3O7. At leas? five different phases have been identified, all of which have lieen characterised as tetragonal pseudocubi c structures derived from the t iuorite structure. However, wide discrepancy exists in their characterisation, particularly in the O/IJ values and the c/a rat io. PXAFS and XPS spec t roscopy have ruceni ly been used, but have tailed to give the oxidaion state of uranium in U^O-j /1 ,2/. The low temperature reaction of UOj , UO3 and U3O3 with ammonium sulphates have recently been interpreted to derive the oxidation states of uranium in U-jOg/3/. In this paper, we report the results of the solid state reaction o.f U3O7 with (NH/^^SO^ and the analysis of the react ion products. II. EXPERIMENTAL: •JO;) prepared through ADU route was equil iibrated at 700 C in argon- HXhydrogen gas atmosphere to obtain stoichiometric UO2 which was converted to U3O7 by heating in air at 250+5°C for 3 hrs. The 0/U was determined by thermogravimetric equ i i I i bra t i on method. It was mixed with excess of (NH^^SO/, (1:10 molar- ratio) and then heated in air at 2S0C for ~ 1 4 hrs. The reaction products were subjected to chemical, thermal and X-ray analysis. II I.RESULTS AND DISCUSSION: The X-ray pattern of U3O7 used is in agreement with the data reported. The products formed by the solid state reaction of U3O7 with (NH^)2 SO4 was examined by IR and UV .spectroscopy. The IR ti.pect rum recorded in the region 4000-600 cm on samples made into KBr. discs showed absorption bands ascribable to NU4 , SO4 and UO2 ions. The UV - visible spectra obtained on solutions of the product in Cone H3PO4 showed the presence of both U(IV) and U(VI). Chemical analysis of the solution for U(IV) and total uranium, by the potent iometri c method gave a value of. 2:1 for the U(IV)/total U ratio. X-ray powder patterns recorded using Cu—K(alpha) radiation showed the solid state reaction product to be a mixture of , and (NH4) ^UC^CSOij)2 .The thermograw obtained by heating the sample in a flowing stream of dry air is shown in the figure. The final product obtained at 800 C was identified to be Back calculation showed that the mixture analysed has a SSC - 32.1 composition 2 ( NH4 ) z,U ( SO^ )2 2UO2(SO/,)2 in agreement with the results of X-ray,chemira1 and spectroscopic measurments discussed above. The TG curve can be interpretted to give the following ^'•composition sequence. 3UO2SO,, Just as in the case of U3OQ which when heated with ^^ at 25O°C gave n mixture of (NH4 ) ^IKSO*, ) 4 and(NH4)2UO2(SO4)2 in the 1 :2 ratio, U3O7 gave a mixture of the same two double sulphates in 1 he 2:1 ratio. Sincce in the reaction with at such a low temperature of 250 C is not expected to disproportionate U(V) if present in 1)367, these results indicate that U3O7 f.can represented as a mixed valance oxide of composition U2' *U'"07U O . Allen et al have observed that the EXAFS spectra of the B-U3O7 and UO2 have the same peak positions, but the relative intensities of the peaks are different, indicating a modified fluorite type structure. The XPS results, however, showed that the binding energy of the U4f pertk is significantly higher than that for UO2, indicating an increase in the average charge on the I) atoms. For U3OQ, the XPS data have been interpretted in terms of U and U cations in tlie crystal leading to the formation of the compound as IV.REFERENCES: 1.G.C.Allen, CD.Garner, D.J.Jones, I.Ross and P.A.Tempeet, J.Phys.Chem.,89, 1334 (1985). 2.G.C.Allen, P.M.Tucker and J.W.Tyler, Vacuum, 32, *t81 (1982) 3.K.D.Singh Mudher, A.Chadha and N.C.Jayadevan, Radiochem. Radiation Chem. Symposium, Tirupathi (1986). 4.G.C.Alleji and P.A.Tempest, Proc . P.. Soc . Lond - , A 406, 325 (1986). - TG '' OTA - DTAK)£V FIG. TG & DTA CURVES FOR TG 2mg THE DECOMPOSITION OF U3O7. - JL \ vn ^——"v * J ill f I J 1 V. I.I 190 390 590 790 TEMPERATURE/C SSC - 32.2 Position of Actinide Eiemants in the Periodic Tablet Evolution of a Generalised Font). T. Newton Nathanitil. Leser and Plasma Technology Oivition. Bhabha Atomic Research Centre, 8oenbay-4M §85. All the actinides and lanthanides are placed in III group and are shown separately, in the most popular forms of the periodic table, including the latest version of IUPACO,}). This has been leading to an avoidable debate (>>! ie. whether to place La and Ac or Lu and Lw 'jelow Sc and Y in the III-B gr&up. Apart from this, the group characteristic nature of most of the actinide elements, demands that a more appropriate position is to be given to thesa f block elements. Thb situation is more complex in the higher periods, in which new elements belonging to g block, h block, ... are to be included. Also, there seem3 to be no consensus in following « uniform and acceptable subgroup notation<4,2). The author arrived at a new form of the periodic table (see fig.), to successfully sort out all these issues, as described here. The striking resemblance of Th, Pa, U. Np and Pu to Hf, Ta, W, Re and Oa respectively is undisputable. However, the resemblance of Ce, Pr, Md, Pm and Sm, with Hf, Ta, W, Re and Os respectively, is not as striking as that of their congeners in actinides with Ce being the exception. At times when U and Th were the only actinides known, the position of the f block elements in the periodic table was guided by the lanthanides, which have many special features of their own and distinct from rest of the elements. Actually, in almost all the early versions(B,9) of the periodic table including that of Mendeleyev <1#), Th and U were placed in IV and VI groups respectivaiy below Zr and W. Also, Th and Ce were often given the position of Hf before 'its discovery. Since the number of lanthanides is more than 10, and also as they come before the 5d transition series, they could not be placed with the third row of the transition series. Hence the necessity to keep them separate. Later, whan the transuranium elements were being discovered one by one, they are given the corresponding places below lanthanides. In that process, the remarkable group character stic nature of the lighter actinides. which is less pronounced in their lanthanide congeners, has been ignored. Thi6 aspect is well known in the d block elements, where lighter members of the 5d transition series show most stable oxidation states characteristic of the raspective group number: ie. Hf<+4>, Ta(+5). W(+6). Re<+7) and 0s(+8), which belong to IV, V, VI, VI3. and VIII groups respectively. Interestingly, the last two members of the f block elements also show oxidation states consistent with I and II groups. The striking resemblance of No, having +2 as the most stable oxidation state, to the II group alkaline earth metals and the conspicuous absence of +5 oxidation state, support this. The same is true to some extent with Yb, which has a stable +2 oxidation state. Also +1 oxidation state is known for Md. though it is not known for Tm. Thus, the extension of 8 group classification from the 8 columns of the s and p blocks to the 19 columns of the d block elements, can be very conveniently adapted to the 14 columns of the f block elements also. Th. Pa, and U. with +4, + > and -1-6 as the most stable states respectively, perfectly fit in the IV. V. and VI groups and it is highly incorrect to place them in III group along with Sc. Y. La and Ac. Also, Np. known to have +7 as one of the stable utates. fits well in VII group. It has been established that Pu has SSC - 33.1 Fig. PERIODIC TABL2 FOR INFINITE NUMBER OF ELEMENTS WTTH A SYSTEMATIC SUBGROUP DESIGNATION an oxidation state equal to almost +8 in solid state in mixed oxides like Li«PuO6. Thus. in this systematic classification, Pu automatically falls in VIII group, while its congener Sm, which is not known to have any state higher than +> is getting forced in to VIII group as an exception like Fe and Ru in d block elements. Thus, the author tried to give actiiides and lanthamdes a befitting position, which has been denied hither to. In this process a generalised version of the periodic table (see fig.) with a systematic subgroup designation, has been evolved to accommodate infinite number of elements with out any ambiguity, while restricting to the hither to used and most acceptable eight group notation, by simply extending the classification to the g block, h block, i block, ... etc. of elements in the highsr periods. Seaborg(6> termed all the 32 elements with atomic numbers from 1Z2 to 153 (g and f blocks together) as "Supersetinides" and suggested to place them below the 14 actinides. Instead, the author's suggestion to place them as per the generalised form given here is more appropriate, as the identity of each of the blocks is retained here. Also, the g block is experted to contain just 18 columns, h block 22 columns. ... etc. Though the theoretical calculations suggest slightly different outermost electronic configurations due to the overlapping of the energy levels of the corresponding p, d, f, g. and h orbitals, where ever relevant, it is definitely rational and very convenient to place the elements in the groups as shown in this form. The subgroup notation suggested here is veiy rational, explicitly retaining the identity of most of the subgroups, while avoiding all the controversies. The subgroup designation A refers to s and p blocks and B, C, 0, E. ... refer to d, f, g, h, ... blocks respectively and it is no longer arbitrary. Thus, B, Al, Sc, Y, Lu, Lw, La and Ac all fall in 111 group, but conveniently sorted out into the respective subgroups A, B and C. However, VIII group remains with unique features, as a composite group consisting of several elements of diverse nature. from noble gases to the transition and inner transition metals, with al! the valencies from 0 to G, but in different subgroups. References. 1. New Notations in the Periodic Table. E. Fluck, Pure and Appl. Chem., 69, 431-36. 1988. 2. Some Reflections in the Periodic Table and its Use, W.C. Fernelius, J.Cbem. Ed.. 6}, 263-66, 1986. 5. Recommended Format for the Periodic Table of Elements, by Am. Chem. Soc. Committee on Nomenclature, K.L.Loeing, J. Chem. Ed., 61, 136. 1984. 4. Confusion in the Periodic Table, W.C.Fernelius and W.H.Powell, 3. Chem. Ed.. 59. 584-8, 1982. 5. The Position of Lanthanum (Actinium) and Lutecium (l.awrencium) in the Periodic Table. W.B.Jenson. J. Cham Ed.. 59. 634, 1982. 6. Prospects for Further Considerable Extension of the Periodic Table. G.T.Seaborg. J. Chem. Ed., 46. 626-34, 1969. 7. The Periodic System of Chemical Elements, J.W. van Sporensen, Elaevier. 1969. 8. Niels Bohr, Z. Phys.. 9. 1. 1922: Ref. 7. pp 137-38. 9. Alfred Wernur. Ber. Dsch. Chem. Gss., 38. 914-21. 1995; Ref. 7. pp 152-J4. 18. Mendeleyev, D.I. Ref. 7. PP 137-38. SSC - 33.3 SOLVENT EXTRACTION OF ORANIUM(VI) FROM AQUEOUS SOLUTIONS USING MONO-OCTYLPHSNYL PHOSPHORIC ACIDC MOPPA) P.D.Mithapara, V. Shivar udcsppa/ and H.C.Jain Fuel Chemistry Division, B. A. R. C. , Bombay 400 085. SUMMARY Studies have been carried out for the; extraction of U< VI) from different acidic media by mono- I.INTRODUCTION Or ganophosphor ous compounds; have bet?n widely rfc:<~'ogn i std as efficient extract ants for metals from acidic solutions. Mono basic organophorous scids have bten extensively studied and their mechanism of extraction of metals have been irgptiV tRii 6 ' ~^i-. Ifl Pvifit Ffiftt, the dih&sic organophosphorous esters of the type ( GO) P0< 0H> - ("where, n is alkyl, aryl oi alkylaryl group) have not .been studied intensively. Peppard et.al*^' in their studies on the extraction of trivalerit lanthanide =>nd act inide ions using HaMO II. EXPERIMENTAL HsMOiJ)P was isolated- isolated from its commercially available mixture, octylphenyl .~*cid phosprir Activity of '-"^U/ Ml of the organic phase ______„ . ^ D — • Activity of 3:3I'0/inl of the agueous SSC - 34.1 III. RESULTS AND DISCUSSION The distribution ratio of U o) U0-< HMOOP> s. H^ 2> and the equilibrium constant Ke* is given by, a UOs(Hy>a. HaY (o) M- "(a) DxFx H" \ S U0a** (a) HSY - lo) H^Y -(o> Where F is the anicn complex ing factor foi tthhe metal ion in the aqueous phase used. The distribution fipl.o wois used *• o Calculate equilibrium constantfKex^ for the above react inn. The values are given in Table-1. It is in the oid^r: HC104 > HNOa > HC1 > H-SO^ which is in sjreement with the kwown complex ing tendency of UQ2++ by these an ions. IV. REFERENCES 1. C.A.BJake, C. F. Col eman, K. B. Brown, C F. £Ws and J.C. White 2nd International Conference on the Peaceful Uses of Atomic Energy, Volume 28, United Nations, (1958)289 D.F.Peppard, G.W.Mason and S.McCarthy, J. Tiiorg. Nucl. Chem. , 13, (I960) 138. G. M. Mason, I Hilobran and D. F. Peppard. J.Inorg.NUcl. Chem. 40, f 1978) 1807. 4. C.A.Baes,Jr, J. Inorg . Nuc 1 . Chen.. , 24, M962>707. 5. D. F. Peppard, G. W. Mason, W.J.Driscoll and R. J.Siron«n, J. Inorg.Nucl Chem. , 7, (1958) 276 D. F. F'eppard, J. R. F'erraro and G. W. Mason, J. Inorg. NUcl. Chem, , 7, (1958) 23 1. Table 1 Equilibrium Constant values for U Acid Log Kg., 9. 67 HNO= 9. 52 Hcr 9.56 9. 44 •* calculated after correcting for Chloride complex ing of U( VI) ** calculated after correcting for sulphate complexing of CKVI) SSC - 34.2 STUDiES ON THE OXIDATION BEHAVIOUR OF i NTERMETAL.L IC UPd,, II. T.Nswada, P. Si i i ama inurti, (j.Seanivssan and 1.1-^11 l-ipp^n Kadior-liemisti y programme?, I nili ra Gandhi Cenlio ,'ui Al umio Research, K H I [)B k V. a in HO? 1 < >'/ . .SUMMARY : • The oxidation process of U - PiJ alloy has bpsii invest igdtpd in detail by t he r mug r a v i in t. i y a rid with support irijj evidences f i om oxidation tit Uranium and Pa I I ri i y in metal to umiui s>?n KEY WORDS: UFd.-,, Pa I 1 ad i uin , Uranium, t hermog ravi me t 1 y , ox i it.* r 1 i.n I.INUNUDUCJTION: Actinide (U,Pu> - nubk metal (Ru, Hh, Pii< a I I u y * .411- ..( cmrent interest, sincp it has been found that the riuM*- n . t ^ I alloy inclusions ar& formed in fast reactor fuels during ii radiation. /I/. 1 lies e metallic alloys are alsj { uuiui in the residue frfs the dissolution of carbide f DP ! i i- 10 N n i > • 1 c acid/2/. Oxidation of fuel as a single ci.ep or oxitlat joti of insoi ubies to separate actinides i r. to more soluble f u 1 in is considered. Wsjbenga /3/ has mpr. cioned incomplete oomi'ws t i on of the j i 1 oy due to coating of ; e I eased Pd over alloy in his bo ml) calorimetry experiments No information exists on the oxidation behaviour of U P d3 . 11.EXPERIMENT/^:- U-Pd alloy was preapared by arc melting followed by annoaling in a vacuum sealed quartz capsule at BOO -850° C and I h& phase was confirmed by XRD- Oxidation behaviour of the alloy ond palladium was studied employing the Neztch STA 4 29 t lie 1 mo I analyser using heating rate ranging from 0.2 to 2O°C/min in tl.a nonisotherma1 mode and also in isothermal mode <40O-fc)O0' C>. The reaction products cor respond i tig to the i nte rut d i a to stages of oxidation were soaked at thg respective temperatures for sufficient period of time and the cooled products were analysed for Pd by giavimetry with Nioxime as comp i <=> x i ng agent and plu)se characterisation by XKD. II I.RESULTS AND D1SCUSSI0N:- Oxidation characterstios of U-Pd alIoy have been found to b* strongly influenced by heating rate as shown in fig.], illustrating effect "of heating rate on oxidation, viz., describing the variation of the maximum weight gain and weight plateau. As heating rate is reduced from 5 to 0.2" C/min the extent of oxidation lias increased from 50% to 90%. In order to compare the progress of the oxidation reaction as a function of temperature, thermogravimetric results from nonisothermaI runs of U-Pd alloy, U metal and Pd metal are shown in fig.2 which is a plot of overal 0/M as function of temperature recorded rollowing trie heating rate of 0.2*C/inin. The air oxidation of U Pd alloy incepts around 340 C and reaches maximum weight gain plateau around 750°C and on futher heating a sharp weight loss step was noticed around 778t C which corresponds to PdO decomposition, lntial stages of oxidation of U and U Pd alloy were similar upto 350° C. With f-urther increase in temperature uranium reacts rapidly compared to U-Pd alloy. Two peaks in L>!G curves in the first weight gain step of U-Pd alloy oxidation and quantative chemical analysis of Pal ladium has indicated uraniut.. preferrential oxidation leaving Pd in U-Pd alloy upto 450 C. At temporatures above 450*C Pd also undergoes oxidation From the in ox i dat ion-1 hermogramme of U-Pd alloy,Pd & a -synthetic mixture of U03 + PdO it is seen that PdO decomposition in U-Pd alloy is quite sluggish compared to pure Pd & synthetic mixture, eventhough decomposition temperature was not much affected. Similar systematic studies in understanding mechanistic aspects of oxidation of URu-& SSC - 35.1 REFERENCES: 1 . J.l.Br-amman.R.M.Sharpe.D.Thron S* G. Yat.-s, Jour Nucl Mat 1968,75,201 . 2. H.-K ] e> Kamp &< R.Pejsa, Jour Nuc i Mat 1984, 124, 56. 3. G.U'ijbenga, Jour Chem Thermmodynamics,1982,14,483. 1 0.2 2,05 3 2 0 deg/ rn/n i, 5 0 diV min 5 20 Odeg/ mm 500 700 900 TEMP °C "100 300 500 700 900 TEMP. °C FIG. 1 VARIATION OF OXIDATION THERMO- GRAM OF Upd^ WITH HEATING FIG 2 COMPARISON OF OXIDATION PATTERN RATE OF UPd, ,U AND Pd SSL - 35.2 EFFECT OF PREPARATION HISTORY ON OXIDATION CHARACTEREST1CS OF URANIUH MONO CARBIDE ll.P. Nawada, P.Srirama murti, G.Seenivasan and S. AnUmyssmy Rad i ocliedi i s t ry programme, Inuira Gandhi Centre for Atomic Research, Kalpakkarn- 603 102. SUMMARY:- The factors control]ing the oxidation behaviour of UC viz., preparation history and heating rate have been investigated by therrnogravimetry and suuported by chemical characterisation experiments to deliniate nature of oxidation process. KEY WORDS: Oxidation, ignition,thermogravimetry, preparation history, uranium mono carbide. 1 . 1NDRODUCT] ON : - Studies on the oxidation and ignition behaviour of caibide fuels are of interest : i) as a possible head end stage of reprocessing ii) in the context ofconversion of fuel fabrication waste for refabrication iii) for safe intermediate storage of irradiated fuel in the form of oxide iv) proper handling of pyrophoric carbide/1/. The literature available on kinetics and mechanistic aspects of oxidation in high oxygen pressure leading to bulk oxidation. process is misleading due to ignition and pretest sample history/2/. As carbide ignites, the stages of oxidation which it undergoes need not be similar with the oxidation under programmmed smooth conditions because of localized sharp raise of temperature at few spots. A knowleflge of the ignition behaviour of carbide is necessary in order to Ofjtimiza the conditions for a control le'd oxidation of tho carbide. There were no systematic studies in understanding the role of ignition &< pretest history in oxidation of UC. We have recently studied/3,4/ the ignition and oxidation behaviour of UC2 In the present work detailed E.tudies have been carried out on the oxidation and ignition of behaviour of UC primarily employing thermogravimetry and supplemented by chemical analysis and XRD. I I.EXPERIMENTAL:- In the present work uranium carbide pel lets were prepared by carbothermic reduction and were denoted "CTR-UC" henceforth. Some CTR-UC pellets were repeatedly arc melted in water cooled hearth furnace and the carbide thus obtained from the arc melting was called "AM-UC". Some of the AM-UC buttons were annealed in vacuum sealed quartz capsule at 600°C for one month which were called as "AM & AN-UC". Similar procedure was followed for obtaining uranium metal viz., arc melting and annealing. The air oxidation of UC was followed by thermograv i rne t ry employing Neztch STA 429 thermal analyser. The reaction products corresponding to the intermediate stages of oxidation were soaked at the respective temperatures for a sufficient period of time and the cooled products were analysed for 0/U by spectrophotometry and phase characterisation by XRD. 111.RESULTS AND D1SCUSSI0N:- The ignition temperature(Tig) of CTR-UC showed a gradual increase from 170°Co to 210" C as the heating rats(B) decreased from 20° C/min to 0.5°C/min. When the heating rate was further reduced to 0.2 C/min, CTR-UC did not undergo ignition, indicating that at sufficiently slow heating rates ignition car be altogether prevented. A linear correlation was noticed between the reciprocal of Tig of CTR-UC and iog(B). Ignition temperature has exihibited profound dependence or. the preparation history of UC e.g., the CTR-UC ignited around 190°C, AM & AN-UC ignited around 300°C and fresh AM-UC could not be ignited with in 600" C under 5" C/min programmed heating conditions. It appears thcit increase in oxygen content in UC and lowering of density of UC decreases ignition temperature. SSC - 36.1 Fig.l shows the effect of the heating rate on the oxidation thermogram of CTR-UC. Fig.2 shows the oxidation thfirmoerams of CTR-UC, AM-UC and AM & AN-UC under 0.1°C/min heating rate, viz., illustrating the variation of the maximum weight gain, weight plateau and inception temperature. It was found that AM UC reacts more sluggishly than that for AM-U: With annealing of AM-UC, carbide loses its oxidation resistance and oxidation inception temperature is lower than AM-U. It was found that the air oxidation of the CTR-UC proceeds through the intermediacy of uranyl carbonate adduct, << -U0, and the later decomposing initially to {JO and finally to U, 0B REFERENCES: 2"9 3 8 1.U.Benedict, M.Richter, J.Tuerlimx, Trans.Amer.Nuc.Soc. 31 19 79, 512 2.H.Matzke, Science of LMFBR fuels North Holland, Amesterdam.1985 3. H. P. Nawada, F'.Srirama murti, G.Seenivasan and S.Antonysamv Thermo.Chim.Acta, 144, 1989, 357. 4 . H. P.Nawada, P.Srirama murti, G.Seenivasan, S.An tonysamy a.u C.K.Mahtews J . Therm. Ana.1 . 39, 1989, 1145. TEMPERATURE'C IQD 300 400 500 600 700 TEMPERATURE'C FIG.1 COMPARISON OF CTR-.,: OXIDATION BEHAVIOUR FIG. 2 UC OXIDATION DEPENDSNCE ON PREPARATION AT DIFFERENT HEATIN3 HISTDRY SSC - 36.2 DIFFUSION OF MANGANESE IN ST.AINi EbS STEEL V. Ganesan, P. MurJ!idharan, K. Chandran and 6. Periaswami Radiochemistry Programme, Indira Gandhi Centre for Atomic Research, Kalpakkam - 603 102. INTRODUCTION Release and transport of radionuclides of fairly long half-lives such as Mn-54, Co-60 and Cs-137 through liquid sodium in the primary circuits of fast reactors would lead to high radiation fields, resulting in operation and maintenance problems [1]. Upon prolonged operation, these nudides get deposited in various regions of the coolant circuit and penetrate into the stainless steel matrix by diffusion thereby making simple decontamination procedures inadequate. A knowledge of diffusion coefficients of various radionuclides in stainless steels, being used as structural materials, becomes very important. These data are also important in modelling the activity release that occur in the core of an operating fast reactor- This paper describes the tracer technique used in determining the diffusion of Mn-54 in stainless steel type 304 in the temperature range of 723 - 9?3 K. Measurement of concentration profiles :if Mn-54 in stainless steel was achieved by residual radioactivity technique. EXPERIMENTAL Sodium containing a known activity of Mn-54 was taken in an alumina crucible. Am.edled samples of austenitic stainless steel type 304 was kept immersed in liquid sodium. The alumina crucible along with its contents was placed inside d stainless, steel reaction vessel and was secured leaktight using knife-edge flanges employing annealed copper ring as gasket material. The entire operation was carried out inside an argon atmosphere glove box. Diffusion annealing of the sample was. carried out at suitable temperatures ranging from 723 - 923 K for specified lengths of time. At the end of diffusion annealing, the sample was cleaned free of sodium and the residual activity of Mn-54 was measured using a well type Nal(Tl) detector. Concentration profile of Mn-54 in the sample was obtained by measuring the residual activity as a function of distance from the surface after removal of the successive layers by electropolishinq technique. A mixture of phosphoric acid, sulphuric acid and distilled water in the ratio of 65:35:15 by «volume was used as the electrlytic bath and a current density of 0.25 A cm was employed. RESULTS AND DISCUSSION Since the concentration of Mn-54 on the sample surface is maintained constant and the diffusion path of the nuclide is extremely sm,ill in comparison with the actual thickness of the sample, the solution of Ficks second law of diffusion as applied to semi-infinite body is considered applicable in the present case. The residual activity in such cases can be related to the thickness of the layer removed as indicated by the following expression. „ where, A = residual activity (EJq) Co x Co = concentration of Mn--54 on the In A = In "-"^^- surface of the sample (constant) 2\//T Dt 4Dt D = diffusion coefficient (m /s) t = diffusion annealing time (s) x = diffusion length (m) SSC - 37.1 T ( a .11 650 600 K0 503 10 11 •!. \ RKKEM .i-i. ; o PRESENT \ 1 \ 1 \ -17 \ 10 \ 2 3 i 5 E Squore oi ihs Thickness removed from surface 1.2 Fig.'l. Residual activity of Mn-54 as Fig.2. Bulk diffusion of Mn-54 in 304 a function of x at 773 K. stainless steel. Fig.1 shows the typical plot of In A vs x at 773 K. . From the slope of the plot the diffusion coefficient at 773 K is calculated. Fig.2 shows the variation of log 0 as a function of reciprocal temperature in the temperature range of 723 - 923 K. The diffusion coefficient can be expressed by the Arrhenius expression: -89536 D = 2.47 x 10~12 exp where R = gas const. (J mol" K~ ) RT Also shown in the figure is the diffusion coefficient at 855 and 866 K reported by Brehm et al.[2] and Sagawa et al.[3] respectively. CONCLUSIONS The diffusion coefficient of Mn-54 in austenitic stainless steel type 304 in contact with liquid sodium was measured in the temperature range of 723 - 923 K by residual activity technique. The activation energy for the diffusion process is calculated and is given as 89536 J rnol" . REFERENCES [1] H. Fuerstein, A.J. Hooper and F.A. Johnson, Atomic Energy Review, 17 (1979) 697. [2] W.F. Brehm, P.L. Koehmstedt, E.A. Kovacevich and O.W. Shannon, (Proc. Meeting Bensberg, 1971) Report IWGFR, IAEA, Vienna, paper 8. [3] N. Sagawa, H. Ib<>, M. Urata and Y. Ozawa, J. Nucl. Sci. Technol., 13 (l976) 35°- SSC - 37.2 SST - Separation Science & Technology Papers : SST - 01 to SST -18 RECYCLING OF PLUTONIUM FROM FUEL FABRITAlION SCRAP A,R. Joshi , M.M. Charyulu, D.R. Ghadse, A.v. Kadam, U.M. Kasar, D.M. Naronha, S.M. Pawar, l.C. Pius, M. Ray, V.B. Sag II EXPSRIMENTAL Macroporous anion exchange resin Amberlyst A-76 was obtained from M/S Rohm 8< Haas Co. , Paris in chloride form which was converted to nitrate form before use. All chemicals used were of A.R. grade. Detailed procedure followed for ion exchange purification of Plutonium is described elsewhere . At the end of each cycle consisting of loading (7 M HIMOT) , washing (7 M HIMO3) , elution (0.5 M HNG-j> and recondi ti oni ng (7 M HNO,.) , samples of resin were removed from the coloumn bed and their ion exchange capacities were determined. 5 ml of the resin taken in a glass column was converted to chloride form by washing with 250 ml of 4 M HC1. This was washed with distilled water until washings were free from chloride. The resin bed was subsequently washed with 100 ml of 4 M HNQ^ and washings were collected in a 10ifl ml flash. 5 ml of the solution was diluted with 50 ml of distilled water and adjusted to pH 5 with NH^OH. Chloride in SST -01.1 this was deter fni netl by adsorption indicator method using standard AgNQ-.r solution . The capacity of the resin was calculated -from these dat a. Concentration o-f plutomum in the ion exchange product solution was sufficiently high ("5(3--4 fj g/1) and the plutonium -from the solution was directly precipitated as f-'u(IV) oxalate after adjusting the oxidation state and acidity. Precipitation was carried out Jl in batches from Pu(.lV) fetid solutions having initial HNQ-r ranging from 3 to 4.4M. The solution was heated at 55DC and plutonium was adjusted to f-'u(lV) by addition of 30Z H-^O-p ( 1 ml/g of Fu). PIutonium(IV) was then precipitated at this temperature by adding 1 M oxalic acid (8.4 ml /g of F-'u) . Additional oxalic acid was addt-'d ta male the SII|ILT luUml W. t M in o>: ctl .-_* ti?. fhe precipitate was digested at U'.J C for one hour and left overnighL: for settling. The precipitate was filtered using Whatman paper 541. washed initially with 2 M UNO., i W.IS'J M oj;aiic acid mixture and finally with absolute alcohol. The precipitate was dried by sucking air through it and subsequently calcined in a flow of oxygen at 55(3°C for 1 hour. PuQ-> thus produced was analysed for PI.I content and the impurities. The surface area of PuO., was also measured by BET method in Radi ometal 1 urgy D:i vi si on . Ill RESULTS AND DISCUSSION When the HNO-.: concei itr a L I on of plutomum solution varied from 3 M to 4.4 M in the oxalate precipitation, the losses of plutonium in the supernatant increased train r-> 20 rag/litre to r-J 5(3 rag/1 itre . However the precipitate obtained at lower acidities were finer taking more time for settling and filtering. Finer particle size of the precipitate obtained at lower acidities was corroborated by measuring the surface area of the PuO , obtained by thermal decomposition of the oxalate precipitate. The surface area of PuO ••-, obtained from oxalate precipitated from 3.4 M HMO.- sne'Jium was > 2W rn^/g of PuO^ and the corresponding figure when precipitation was carried out from 4.4 M HNO-.T was < 10 m^/g of PuOv- lon exchange capatitieis of the Amberlyst A—26 resin determined before its use and after 3rd, 4th and 6th cycle of operation are 0.92, 0.86, 0.85, 0.81 nit?q/.ill respectively which indicate that there is only marginal decrease in its capacity. This suggests that the same resin can be reused successfully. • ACKNOWLEDGEMENT The authors express their thanks to Rhri B.C. Jain, Radi ometal 1 urgy Division for measuring the surface &reat of PuO^ samples. REFERENCES 1. M.M. Charyulu et.al, Recovery and Purification of Plutonium from fuel fabrication scrap using macroporous anion exchange process, Int. Symp. on Thermochem. & Chem. Process.,1989, Kalpakkam, 2 Vogel 's te?>:t book on quantitative Inorganic Analysis 339, 4th Edition, Enqlish Language Book Society (1985>. 3. J.M. Cleveland, "The Chemistry of Plutonium" published by Gordon and Breach Science Publishers, p. 527 (1970). SST - 01.2 HliMOVAl 01 IOW I rvill.S OP URANIUM I'ROM AOUllOUS SOIOTrONS (!Y COI'RrCIPJTAriON AND ION rxrilANGf M.Sudersanan and H.K.fyer Analytical Chemistry Division Uhabha Atomic Research fontie fiuinbay, lloinliay /,00 08G SUMMARY ( oprecip i1 ation of uiaru um( VI) from aqueous solution:; with fciiic: hydroMide has boon evaluated as a nii-ans of roraovjng uraiuuin from aqueous; effluents. I xpei iment a with different amounts of uranium arid added carbonate showed I.hat. it was possible l.o remove better l:han 957 of uranium in a single precipitation yf low concentrations of uranium. Sorpti.on on weak acid cat: ion exchange resin has also been studied and can be used if the uranium is to tie; iccovciwl. (Keywords: Uranium. docunt.amiiiiili.on, coprec i p) 1. a t. i on , ion exchange) INTRODUCTION: The removal of trace levels of uranium present in effluunt streams of NI:C Hyderabad, required the dove) opment of a simple and economic mot hod Al the same time, * ho method should not give rise to any cnvi ruranenU 1 problems. Precipitation of urur • •!"» in a suitable foim is acceptable from those considerations, ' < the incomplete prec.ipi tat. i on of ammonium di uranalu from .ol.uti.ons containing tarlion dioxide and !he low levels of uranium necessitated the choice of an alternate method. Coprecipita I: ion with ferric hydroxide was therefore tried. Treatment of t fu solution with weak acid cation exchanger offers another possibility of removing uranium from solutions, though the material is more expensive. >:'XPI:ttIMI:N''.AI A solution of urani um( V ( ) containing 1.7? mg/ml was prepared ami suitable aliquots were used. Sodium hydroxide, ammonium hydroxide and hexamine wore used as thr proci pi tant.s. Oowox CC? resin was used as the weakly acidic cation exchange tesiri. Vstimation of uranium was carried out using pyridyla^o resorci.nol (I'Att) by treating the solution containing 0-00 \ig of uiarnum wilh 1 ml of 0.1 M IOTA, 1 ml of 0.0?7. PAH and ? ml of ?0T. ammonium hydroxide. The solution was made upt o 10 ml and the optical density was measured at C>10 nm vs. water blank. Copi ec ip i t at ion studies were can led out witti sodium hydioxicle. ammonium hydroxide and hexamine as precipitants . I'ho amount of iron was kept constant at 0 oi 16 mg but the volume of solution was varied from ?.'} to ?'JO ml. Sodium chloride (0.0'j to 0.10 M) was used to flocculate the precipitate and 0.06 M NalICO was added to simulate 1 he- conditions of carbon dioxide saturated solution. Precipitation was carried out in the cold. Solutions containing 7 to If? ppm of uranium, sodium chloride and NailCO. so as to have a pit of 8 was treated with 8 to 16 mg of ferrictill) ion. The pll which was acidic at this stage was adjusted to alkaline pll with the precipitant. The concentration of uranium in 1 ho supernatant solution was estimated as described SST - 02.1 cat) i Z of uianium f I oin dilute solutions [7-'I ppiri). I ri I he ia<,e of higher ( nn< cut ta I i on •: of uiauiiim containing a liighcr amount of N.illCO. , Z of in <• i uni and a double preripjl,)! ion reduced I lie i. urn en) ra t i on ot uunuim below 'he ili'tci t IUM 1 i in I 1 of 0 O'J ppiri wh 1< h is cons i rlorod safi.1 for disposal the results in t IK; I aii1 of hex aw i nc was < omp;u ahi e In the ';i>1 nb i 1 i I y r> f Mil in wsli'i w;i'. a 1 '.:n >••; t i in;i led. II w.i •; f oi/rid thai 1 hf i omciitiiil inn nf vi i .m i inn «i;i •. 1 !••. f()N fXCHANfil' SlllOriS: fon fxrb.inge uf uranium by CC? resin (weakly «i i d c i ( a i boxyl a I <• I ype i u' lhu usu of this l'usni for I.I10 romoval of up to 69 ppm of uranium from solutions i out i\ 1 n i rig O.Oli 1o n.tO K sodium chloiicle was studied. When the solution pll was ', In 6 .'i the.- uptake was over 91) I will) ?r> ml of solut ion and O.?'J 9 of n:r:iri. llowevoi I he pll of the solut ion has to be udju^iod 1 ti l:he pll range of 1 1.0 I ('OM(-I UNIONS : Ammonium riydtox u!" or tiexainine Hie equally eFfeitive fin I: ho removal of uianium present- in solul ion. Sodium tiyiJioxnie i :; 1 ..> s i effective A double prec i pi ( a I i on is nei esisar y foi fuyhei oont out 1 al i oni of uianium su as l;o reiiuie tin; i:oin:i;nl 1 a 1.1 on of uranium to a< 1 t-pl alii e levels The use of a 1 a I 1 wi exitiamjei is ail.'aitiv if t.hu lOiK^enl.ial ion of uranium is high enough l.o warrant, lurovery of 111 an 1 uin . SST - 02.2 SEPARATION OF URANIUM AND IRON BY ANION EXCHANGE IN HYDROCHLORIC ACID Mary Xavier, K.A.Mathew, P.R.Nair, B.N.Patil, H.C.Jain Fuel Chemistry Division, Bhabha Atomic Research Centre Trombay, Bombay - 400 085, India SUMMARY A procedure is described for the separation of uranium from iron by adsorbing it on Dowex 1X4 anion exchange resin in 9M HC1 as the anionic U(IV)-chloride complex. Ti(III) is employed for the reduction of uranium to the quadrivalent state, and iron to the ferrous state, which is only slightly adsorbed. Decontamination factors ranging from 150-600 were obtained and the recovery of uranium was better than 9b%. Introduction o r> o U is an important fissile material and has to be recovered from analytical waste solutions and purified. The major impurity associated with uranium in such solutions is. iron. This paper presents the results of some of the studies carried out to separate uranium and iron by anion exchange in HC1 medium. Ti(III) is employed for reducing iron to unadsorbable Fe(II) and uranium to U(IV) which is adsorbed as its anionic chloride complex. Exj>er imenta 1 Solutions of uranium, iron and titanium respectively were prepared in 9N HC1 from uranium metal, iron powder and titanium sponge. Dowex-lX4 resin was used as the anion exchanger. Known amounts of uranium and iron were taken in a beaker and Ti(III) solution was added till the violet colour of Ti(III) persisted. This solution was transferred quantitatively to an ion exchange column (1.5 cm dia and 15 cm long) which was conditioned with 9^ HC1. The flow rate was kept at 5-6 drops/rain. The column was first washed with 9N HC1 containing Ti(III) and then with 9N HC1 alone till free from Ti(III). Elution was done with 0.2N HC1. The eluate was analysed for uranium by titration^ ' and iron by spec^rophotornetry using o-phenanthroline as the complexing agent^'. Results end Discussion In hydrochloric acid solutions of higher concentrations, U(VI), U(IV) and Fe(III) form anionic complexes and are adsorbed on strong base anion exchange resins. A look at the distribution coefficients of these ions shows very high values for U(VI) and Fe(III) in HC1 concentrations above 6N whereas U(IV) shows appreciable adsorption only above 8N acidity^-. Fe(II) has no adsorption below 7N acidity but shows slight adsorption at higher acidities. From these data, with the selective reduction of SST - 03.1 iron it should be pofisible to prevent the adsorption of iron. Attempts to reduce iron selectively with hydroxylamine hydrochloride, and sodium bisulphite were not successful. Therefore, the feasibility of adsorbing uranium as U{IV) after reducing both U and Fe to U(IV) arid Fe(II) has been investigated. Titanous chloride was found to be a suitable reductant for this purpose and it has only negligible adsorption in 9N HC1. Experiments were carried out'at different conditions and the decontamination factors determined. In some of the experiments the columns were left overnight after loading and washing was done only with 9N KC1 on the next day. The decontamination factor obtained in these cases were about 10. This could be improved to about 25 when the loading and washing were done on the same day. It was seen that washing the column with HC1 containing Ti(III) gives much better separation from iron. The best results were obtained when loading and washing were done on the same day and the first few washings contained Ti(III). Thus decontamination factors hanging from 150 -600 could be obtained. Table-1 summarises the results. As seen in the table the recovery of uranium was better than 95%. This procedure was employed for the purification of from analytical waste solutions. References 1. P.R.Nair, K.V.Lohithakshan, Mary Xavier, S.G.Marathe, H.C.Jain, Radiochemistry & Radiation Chemistry Symposium, IGCAR, Kalpakkam, India, Jan. 4-7,1989,Paper RA-22. 2. J.E.Rein, G.M.Matlack, G.R.Waterbury, R.T.Pelps, C.F.Metz (Ed.). Report LA-4622, 1971, 55. 3. K.A.Kraus, F.Nelson, Int. Conf. Peaceful uses Atomic Energy (Prcc. Conf., Geneva, 1955), 7, UN, New York (1956) 113. Table-1 Results of anion exchange separation of uranium and iron Uranium Iron s. Taken Found in Recovery Taken Found in D.F. No. (mg) eluate \ fo ) (mg) eluate (wig) ( MS) 1 63.7 61.6 96.7 53. 5 194 266 2 63.6 61.9 97.4 50.5 194 253 3 25.4 24. 1 94.8 62.3 248 238 4 60.6 58.3 96. 1 105.6 542 187 5 60.2 58.3 97.0 103.2 664 150 6 54.3 53.2 98.0 ! 102.3 204 ! 491 7 56.9 54.2 95.3 91.5 150 581 SST - 03.2 MASS TRANSFER CHARACTERISTICS OF URANIUM IN DI(2 ETHYL HEXYL)PHOSPHORIC ACID -TRIOCTYLPHOSHINEOXIDE-KEROSENE SYSYEM. K.N.Hareendran, Pushparaja* and M.S.Subramaniam** Uranium Extraction Division Bhabha Atomic Research Centre Bombay-400085, India. * Health Physics Division ** Radio Chemistry Division SUMMARY The mass transfer characteristics of uranium in di(2- ethylhexyl) phoshoric acid (D2EHPA)-Trioctylpho3hineoxide(T0P0) in kerosene medium have been studied. The extraction runs of uranium(vi) from 2M HNO3 solution were carried out in a specially designed extraction cell of Lewis type. The effects of interphase area and stirrer Reynolds number were investigated for the sys- tem. The overall mass transfer coefficient increases with in- crease in Reynolds number. However, the values for the coeffi- cient decreased with increase in interphase area. (Key words; Mass transfer coefficient,synergism,uranium,organo phosporous compounds) I. INTRODUCTION Extraction of uranium from HNO3 solutions using syner- gistlc mixtures of organophosphorou3 compounds has potential ap- plication in the reprocessing of spent reactor fuel. The general objective of the study of mass transfer chracteristics of uranium is to aid development of the extraction process. In the present study, overall mass transfer coefficients of U(VI) from HNO3 solution using D2EBPA-TOPO synergistic mixture in kerosene medi- um, were determined using a baffled extraction cell of Lewi3 type. II. EXPERIMENTAL The details about the extraction cell (Fig.l) and exper- imental arrangement are described else where /!/. The aqueous phase containing 5 g/.L natural uranium in 2 M HNO3 and the organ- ic phaae containing 5% (v/v) D2EHPA (BDH) + i% (w/v) TOPO in kerosene were introduced into the cell carefully. The two phases separated by a SS plate of known open area to serve as interphase area, were stirred independently using a contra-rotating stirrer system with control mounted at the top of the cell. The extrac- tion runs were of 180 minutes duration with constant stirrer speed. Samples were collected periodically for uranium estima- tion. The equilibrium data required for the study were obtained by shake out tests by varying aqueous to organic phase ratios at room temperature (24°C). Reynolds numbers for the system were calculated by measuring the viscosity, density and knowing the impeller dimensions and stirrer speed. I]I. RESULTS AND DISCUSSION The overall masstransfer coefficient,on the basis of 3ST - 04.1 organic phase, was calculated using the equatipn- Ko = Vo/A t*A / Co(t»_ U" - Co where Vo is organic phase volume, A isi Tn"the einterphase area At is the time interval of the run, Co is the uranium concentration in the organic phase and Co* is the equilibrium concentration of uranium in the aqueous phase (obtained from the extraction iso- therm for the system). The integral in the above equation was obtained graphically. The overall mass transfer coefficients(Ko)_ obtained for different interphase areas were as follows: 2.13*10* cm/s(50cm2); 1.73*10"*cm/s(86cm2) and 1.65*10 cm/s(116cm2). There is slight decrease in the values as the area increases. Fig-2 presents the Ko values Vs Reynolds numbers, which shows an increase in Ko with increase in Reynolds number. During the extraction run, the Ko had initial higher value which decreased with time. In the absence of turbulance at the interphase, the build up of a strong synergistic complex species ai the inter- phase may result in the increase in interphase viscosity which suppresses the surface renewai/2/. This hinders the diffusion of free extractant molecules to the interphase and decreases the mass transfer process. However, at higher Reynolds numbers, the surface renewal i3 enhanced resulting in higher values for mass transfer coefficients. IV REFERENCES. 1. J.V.Abraham, MSc Thesis, University of Bombay,(1988) 2. D.ROGERS, P.J.Thompson and J. D. Thronton, -I.Chem.E. Symposium series No.103, (1987) pp 15-27 A2 HP GEA3EO MOTOR PL( T OF IWSS RANSF iR CO JFFIQ TNT Vs IEYNOL3S NUMBER FOR d- 02 EH -TOHD SYS EM. t)£VEL ttEAK SYSTEM 2.0 -- SAMPI tNG HOI FS 15 SAFFI.ES PHASE SEPARATOR 10 OS - — STII CON I. EVEL 500 1000 1500 2000 2500 3060 AOJUSTORS REYNOLDS NUMBER SST - 04.2 SEPARATION OF URANIUM AND PLUTONIUM USING MACKOPOROUS ANION EXCHANGE RESINS FROM MIXED SOLVENT MEDIA - COLUMN STUDIES K. V. Chetty, A. G. Godbole, P. M. Mapara and Ra.iendra S war UP. Fuel Chemistry Division, Bhabha Atomic Research Centre, Trombay, Bombay - 400 085, INDIA. SUMMARY Column studies on the separation of plutonium and uranium using macroporous anion exchange resin from nitric acid + methanol medium were carried out. From the results obtained it was observed that a good separation of plutonium and uranium could be achieved. Key words: plutonium, uranium,macroporous resin, separation, mixed solvent. I.INTRODUCTION Ion-exchange in solutions containing water miscible organic solvents such as alcohols, ethers, ketones show great promise in the separtion of metal . ions^ '. The present studies were undertaken to find out the possibility of separation between plutonium and uranium using macroporous (MP) anion exchange resins from mixed nitric acid and methanol medium. The distributioin data* ' for Pu{IV) and U(VI) obtained earlier indicated that a reasonably good separation of Plutonium and Uranium could be achieved. Further, column experiments vere initiated and the results of which are reported here. II.EXPERIMENTAL In the present WGrk an ion-exchange column of 8mm I.D was used. The resin bed volume is of 5 ml in water. The column was conditioned with a mixture of 1 M HNOq and 60% methanol before loading. During loading and washing the residence time of 15 minutes and during elution the residence time of 60 minutes were rr.ai ntained. The breakthrough capacities for Pu(IV) using macroporous an ion-exchange resins Tulsion-A27, Amberlyst-A26 and Amberlyte-XE 270 from 1 M HNG^ + 60% methanol medium were determined using the reiati on Breakthrough(%) = (C1/CQ) xl-00 where C^ = Concentration of Pu in the effluent CQ = Concentration of Pu in the feed. III.RESULTS From the results (Table 1) Tulsion-A27 was found to be better as compared to the other two and further experiments in the presence of uranium was carried out with Tulsion-A27 only and the results are included in the Table 1. The washings were SST - 05.1 carried out with the same solvent mixture, 1 M HNO3 + 60% methanol. In the washings uranium was found to be almost completely removed from the column in about 5 column volumes(from 13.3 mg/ml to 0.24 mg/ral). The loss of Pu in the total washings was found to be negligible. Plutonium was eluted using 0.35 M nitric acid. The data on the product recovered are given in Table 2 which indicated almost quantitative recovery in 5 bed volumes. Major portion of the loaded Pu (around more than 95% of loaded Pu ) can be eluted in 3 column volumes giving good product which is useful for further processing. From the results, it could be inferred that a good separation of Pu from U is possible from 1 M HNO3 + 60% methanol medium. IV.REFERENCES 1. J.Korkisch, Progr.Nucl.Ener.Ser.IX, Eds.D.C.Stewart and H.A.Elion, Perg.Press. Vol.6(1966),pi. 2. K.V.Chetty, P.M.Mapara and Rajendra Swarup, Radiochem. & Radiation Chem. Symp., IGCAR, Kalpakkam, Jan.1989, ST-19. TABLE 1 Breakthrough capacity data for Pu(IV) Resin %Breakthrough Capacity g Pu/1 resin Feed I [Pu] = lir.g/ml Tulsion-A27 0. 12 19 [U] =Nil 0. 32 31 Solvent=l M Amberlyst-A26 0. 38 13 HNOo + 60% AKiberlyte-XE270 1.0 3 Metnanol Feed II As above + Tulsion~A27 0. 13 6 [U] = 13mg/ml Residence time - 15 minutes. TABLE 2 Data on Pu elution [Pu],mg/ml % Pu £Pu % Fore cut 0,. 103 0 .238 0..238 I bed volume 9..80 37 .964 38. 202 II bed volume 13.,66 52 .932 91. 134 III bed volume 1. 71 6 .622 97. 756 IV bed volume 0. 33 1.293 99. 049 V bed volume 0. 14 0 .537 99. 590 Residence tine - 60 minutea. SST - 05.2 Recovery of plutonium from phosphate containing aqueous analytical waste solutions using macroporotis anion exchange resin I.C.Pius, M.M.Charyulu and C. K. Sivaramakrishnan Fuel Chemistry Division Bhabha Atomic Research Centre Bombay, INDIA 400 085. SUMMARY Distribution ratio of Pu(IV) between 7.5 M HNO^ + 0.75 H?P04 + 0.3 M H->S04 medium and macroporous anion exchange rpsin Amberlyst A-$6 enhanced from 40 tc 250 when 1 M aluminium nitrate was present in the aqueous medium. When 1 M -ferric nitrate w^s used in place of aluminium nitrates the distribution ratio furthe: increased to 850 indicating the feasibility of recovery of plutonium from aqueous analytical waste solutions containing phosphoric acid in presence of these trivalent cations. (Key words: Plutonium, Macroporous anion exchange, Phosphate medium ) I. INTRODUCTION Plutonium bearing nuclear fuels contaning uranium are routinely analysed for uranium content by Davies and Bray method . Recovery of plutonium from the aqueous waste generated from the above method cannot be achieved by conventional methods like IBP extraction or anion exchange directly because of the presence cf phosporic acid. ^owever the presence of large excess of trivalent rations like Al*"" or Fe~ suppresses the effect of phosphate ions by complex ing, the;reby enabling uptake of Pu It has been observed that triv^lent cations like Al3+/Fs3+ enhances the extraction of Pu(IV> by Aliquat-336 from aqueous oxalate medium *-. A similar enhancement of distribution ratios of Pu(IV) between macroporous anion exchange resin Amberlyst A-76 and phosphate containing aqueous medium was also observed in presence of these trivalent cations. Distribution ratios of Pu(IV) between Amberlyst A-26 and aqueous medium containing 7.5 M HN^ +^+0.75 H3P04 + 0.3 M H2SO4 varying concentrations of Al~"/FeJ ar •' listed in table 2.'"" The high distribution ratios of 250 in presence of 1 M aluminium nitrate and 850 in presence of 1 M ferric nitrate indicate the feasibility of recovery of Plutonium from the phosphate containing aqueous analytical wastes. Column operations with synthetic mixtures for optimising the parameters are in progress. IV. REFERENCES 1. W.Davies , U'.Bray, Talanta 11, 1203 (1964) 2. I.C.Pius, M.M.Charyulu and C.K.Sivaramakrishnan " Studies on the extraction, of Pu(IV) from aqueous nitric acid-oxai ic acid mixtures by AIiquat-336 in presence of trivalent cations AlJ ,Fe^ . Paper being presented in this symposium. Table 1 Effect o* CH-jPO43 on distribution ratio of Pu(IV> Aq.phase : 7.5 M HN03 + varying CH3PO43 Ion Exchanger: Amberlyst A-26 CH3PO43 M 0 0.25 0.5 1.0 1.3 2,0 D 50GI0 430 100 20 8 5 Table 2 3* 3+ Oistributilon ratio of Pu(IV) in presence of M or Fa' Aq. phases 7.5 11 HN,OX + 0.75 H3P04 + 0. 3 M H2SD4 + varying concentrations of A1J /Fe° CA13+/F e 3 M 0 0. 1 0 .25 0.5 0. 75 1.0 D I.) (Fe3+) 40 20 30 125 450 850 SST - 06.2 STUDIES ON THE DEGRADATION OF ANION EXCHANGER EMPLOYED FOR PLUTONIUM PURIFICATION P.S.Dhami, V.Gopalakrishnan, A.RamanuJam, R.K.Dhumwad Fuel Reprocessing Division, & M.Sundaresan Analytical Chemistry Division, Bhabha Atomic Research Centre, Trombay, Bombay- 400 085, India SUMMARY Strongly basic quarternary ammonium type anion exchangers employed for the purification of Plutonium undergo thermal, radiolytic and chemical degradations leading to the delinking of quarternary ammonium type groups which precipitate plutonium at higher nitric acid concentrations. The paper describes an attempt made to characterise this type of precipitate. ( Key Words : PLUTONIUM, PUREX PROCESS, ANION EXCHANGE ) I.INTRODUCTION - Among: the methods available for large scale Plutonium purification, anion exchange method /I/ offers several advantages. Plutonium is absorbed on a strongly basic quarternary ammonium type anion exchanger mainly as Pu(NO3)6-- at 7.2 M HNO3, and after washing. it is eluted with 0.5 M HNO3. Elevated temperature is often employed to improve the kinetics. During repeated and prolonged ion exchange cycle, the resin undergoes thermal, radiolytic and chemical degradations resulting in chemical and physical damages that impair the process performance/2/. Besides, significant „ quantities of white precipitates with fairly high concentrations of plutonium are observed in the eluted plutonium streams. An attempt has been made to investigate these precipitates. II.EXPERIMENTAL - Various anion exchangers including Dowex- 1x4 in nitrsite form were heated with 8 M HNO3 at temperatures 60°C, 70"C and 80" C for 50 hours. After every 5 hours' heating and cooling, samples were drawn and the extent of degradation was assessed by the ability of a sample aliquot to reduce the intensity of thorium-thoron colour complex. Dowex-lX4 resin extract after 50 Hrs'degradation was tested with mono, di, tri and tetravalent metal ions including thorium and Plutonium for precipitate formation. Thorium solutions were added to bench reagent, tetra propyl ammonium hydroxide in 8 M HNO3 ( TPAN ) and the leach liquor of the rsoin Dowex-lX4 (resin degradation product ,RDP) and the precipitates formed,Th-TPAN and Th-RDP were filtered, mildly washed with water (precipitate's are water soluble ) followed by ethanol and dried in air. They were subjected to various analyses viz. elemental analysis, IR spectra measurements and thermal analysis coupled with mass spectroraetry. II[.RESULTS AND DISCUSSION - The extent of thermal degradation suf t'ered by anion exchangers tested was found to be in the order Dowex-ix4(commercial)> Amberlyat A-26> Tulsion A-35> Tulsion A~2? SST - 07.1 > Tulsion A-23 and therefore Tulsion A-23 appears to have better stability than Dowex-lx4, but its exchange capacity is low. The leach extracts of the resins formed precipitates with tetravalent cations like Th(IV) and Pu(IV) but not with raono.dl or trivalent metal ions. These precipitates are insoluble in 8 M HNO3 at room temperatures but soluble at 65-70* C and reappearing on cooling. They are also soluble in water/dilute acids and TBP.These characteristics are comparable with ThTPAN and with the white precipitate observed in Plutonium elution stream. It is felt that in the case of strongly basic quarternary ammonium type anlon exchange resins the functional groups get delinked from the resin matrix due to resin degradation and form precipitate with tetravalent cations at higher acidity. The percentage constituents of Th-TPAN and Th-RDP are presented in Table 1. Using these data the emperical formula for both the compounds were worked out. For Th-TPAN, the percentage of constituents observed are in reasonable agreement with the theoretical values. Hence the % constituents estimated for Th-RDP using the same method can be considered to be reasonably accurate. Nitrate estimations have shown that the ratio of Th to NO3- in Th-TPAN is 6 while that in Th-RDP is 5. Eased on thermal decomposition studies and analytical data obtained for Th-RDP, the compound has been assigned the following structure : C3H7 Th(NO3)5 l^ ^ .3H2O As such. not withstanding the evidence observed. other possibilities are not ruled out. The evidence, cited above, however, is sufficient for a tentative start IV.REFERENCES 1. J.L.Ryan, E.J.Wheelright, HW - 55893, Hanford Atomic Products Operation. Richland, Washington (1959). 2. E.V.Egorov, P.D.Novikov, Action of Ionising Radiation on Ion Exchange Materials, Transl.J.Schomorak, Chapter IV, Israel Programme for Scientific Translations, Jerusalem (1967). Table 1. Composition of Th-TPAN and Th~RDP Th-TPAN Th-RDP Costituenta Observed Theoretical Observed values (%) values (%) values (%) Th 23.5 23.8 32.7 C 29.0 29.5 14.0 H 5.5 5.7 2.7 N (org) 8.0 8 6 9.3 N (total) 11.9 11.5 11.1 SST - 07.2 (STUDIES ON POM SUING OF DEGRADED PUREX SOLVENT KOR PARTIAL RECYCLE B.K,SiPKb.M-K.T.Mair.D.D.Bajpai,N.Varadarajan, P. B.Gurbti, K.K.Gupta and Rajendra Kumar PREFRE Plant, B.A.R.C. Tarapur SUMMARY Use of weak carboxyltc acids has been suggested for the splitting of degraded PUREX solvent by the mechanism of adduct formation. The diluent separates out as a new phase leaving degraded Tributyl phosphate (TBP) and fission products in TBF':Acid phase. The diluent recovered is found to be quite suitable for the purpose of recycling in the plant . Keywords: PUREX, Spent solvent I.INTRODUCTION PUREX process widely used in nuclear fuel reprocessing employs 30 percent TBP in normal dodecane (nDD) as the solvent. High radiation level, presence of acid and temperature in the process cause the degradation of TBP as well as nDD. TBP degradation products can be effectively removed by the primary clean up method of alkaline wash. Secondary clean up, using inorganic adsorbents/1/, is not very effective and this results in the accumulation of diluent degradation products in spent solvent which gives rise to increased formation of complexants, crud formers arjd emulsifiers. As a result, the rejection of degraded solvent becomes unavoidable. Use of phosphoric acid has been reported/2/ for the splitting of solvent and recovery of diluent. The viscous nature of phosphoric acid and need for thorough washing of the diluent to make it totally free from phosphate are two major drawbacks with this method. In the search for suitable alternatives for splitting the solvent, studies have been carried out in our laboratory using acetic, formic and lactic acid strippants. PRINCIPLE: Weak carboxylic acids form adducts with TBP which are Me33 soluble in diluent and hence get separated from the diluent forming another phase as follows; TBP nDD *• Acid = -•- TBP: Acid + nDl) II.EXPERIMENTAL Initial studies were carried out using acetic, formic and lactic acid on fresh solvent "to determine the required Organic/Aqueous phase ratios. For further studies, 3pent solvent washed with sodium carbonate was used. Operation was carried out batchwise. After each equilibration contact, lasting for about b minutes, / top riDD phase, lean in TBP is separated and subjected tu another contact with fresh acid. At the endjthe acid. TBP:Acid phase and diluent collected are pooled and kept separately. SST - 08.1 Tl»« diluent- was given a separate treatment before assessing its quality and performance . Volumes of n.DD phase, acid, and TBP phase were measured and distribution of gamma activity was also monitored to follow-retention of activity. Nitric acid equilibration method was used for the estimation of percentage of TBP in nDD phase. Plutonium loaded & retained in diluent, even after several washings, was determined as moles per billion litres and was expressed as quality of diluent in terms of Plutonium retention number. III.RESULTS & DISCDSSION Among the three acids, acetic, formic and lactic acid, the first two acids are found to have better stripping efficiencies associated with quick phase separation. To avoid the reaction between diluent. and formic acid, use of only 70 -80% acid and dilution of TBP:Acid phase without delay on stripping is found to be necessary. 60 70 percent lactic acid is also found to be effective and the possibility of recycling exists. All the three acids show better stripping efficiency with spent solvent as compared to fresh solvent. In two contacts, decontamination and improvement factors with respect to gamma activity & piuIonium retention number (feed to separated diluent). as required, were obtained. The separated diluent in each case was washed, made up to 30% in TOP and subjected to extraction / stripping studies. Performance is found to be satisfactory. IV„ACKNOWLEDGEMENT Authors are indebted to Shri A.NPrasad. Director. Reprocessing & Nuclear Waste Management group, for his support and encouragement during the course of this work. V.REFERENCES 1. J.C.Mailen, Secondary clean up of Idaho chemical processing plant solvent, Sepn. sci.8c technology, 22(2&3), PP-335.(1967) 2. H.E.Scrich et al. studies on the treatment of organic wastes. Part V. The Eurowatt process, Rep.ETR-287 (1980). Table 1 Studies on diluent recovery from spent solvent using different splitting agents 1 Acid , 0/A i TOP in ! Gamma Recycle Pu.Reten. I strength Ratio nDD % DF Volume % Improve.F. ! Acetic 90 % 4 4 45 70 150 lAcid [Formic JAcid ! 80 % 2 5 50 70 150 •i Lactic ; 70 % 2 , 10 17 70 24 ,Acid SST - 08.2 CARRIER-MEDIATED TRANSPORT OF TRACE QUANTITY OF PLUTONIUM ACROSS DICYCLOHEXANO-18-CROWN-6 /TOLUENE LIQUID MEMBRANE FROM LOW ALPHA WASTE SOLUTIONS Anil Kusar .R.K.Singh, M.K.T.Nair and J.P.Shukla* Power Reactor Fuel Reprocessing Plant Bhabha Atoaic Research Centre, Tarapur-401502 SUMMARY Macrocycle-facilitated transport of Pu(IV) from plutonium nitrate solutions of extremely low concentration levels across bulk liquid membrane (BLM) and supported liquid membrane (SLM) was investigated. Most of the Plutonium could be transported by employing 0.2M DC18C6/ toluene as carrier and sodium carbonate as strippant. 'Enka' Accurel flat sheet polypropylene membrane was used for SLM. (Keywords : Membrane.transport, plutonium. inacrocycle. carrier ) I. INTRODUCTION ^ In view of the ALARA principle (as low as reasonably achievable) currently followed in nuclear installations, the possibility of further removal of trace quantities of plutonium from low alpha wastes is being vigorously pursued. A comprehensive work program was initiated by us to study the feasibility of removing truces of Plutonium using macrocyclic carrier, namely DC18C6 in toluene medium .Effects of important parameters like feed acidity, carrier (DC18C6) concentration . nature and type of strippant were evaluated. 'Enka1 Accurel polypropylene (PP) films were tested as the flat-sheet solid supports for supported liquid membrane (SLM). II. EXPERIMENTAL Dicyclohexano-18-Crown-6 (DCI8C6) obtained from Aldrich Chemicals, U.S.A. was used as the carrier. Valency of plutonium in the feed was adjusted to tetravalent state by the addition of solid sodium nitrite /2/. BLM cells (Shulman Bridge type) consisted of source rphaseC 10 ml ) and recovery phasef 2.4 ml) separated by organic DC18C6/toluene and SLM, simple two compartment permeation cell consisted of a feed so.lution(2.5 ml) and product solution chamber(2.5 ml) separated by a flat-sheet thin membrane with, effective area of 1.13 cm2 were used in the experiments/3/. During the BLM experiments, a ratio of 4 : 1 between the volumes of source phase and strippant was maintained Membrane permeabilities at various stages of transport experiments were determined by monitoring the plutonium concentration in the receiving phase . III. RESULTS AND DISCUSSION Use of neutral crown ethers as carrier for carrier -mediated transport of metal ions across the organic liquid membranes is now well established. As plutonium(IV) is highly extractable by •Radiochemistry Division ** designated for final disposal SST - 09.1 DC18C6/toluene. from nitric acid solution/1/,it permeated easily across the xiquid membranes . With increasing nitric acid molarity. the plutonium permeation first increased passing through- a maximum at nearly 3M HNO3, and then decreased with increase in nitric acid concentrations . Several aqueous strippants such as sodium carbonate, oxalic acid, hydroxylamine hydrochloride were tested. The Plutonium recovery was over 90% and maximum flux exceeded 15.7 x 10 mol/m /sec after about 1 hr. of transport process took place from relatively lower feed acidity up to 3M HNO3 with 0.2M DC18C6 . With increasing carrier concentration in diluent like toluene, plutonium transport gradually increased reaching a maximum with about 0.3M DO18C6 and thereafter a significant decrease in plutonium transport. Typical results obtained for single-ion transport of plutonium across BLM using DC18C6 /toluene and sodium carbonate as the strippant are summarized in Table 1 . IV.ACKNOWLEDGEMENT Authors are thankful to Shri A.N.Prasad, Director, Reprocessing & Nuclear Waste Management group, for his support and encouragement during the course of this work. IV. REFERENCES 1. J.P. Shukla. R.K. Singh and Anil Kumar. Radiochimica Acta. (in press). 2. D.E. Ryan and A.W. Wheelwright, USAEC Report No. HW-55983 (1959). 3. J.P. Shukla and S.K.Mishra, Uranyl ion transport across tri-n-butyl phosphate / n-dodecane liquid membranes. Nat. Symp. on Uranium Tech.. Bombay. Dec 13-15 (1989). Table 1 Initial Feed Concentration : 1.0 mg /L plutonium in HNO3 Carrier concentration : 0.2M DC18C6 / toluene Strippant : 0.5M sodium carbonate Volume ratio of feed to strippant : 4 : 1 Source phase acidity . HNO3 Molarity. M 3 hr 1.9 4.5 14.6 8.9 5.8 2.7 IRate of !transfere 6 hr 1.7 3.0 13.7 5.0 6.2 2.9 '.Plutonium 3 hr 11.7 24.6 74.9 33.2 27.2 14.2 !permeation ! 6 hr 20.8 ! 31.6 97.0 52.2 45.3 30.1 V _ -9 2 a. x 10 mol/m /a SST - 09.2 A MODIFIED METHOD FOR THE SYNTHESIS OF POLYCONDENSATE PHENOLIC RESIN WITH IMINODIACETICACID (IDA) FUNCTIONAL GROUP FOR TREATMENT OF ALKALINE WASTE FROM REPROCESSING PLANTS N. V ar adar ajan , J . 6abr i e J , Pintut Sen and M.K. t.Nair PRFFRE Plant ,BARC ,Tara(3ur . ABSTRACT bynihesi s of bakel i re type res5in i-n I h iraiiKidi -n:et ir aciij \ I l/H •• r|f (i'i(i wan i sr r i cil out" . 1 he product 'furmed by followjnq tht> procedure pnh) ishrri in 1 i t ivr ,»t ur e was inund to hp unstable in aikalim- Siolut.inii and t htr < eatt ion rate waft tonsi dw dbl v siuw. H modified method tor pr epar i nq the resin with improved stability is rliscjnssfiJ in thin paper . INTRODUCTION I 1 ttr* r esins tor inpd i r om i esut riiml anil lui niri I d«-.'h yi lie usinq bast? catalvst , is hiqhly selective tor cesium i.iptdke from all'alirtR ?>n.lul iod. A more ver sal i Je i PBJU product is reported in literature (I,.:! ) which si mu I, f-..aiie(ji is J v removes resinm and atront i 1.1 in _ It can tie pr t'pdi t - 1 Bynt liiih.i B: I bA eincJ i tstirunol wati ili &BO1 v«il in wr. II. Wash put stability Jin acid and alkali: Sit rsl.»i 1 i t y in acni and ial kal i was determined by weiqhr loss in ,,'li ftirmic acid ctniJ 6 h Sodium hydro;; l"de» solution respectively. The contact time was ^''1 hrs and we i qhi nqti were done atlter c.onver I i m.j t. ht.j rosin into H form. 6I'I <>ud:iii/ii hydro;: ide was used to obtain measur rfh I i; deqradatitm at: e.u.i el ar a !.ed attaik ttiuuqh this rt-sin wi J 1 be normally seeinq comparitJvelv lower alkali concentration* in the r anqe of 0.01 to 1.0 molar in the actual # W III Physi cal pr oper ties of various Resort inoi -1 DA pol yiner. Product R : IDA: 50 C i no C 1 inear Lr ossed Li near Irossed A 4 : 1 : 15 R\G RB\B RAG RB\S B •r : 1 : 12 R\l RB\S R\6 RB\S ("': : J : 9 R\l_ R\l.. R\L RH \ S D 1 : 1: 6 R\L R\l. R\l RB\B R = Red RB - Reddish Brown L •= Liquid G = Bel S = Brit hie solid RESULT AND DISCUSSION in the inter niedi ate steqe o ( polymerisation, IY>Pi is attached to the phenolic rinq through the methylo) qroi.ip and no fnrt.tier reaction -for this site is pass! hie and inhibit cross linkinq. The polymerisation reaction is very slow in the case of product. D. Only a liquid product is obtained after keepinq tor 24 hi's. In the case of Product A & B, a qel product is obtained i n . 1 •*•• fri hairs respectively. In all cases, a solid reddish brown product is formed when cured at 1OO " C for 24 hours. All products are reasonably stable in 2 molar formic acid. Product. D is of very poor stability and 30% by weight, is washed out whi)e Lontariinq with 6M alkali solution, the product. A hasvery low wash out effect and its distribution coefficient is comparable with other products < Kd for ..cesium is 2 3;; J 0 and for strontium is ol the order of lx 10"' ). So it is recommended to (Modify the procedure by chanqinq the mole ratio of react ants 1:4 with respect to i mi nodi aceti c acid and resorcmo) . ACKNOWLEDGMENT tiutiioi s are sincerely thankful to Shr i . A.N.I-'r asad. Direr tor , f-uel Reprocessing & Nuclear' Waste Management Group and bhri. M.S.Kumra, Head, Waste Manaqemeiit Division tor them guidance and support in carrying out this work. REFERENCES 1. P. l; .. Dduraqartcin , I'l.rt.Ehr a Waste Hcmaqemeiit. , I9B1 1'ur.Bon. febnidry s.S 26, 1V61. 2. R. H. Wai 1 ar.e, R.B. Ferguson in t er fiat i ona J Sv'niposii urn on Scientific. Hasis ol Muc 1 eai Waste Manacjement. ( boston, Masssac iietus, November o •-- 20 , 1980 >- SST - 10.2 FEASIBILITY STUDY OF VACUUM DISTILLATION AS A TECHNIQUE FOR DECONTAMINATION OF KEROSENE RADIOACTIVE WASTE. " RG Yeotikar, CP Kaushik and Kanwar Raj WIF, WMD, BARC, TARAPUR COMPLEX, PO. GHIVALI, DIST. THANE, 401 502. (M.S.). ABSTRACT : Radioactive kerosene waste is generated in various operations in bituminisation plant and active laboratory where analysis of waste product is carried out. Major quantity of waste is generated due to cleaning of various systems by kerosene during mechanical maintenance operations in bituminisation plant. This waste constitutes mainly kerosene with some bitumin in it. Vacuum distillation technique has been studied as one of the ways for decontamination and volume reduction of this type of waste. The distillate which is having traces of activity can be reused for cleaning operations. KEY WORDS : Radioactive kerosene waste, bitumin waste product, decontamination, vacuum distillation. INTRODUCTION : Bitumin has been adopted as a matrix for incorporation of intermediate level vaste of the specific activity in the range of 1 Ci/lit. Bituminisation of this waste is carried out by feeding the waste alongwith emulsified or straight run bitumen in an evaporator- assembly. In India, waste and emulsified bitumen are fed to a wipe film evaporator to get a bituminised waste product (BWf)[1J. During mechanical maintenance operations in the bituminisation plant, cleaning of various assemblies/systems is required. Kerosene has been used for cleaning purposes in this plant, thereby, generating radioactive kerosene waste. Secondly, during characterisation of radioactive BWP, kerosene is also used as a solvent for bitumin and fcx- general cleaning of bitumin contaminated material/article. This also produces radioactive kerosene waste. Various methods are available for management of liquid organic waste. These include incineration, fixation in either cement or polymer, degradation etc. [2]. These methods basically destroy the active organic liquid waste. Since kerosene waste generated is not highly radioactive, re-use of the kerosene has been thought. Vacuum distillation of this waste gives the distillate of negligible activity and, hence, can be re-used for cleaning purposes again. This paper describes the feasibility study of vacuum distillation process for decontamination of active kerosene waste. Parameters like distillation start-up temperature, maximum disti1lation temperature, decontamination factor (DF) and time required for maximum recovery of kerosene were measured against applied vacuum to the system. EXPERIMENTAL : Corning quick-fit glass apparatus was assembled for vacuum distillation study. A three-litre multineck flask was used as a distillation chamber. This flask was th TABLE : RESULTS OF VACUUM DISTILLATION OF ACTIVE KEROSENE WASTE. Sr • Applied Distillation Temp. •Time for Activity D.F. No. Vacuum Start up | Haximum recovery Waste [Distillate of 96% (mm of Hg) <°C) (CC) (mins) (juci/ml) (Appx) 1. 250 195 245 480 2.5x10^ <10~* 250 2. 350 175 235 450 2.1x10 <10 210 3. 400 168 230 360 2.8x10" <10":S 280 4. 450 160 225 300 4.5x10 <10~* 4 500 5. 500 150 210 240 4.5x10* <10~s 45160 6. 590 135 190 165 4.5x10 SST -11.2 ELECTRO-OXIDATIVE DECOMPOSITION OF ALPHA CONTAMINATED ION EXCHANGE RESIN ILJBLGurf>fl.R.K.Singh,M.K.T.Nair,A.K.Venugopal, D.D.Bajpai,R.R.Singh and S.P.Dhakras PREFRE Plant, B.A.H.C, Tarapur SUMMARY: Application of silver catalyzed electro-oxidative method for the decomposition of alpha contaminated ion exchange resin has been suggested. The electrolyte used is 4 molar nitric acid containing 0.01 to o.05 molar silver nitrate. Intermediate temperature of ~55- C is found to be adequate. Simplicity of process, effective volume reduction and complete control over the operation are the advantages of this method. Keywords : Electrolysis, Ion exchange resin . I.INTRODUCTION Ion exchange resins have wide applications in nuclear industry. They are used for water treatment .treatment of active liquid effluents and final purification of plutonium in spent fuel reprocessing. Disposal of used resiri, specially plutonium contaminated resins , is a tedious job involving many safety constraints. No simple method is available either for near total volume reduction or for complete leaching of activity from the resin before immobilization /I/ . Oxidative methods have the potential to overcome this situation . Hence the catalyzed electro-oxidative technique is considered* for the decomposition of resin . This is mainly because of our past experience /2/ in the application of this technique and also due to the encouraging results obtained in this process/3/. The purpose of initiating investigations on the decomposition of alpha contaminated ion exchange resin is to try and tackle the waste generation at the source itself . PRINCIPLE: At anode Ag?I) is oxidized to Ag(ll). In the anolyte Ag(II) forms oxidizing and reactive species which react with the organic resin and decomposes it. . 2Ag2+ + H2O > 2Ag» + 2H* + (0) Org. resiri + (0) > CO2 + CO + water + Inorg. products Reacted silver ions are continuously re-oxidized at anode. II.EXPERIMENTAL A small glass cell fitted with a frit to separate anoiyte and catholyte was used for eleetro-oxidative decomposition of resin. He3in was fed to anulyte chamber . Water bath was used to maintain the cell temp, at "55- C. Platinum mesh as anode and platinum wire as cathode were used along with magnetic st.irx'ing. In all the cases 4M nitric acid medium was used.Effect of change of current density and catalyst concentration on the rate of decomposition was studied. Dowex 1x4 resin usea for plutonium purification was employed for electro-decomposition studies. SST - 12.1 II[.RESULTS & DISCUSSION Electrolysis experiments clearly revealed that the time, corresponding to 30 -40 % of resin decomposition.was adequate for the . . complete removal of traces of trapped plutonium. A possibility of resin decomposition by anodic oxidation in absence of catalyst ,though at lesser rate, was observed Apparent rates of resin decomposition with 0.01M and 0.05M catalyst were nearly same. During the experimental work it was observed that higher catalyst concentration, higher anode current density and temperature of about 60- C help in achieving lowest concentration of dissolved organic. Studies regarding the nature of intermediate and end products is in progress. IV.ACKNOWLEDGEMENTS Authors are thankful to Shri.A.N.Prasad,Director,Fuel Reprocessing and Nuclear Waste Management Group for his guidance .support and encouragement given during the course of this work . V.REFERENCES 1.Treatment of spent ion exchange resins for storage and disposal Tech. Rep. Ser.No. 254, IAEA (1985). 2.R.K.Singh et al,Preferential PuO2 dissolution using silver catalyzed electrolyser, CT-3,Radiation & Radiochem.Symp. (1988 ) . 3.D.F.Steele, A novel approach to organic waste disposal, Atom, 393, (1983). Table-1 Results of electrolytic elution of Pu trapped in used resin No catalyst Time % of resin 4 M HNO3 mints. 30 60 90 120 150 180 decomposed 5 mA/cm2 % Eluted 50 67 81 86 93 94 33 0.01 M Ag Time 15 30 45 60 75 90 4 M HNO3 mints. 5 mA/cm2 % Eluted 47 66 80 84 89 97 45 Table-2 Results of electro-oxidative decomposition of used resin Resin Ag cone. Anode CD. Rate of decomposition , M mA/cm2 g. of resin/Amp.hr. Nil '0. 10 Dowex 1x4 0.01 5 0.32-0.36 powder 0.025 5 0.33-0.35 0.05 5 0.33-0.35 4 M HNO3 0.05 10 0.29-0.33 0.05 25 0.20-0.21 0.01 5 0.29-0.31 Dowex 1x4' 0.05 5 0.31 beads 0.05 10 0.26-0.27 4 M HNO3 0.05 20 0.14 (35- C) SST - 12.d ELECTROLYTIC DESTRUCTION OF NITRIC ACID IN REPROCESSING STREAM A.Palamalai, S.K.Rajan, M.Sampsth, A.Chinnusamy, F.Govindan, S.V.Mohan, V.R. Raman and G. R. Balasubr*m=uiiari. Reprocessing Programme Indira Gandhi Centre for Atomic Research Kalpakkam 603 102, INDIA. SUtfilAEY The acidity of the first cycle raffinate of F-Tst Reactor Fuel Reprocessing ha.v. to be reduced from 4M to 0.1M in order to effectively reduce i?s volume by evaporation to an acidity of a 6M.This could be achieved electrolytically in the cathode nl. Current ef 3 'ecioncies ranged from 42 to 03%. Key Words : Fat>t Reactor Fuel Reprocessing /Kaf f inate/lioduction in the acidity/Reduction of waste volume. IHlfiQEUClIOJi : The first cycle raffinate of the reprocessing plant, wherein Fast Reactor Fuel containing 20% or more Plutonium is reprocessed contains 4M Nitric Acid.This waste stream has to be atos.-ed in SS tanks until it is immobalised in glass matrix. The upper limit of acidity that can be stored in GS tanks is 6M nitric acid, since acidity greater than 6M wilJ cause sewert; corrosion problem.Thus the raffinate waste can by concentrated 50% by volume only through evaporation.If the acidity is brought down to 0.1M,then the volume EXPERIMENTAL •' 4M nitric acid, containing the various fission and <:;orro.-;i on products in the inactive form him bt-eri prepared. One litre *_-f this electrolyte is taken in the catliodio compartment of a twr, litre glass cell. One end closed porcelain tube of 35mm ID functions cis the anod'j •-••>mpirtin<;nt . Tlie iinoiyte ia 4M nitric acid. Ti t.iuiium expanded metal wade into throe turns is used as the cathode. The pt-platt-d titanium expanded metal is used as the anode. In those experiments, where diaphragm is not u.seJ Mixed Oxide Coated Titanium Anode (MOCTA) developed in MDL.has been used us the anode. From a rectifier,for tissing 10 amps current,6 Volts hs'/'j to be iipi'liod wlu.:u no diaphragm is ut.-.-d and 12 to 14 volts have to applied for passing 10 amps current when the diaphragm is employed. SST - 13.1 AKD EISJ2USS1QH : 'Hi* acdity has been reduced from 4.02M to 0.096M. At different intervals the samples are taken and acidity Lr> determined by titration against standard sodium hydroxide solution. Initially we hiive carried out electrolysis experiments without diyphragm and the acidity is reduced from 4.02M to 3.12M. Here the current efficiency ir, low, around 10%. On intrdu.:tion of the diaphragm the acidity :i s reduced from 3.12M to 0.096M. Then the current efficiency ranges from 42 to L3"i. The current ef f ici- nci es are rjhown in the table. With the introduction ot' the diaphragm the temperature h»s been found to L-aia>i fr:>m 30 to f0°C,due U> LR hent in£. That, in why the voltage ranged from 12 to 14. Further otudies to sr-ale up thio acid killing method are in progress. Table - Current ef f ioi «'.-n<'?y for the e 1 eel ro Ly Lie destruction of nitric acid. Expt.No. Redu-j ti.-n of Current efficiency /% Hemarks Acidity / M From To 1. 4.02 3.49 10.2 Without Diaphragm 11 . 3 2. 'J.49 3.12 53 .6 3. 3.12 2.64 53.6 With 48 .6 Diaphragm 4. 2.64 2.06 .7 b. 2.06 0.86 42 .6 6. 0.86 0.10 SST - 13.2 ELECTROOXIDATIVE DESTRUCTION OF ORGANIC WASTES GENERATED IN A REPROCESSING PLANT A.Palamalai, S.V. Mohan, P.Govindan, A. Ohinnunamy, M. Jiampath, - •-••-•• •-- - - V.R.Roman and G . R. Bcilasubramanian . Reprocess ing Programme Indira Gandhi Centre for Atomic Research Kalpskkrim - 603 102, INDIA. SUMMARY The organic wastes generated in a reprocessing plant such as gloves, tissue paper, and 30% TBP in n dodecane ar=> destroyed cleci.rox.idjt ively. The anolyte used is 1 MM HNO3 0. 1M AgNO3 .Silver is oxiidised to Ag(II) which in turn uxidises and destroys the org.'inics . Thus silver ion fnn':ti..>ns as I. Iv; redox c ital yet. The J-^IJII i rement :J o C #:l'.v -t- r.t c i ly tY>i des true I i • >n of 1 g or 1 nil of the organic wa..-:>te ary presented in this paper. Key Wi 1(1:3 : Electrolytic dc:;truet i' >r> / Orgariic wastes / R>.-d^^x IHlfiQDIIClIQM : Organic: WCIL;(.-S s!i.;h -,r. t. i.b.;u^ popei. •, :,u t g i«;al ar;d i'Ostmi.-i'tent gloves; tind :5i'ent, '.Wt '[[',[' in n d< •d._-'. ane aro presently treated by incineration me thod. T'I i :.; i u a liU'.h t..-inpr'iv.itu.r-»: proo-.^fs to be performed for the Pl.ut<.nium b>r;iring wasttiS.lio this incineration has to be carried out in a glove bex or jn a containment box. If the wastes are treated e) ectroi yt-i >: a] Jy , the operating 'temperature will be leas than 1 'l.kf <". Wln-n l.lie organic wastes are destroyed by eolectru oxi do t i •.•ri, tin* wastes are converted into carboiidioxide c\n<3 wat-.rr.Tlii Plut.»nium present in the wa^iUa will be in the nitric acid .S'.lulJon v/iiicli can be readily recovered by ion exchange method. •" Th« anolyte erjipjv.yed f.,r the elects o-sidative de^tru..:ti.>n of the organic waal.es is 13M HN< >3 H.1M AgMOaand the cath-'lyte if; 13M i;N<")3 . Tlie two con.part, merit:, are .separated by a porcelain diaphro/irn. A platinum gauz-j of !.• -.in diani'.t.er rmd Y c.-n height is used av. th.j anode and a titanium coj 1 :•? 2 cm diiitneter and 7 cm height made from titanium wire of 1 mm thickne-.i.. in vi.sed as the cathode, me aii'.Jytc i..-:'. li.^ated !o Yf*r.A tliefiiio(-;yph-jn type cell of ] litre capacity made from glasij is t..:ied as the electrolytic coll-Ft^r passing 2. !J amp. current ;it room t'jiuperature (without heating ) i'i vi... 1 t..-j &j>; rt.Mju i red wlitrca:-; for passing the same current with heating '<:<• 7f." C.G voLta are needed.Excepting 30% TBP in n dodecane, -d)out 1 g each of the solid organic waste:.; is introduced in the .n... ly te. When jhu'it one gram of neoprene gauntlet is de&t.rt.yed, u whi *.<•-• prcttipi t/itu of silver chloride is found to sett]<: at the bottom of the c-.sll.The supernalent liquid still irurn::tions an the i .^lyt.e lor destruction of organic wastes,. SST - U.1 RESULTS AND DISCUSSIONS : Initially,the electro oxidative destruction has been carried out at room temperature((«i*-->inut ho a ting). The requirement of electricity for destruction of 1 g P'urthyr work to s<:ale up this method is in progress. Table : Requirement of electricity for destruction of organic wastes S.No. Type of organic No.of Faraday required for lg or lml /F waste. Without heating With heating I. Surgical gloves 1.806 0.2027 2. Postmortem 2.491 0.0411 gloves 3. Tissue pape • 1.621 0.0797 4. 15ml of 30% TBP 3. 704 0.0375 in n-dodecane 5. 50ml of 30% TBP - fc.0130 in n-dodecane 6. 100ml of 30% TBP -• 0.0174 in n-dodecane 7. 200ml of 30% T3P 0.0251 in n-dodecane SST - U.2 EVALUATION OF STABILITY OF EPOXY MATRICES FOR WASTE IMMOBILISATION USINB l37 Cs AS A TRACER. S.V.S RAO, K.B. LAL and R.V.Amalraj Centralized Waste Management Facility Fuel Reprocessing & Nuclear Waste Management Group, 8. A. ft.C . , Kal pakkam-603102. SUMMARY Ihi adaptability of epoxy resins for the solidification of sodiua sulphate frot ion-exchange regenerants has been exaiined. The epoxy resins prepared with 48 »/»I of todiu* sulphate and II7Cs tracer. The polyaer M trices nere subjected to leaching and cotpressive strength tests. Froi the leaching studies the diffusion coefficients Mere calculated and the results Mere coaparid Kith couerciil ifocy resins and PQIVETHVLENE tttnees. Key word* i Epcxy- Resins, Diffusion coefficients, immobilisation, Leach Fractions. I. INTRODUCTION Sodium sulphate *long with other radio nuclides like cttium, strontium is generated from the regeneration of ion- exchange re«in>. Since, cement and bitumen »rm not suitable media for incorporation of sodium sulphate due to swelling and cracking problems, a laboratory study was conducted on the preparation and characterisation of epoxy matrices incorporating oodium sulphate,, The diffusion coefficients were calculated using semi infinite plane source model ac recommended by IAEA (International Atomic Energy Agency). Radioactive cesium is chosen as a tr^cmr due to its high diffusivity from the matrix. The results »ra compared with a commercial epoxy resin and also with polyethylene. II. EXPERIMENTAL The epoxy resins were prepared with selected monomer ratio of bis-phenol A, epichlorohydrin (lil.4> and with different amounts of hardeners namely benzyl dimethyl amine (BDMA) and tris phenol E2# 4, 6-tri-3 (dimethyl aminomethyl > phenol] to optimise the composition. The polymer blocks of epoxy resins of 4.5 cm dia and 4.4 cm height were prepared by adding 60 7. w/w of sodium sulphate along with the hardener which was spiked with 40 vCi- °* '"Cu. Similar blocks were also prepared using commercial epoxy resin ie araldite. The polymer blocks were subjected to compress!ve strength test and leach resistance test which are important parameters in the evaluation of wa&te forms. i. CompreBsive Strength Measurementi The compress!ve strength measurement was. carried out on the standard blocks with thiar test machine which complies with the grade*A'of BSi1610119&4. ii. Leaching Testi The leach tests were c*rri~.~ ?ut a« per the procedure given by the IAEA1 on the standard blocks by immersing them in 35GI ml water and the cesium was «stimated using single channel gamma spectrometer. SST - 15.1 III. RESULTS AND DISCUSSION The compressive strength and cumulative leach fraction* of the matrices i. Epaxy matrix (BDMA as a hardener) ii. Epoxy matrix where V & S are the volume and surface area of the sample reepertively. The equation <1) is derived from the cylindrical plane source modela assuming the surface at the leachant/solid interface *t t > 0. The diffusion coefficient3 in considered to be an useful parameter in predicting the long term release rates of activity from th» matrix. The calculated diffusion coefficients were found to be not in the same order as that of compressive strengths. This may be attributed to the structural differences in tha matrices due to the nature of hardeners employed. The compreatii ve strengths and diffusion coefficients of epoxy matrix (BDMA) are comparable to polyethylene. In contrast the diffusion coefficients for commercial resin (araldite) matrix were found to be two orders higher. IV. REFERENCES* U) R.A.J. Satbell, Nuclear 1 Che»ical Waste Management Vol.3,pp 125 - 129, 1962. 121 riasabuii NishikaiM, Hurt ear & Cheaical Haste Hanageient Vol.5,pp 101 - 111, 1984. 13) S.V.S.Rao, K.B.Lai,R.V. A«lrai,Si*ulated studies on the release of"Sr and 1>7Cs into environaent fret Cecent I Mobilised beclaiiinq wastes. Proceedings an ftadio-cheiistry and Radiation cheaistry.Kalpakkai,1989. 14) Nobaru Horiyua,Nuclear It Cheaical t'aste Nanagevent ValJ.pp 23-2S, 1982. TABLE - I Matrix Coipre^bive Cutulative Diffusion Strength IHPa) leach Coefficients fraction Ici2/day) I. Epony IBOHA) 38 a.21 1.92E-7 Jl. Epoxy 47 8.53 2.72E-5 (Iris phenol) ill. Conner c ul 619 8.32 1.55E-5 (Aralditel IV. Polyethylene4 25 B.4 I.BE-5 to IE-A SST - 15.2 THE EFFECT OF DILUENT ON EXTRACTION OF URANYL NITRATE BY TRI-N-BUTYL PHOSPHATE , T.&.Srinivasan and P.R.Vasudeva Radiochemistry Programme, IGCAR, Kalpakkam 603 102, India. INTRODUCTION: The effect of diluent on solvent extraction of met^l ions, in general, has been an extensively studied subject. The effect of uiluent on the distribution coefficient (K.) for the extraction of Th(IV), U(Vj) and Pu(IV) from nitric acid media into Tri-n-Butylphosphate(TBP) was studied in the early sixties and seventies [1]. There has been, however, no agreement on the nature and properties of the diluent that could explain the variation observed in K. values with different diluents [1]. In the case of paraffin diluents, there has been no systematic study of the effect of the carbon chain length or branching. While the reason for this could be the very small magnitude of the changes expected in K. with cringes in paraffin structure and chain ^ngth, it is also true that due importance has not been given to choice of the experimental conditions such as aqueous phase composition and TBP concentration. The use of macro levels of TBP, and high concentrations of nitric acid or the metal salt result in the coexistance of several species (free TBP, TBP.HNO^ and M(N0J 2TBP) in significant concentrations, and it is < lear that the differences observed in K, values cannot be unequivocally attributed to the properties of the diluent alone. With this in mind, we have carried out studies on the extraction of U(VI) by TBP in various diluents, mainly n- paraffins, employing low concentrations of TBP, HNO^ and IJ(VI), and the results are presented in this paper. EXPERIMENTAL: .The diluents used were of purity 99% or more. TBP-obtained from FLUKA (97% pure) was purified by the procedure suggested by Alcock [2]. Uranium-233 was used as tracer for ;.he distribution coefficient measurements. 5% TBP solutions of diluents under study were made and used for extraction of uranium from an aqueous phase of composition 1.9 M NaNO., - 0.1 M HNOV The equilibration of aqueous and organic phases (2ml each) was carried out at a constant temperature of 303 K, Suitable aliquots were taken from the phases, after equilibration and uranium was determined by liquid scintillation counting. RESULTS AND DISCUSSION: Fig 1 presents the data on K . values obtained with various diluents. The K. value is seen to decrease initially, with increase in carbon number of the diluent. For carbon numbers higher than 10, the change in K, value is very small. It is also seen that the K. value obtained with a branched diluent is higher than thet obtained with an n-paraffin of same total carbon number. The paraffin diluents studied in OT work are not expected to have any specific interactions with TBP or metal solvate. Therefore, the extent of extraction is likely to be mainly influenced by the work done to disrupt the diluent structure in transporting the solvate to the organic phase. Based on molecular geometry, it can be expected that more work should be done for disturbing the structure in n-Paraffins than in branched paraffins, since, in the latter case, steric effects prevent close packing of the diluent molecules, thus making "hole formation" easy. The physical parameter that relates to the cohesiveness of the diluent molecules is the solubility parameter (d) [3J. Thus, it could be expected that K. values can be correlated SST - 16.1 with the d values. The correlation seen in Fig 1 confirms this hypothesis. The d values used in this correlation were calculated using the empirical calculational procedure suggested by Majer et al [4]. The values for straight chain paraffins are in excellent agreement with the reported values [5], confirming the validity of the empirical calculations. The d values for the branched paraffins used are being reported for the first time. It thus becomes possible to understand the reason for the trend in K , values, and especially the higher K, values obtained with branched diluents a§ compared to the value for n-paraff"ns of same carbon number. The higher K . value obtained with benzene and toluene perhaps indicate the presence or additional specific interactions in the organic phase between the TBP solvate and the diluent. REFERENCES: 1."Science and Technology of Tributylphosphate", Vol.1, Eds. W.W.Schulz and J.D.Navratil, CRC Press,inc., Boca Raton, Florida (1984). 2. K.Alcock, S.S.Grimley, T.V.Healy, J.Kennedy and H.A.C.Mckay, Trans. Faraday Soc, 52, 39 (1956). 3."Solvents and Solvent Effects in Organic Chemistry", Christian Reichardt VCH (1984). 4. "Meats of Vaporisation of Fluids", V.Majer, V.Svoboda and J.Pick, Elsevier (1989) p. 148 5, E.P.Horwitz, K.A.Martin and H.Diamond, Kaplan, Sol.Ext. Ion Exchange, 6,859(1988). 1 1 1 DILUENT':", USED : 1:HtXANE; 2:HEPTANE; 3:OCTANE; 4: NON/W:"; 5:DECANE; 6-UNDL"ANE; 7.-DODECANE; 8:TETRftDECANE; 9:HEXADECANE; 1O:2,3-DIMETHYL BUTANE; 11 :2,3-DIMETHYLPENTANE; 12:2,2-DIMETHYL BUTANE; 13 : ISOOCTANE; 14:BENZENE; 15:TOLUENE •a 10 J 25- 3 o o 4 IS J K J i 15 16 17 18 19 SOLUBILITY PARAMETER (8) FIG-1 SST - 16.2 RECOVERY OF URANIUM USING BIFUNCTIONAL RESINS K.N.Sabharwal, P.R.Vasudeva Rao, and M.Srinivasan Radiochemistry Programme, IGCAR, Kalpakkam 603 102 * DepU of Chemistry, IIT, Madras 600 036 INTRODUCTION In reprocessing and recovery operations, waste solutions are often gener- ated which contain low concentrations of uranium or other actinide elements in the presence of various impurities. Recovery of the actinide elements from such waste solutions has received a lot of attention recently, in view of the requirement to minimise the activity levels of these elements in the wastes. Horwitz et al have advocated the use of solvent extraction using novel bifunc- tional organophosphorous solvents for such recovery[1]. Ion exchange being a very suitable choice for such applications, it has been our interest to devel- op ion exchange processes for actinide recovery. Recently, Alexandratos et al [^,3] have synthesised novel bifunctional resins which have shown great prom- ise for a variety of applications where conventional resins are of limited use. However, the studies of Alexandratos et al have been mainly devoted to recovery of transition metals and noble metals. In our study, we have explored in detail the possibility of the use of the bifunctional phosphinic acid resins for the recovery of uranium, as a part of our studies on actinide recovery from waste solutions. The results of batch experiments conducted with the bifunctional resins are presented in this paper. EXPERIMENTAL: 1. Synthesis of bifunctional resins: Styrene-DVB copolymsr beads of crosslinkages 4 %,' 6 %, 8 % and 10 % were received from.M/s Ihermax ltd., Purie and used as such. 2 % Styrene-DVB copoly- mer was syrtthesised by suspension polymerisation using benzoyl peroxide as initiator and polyvinylalcohol tur suspending the styrene-DVB copolymer in aqueous medium.The mixture was refluxed at 90 C for 6 hrs,filtered and dried. The copolymer beads were refluxed with PCK and anhydrous A1CU for 6 hrs, followed by filtration and hydrolysis with 1 M sodium hydroxide at 0 C. This prncpchire yields f-:sin with phosphinic acid groups attched to the benzene ri'.i:/ "f t)).i ]•!)! vsi ,-t cue harHxui';. The phosphinic acid groups act as cation exi hd.-iyers, a.-d Hit i'=U t:rr);.|, ta.i form coordination bond with the metal spe- cis:. J he f iific! !"i :i i istid s<:-sins were characterised [2] with respect to P c,i:!en!, ••?•,)'];\> ",\bh' !iyf!;-'j^en, and °-H bonds. Typical values were: P content 1b%, re; l.icf.ii'in !.y!'v IJIJM i.H meq/tj and P-H bonds 3.5 meq/g. 250 mg of these resii.-s w-i e &c|ti i i ibi at ;:<\ WIMI 3 ml of U-233 solution in nitric acid of differ- ent -...om -••Li "ii \r,•;•;. i ;u I ! Mi) ,!' iri's were carried out at a constant temperature of ~y.) .|r-'i [ :•: ?'i 'i'-. \v, initial pxperiments, it was confirmed that equilib- rium \n,v, s ,>uc!'•••! in IK)': i-ei \<:A. After equilibration, the U-233 in aqueous phase was dttonv.iapi! by iitjuid scintillation counting. Distribution coeffi- cient (K.) was then calculated as (U(VI) per g of resin/U(VI) per mL of aqueous solution). RESULTS AND DISUNION: (Jala on K values nieasured with resins of varying crosslinkages are presented in Kig.1 as a function of aqueous acidity. It is seen that the K. values in general decrease with increase in acidity from 0.1 M to 2 M, ana then exhibit an increase. This behaviour is more pronounced in the case of SST - 17.1 resins with higher crosslinkages. The change in K. with acidity is less pro- nounced, and the value remains fairly high throughout the range for low cross- linkage. The decrease in K , with increase in acidity is understandable, as this is the general trend expected for cation exchange. The significant in- crease in Krf with increase in acidity from 2 to 6 M is clearly due to the bifunctional nature of the resin. The P=0 bond in the resin aids in the ex- traction of U(VI) as the neutral urany.l nitrate complex. Such a mechanism is not expected to be a significant factor in ion exchange using conventional sulphonic acid resins. The trend observed with the bifunctional resin indi- cates that U(VI) can be recovered from solutions of moderately high nitric acid concentrations also. With an increase in crosslinkage, the K . value for 0.1M and 0.5 M acidity is seen to decrease. This also is an expectea trend. In the case of 2, 4 and 6 M acidity, the trend is seen to be different, with the K, value decreasing first with increase in crosslinkage, and then increasing. In general, the K . values obtained with the bifunctional resins were found to be significantly higher than the values obtained with the sulphonic acid resins. The data presented in the present work indicate that the bifunc- tional resins have a lot of potential for applications in actinide from nitric acid solutions of any concentration. recovery REFERENCES: 1. E.P.Horwitz, D.G.Kalina, H.Diamond, G.F.Vandegrift and W.W.Schulz, Sol. Ext. Ion exchange, 3, 75(1985). 2. S.D.Alexandratos, Macromolecules, 18, 829(1985). 3. S.D.Alexandratos and W.J.Mcdowell, Sep. Science and Technol., 22,983(1987). D Cross linkage D 2°/o O4V, • 8% A 8% 1.0 ,O 10% -1.000 -0.301 0.324 0.602 0778 M [6]M LOG [HNO3] F16.1 Variation of Log ( K M.Venkat esan , T.N.Ravi, M.C.Devarajan, R.NataraJan, V.R.Raman and G,R.Balasubramanlan Reprocessing Development Laboratory Indira Gandhi. Centre for Atomic Research Kalpakkam - 603 102 SUMMARY Laboratory studies were carried out for the removal of Tri- butyl phosphate from aqueous solutions by washing with Dodecana. The effectiveness of the wash was found by batch experiments and the removal efficiency was calculated. The removal efficiency was «?lso studied with the variation of contact time of the wash. (Key Words : Reprocessing, Purex process, Tributyl phosphate) I. INTRODUCTION : Among the aqueous processes used for treating the Irradiated fuel elements, Purex process using Tributyl phosphate(TBP ) in an Inert hydrocarbon diluent(Dodecane) as extractent has been widely employed. In spite of many merits of TBP es a solvent It has some weaknesses also which result in process problems. Because of the solubility of TBP in aqueous solutions, frequently the aqueous streams leaving the TBP extraction columns contain organic which finally find their way to product evaporators, where they are subjected to thermal de- gradaticu under high acid conditions. The formation and presence of HDBP can give raise to heavy organic phases and precipitates which Interfere with subsequent process operations. The presence of TBP in aqi aous streams has also been blamed for possible explosion hazasds(l). Therefore TBP from aqueous streams have to be removed as tor and as soon as possible. Steam stripping and diluent washing methods are being followed for the removal of TBP (2). This paper describes the experiments carried out in our Lab to find out the effectiveness of dodecane wasii for TBP removal, II. EXPERIMENTAL : Aqueous solutions of 3N and AN Nitric acid with and without Uranium, end also water were taken as aqueous feeds. These aie shaken with 30% TBPln Dodecane(1:1) for 4 hours Thfi resulting aqueous was subjected to three consequent Dodecane contacts (1 :1 ). for thirty minutes each time. At each contact the aqueous sample wes collected and analysed for Phosphate. To find the removal efficiency with respect to contact time, the initial aqueous that contained TBEJ was shaken with Dodecane (1 :1) for different intervals oi time. Aqueous sample for each contact was collected and analysed. The results are piotted(fig.1}. To find the TBP content of the samples, the organic phosphate was first converted to inorganic ortho phosphate by fuming with Perchloric acid, Then the phosphate content of the sample was determined by spectrophotometry(3). The amount of TBP was calculated as TBP In ppm = (Phosphate in ppm)X(266/95) SST - 18.1 III. RESULTS & DISCUSSION: The results of the batch extraction experiments are given in table. The table shows that with Increase In acidity the TBP solubility decreases. The difficulty in removal of TBP is also more with increase in acidity. Flg.l which gives the rate of removal of TBP confirms the above obser- vation. The same trend is observed with increase in Uranium. IV. CONCLUSION : From flg.l it is clear that TBP removal Is slower as the aqueous becomes leaner in TBP. Unless this problem is overcome by another component which tacan foriorm a stronger bond with TBP this removal procedure has a limitation. V. REFERENCES : (1) & (2) : Engg. for Nuclear Fuel Reprocessing by Justin J.Long. (3) : TID 7015 ORNL, 9 00716011. FIGURE i TABLE TBP in TBP <•j/l> in aqutoui Re»ovsl 100 Aqutoui aquaout af tar Efficiency itti iar wash-1 •9/1 it Nash-1 dash-2 I 4N HNO3 208.3 29.4 BOL BDL 85.9 3M HNO3 245.3 • 33.0 SOL BOL 86.5 HATER 439.6 54.9 N.D N.D 87.5 724/l Uraniua 154.4 5. .8 113.0 BDL in 4N HNO3 72 «/l Uriniua 173.1 34.2 13.4 BOL 81.2 in 3N HMO3 BOL • BELDM DETECTABLE LEVEL N.D • HOT DONE 5 10 IS 20 CONTACT TIME IN MINUTES • 4N HNO j + 3N HNO3 SST - 18.2 AE - Actinsdes and Environment Papers : AE - 01 to AE - 03 GEOCHEMICAL ASSOCIATION OF AMERICIUM IN TROMBAY SEDIMENT V.M^ Matkar, U. Narayanan, I.S. Bhat and K.C. Pillai Health Physics Division Bhabha Atomic Research Centre Bombay - 400 085, India SUMMARYrGeochemical association of americium in Trombay coastal Sediment is studied. The amount of americium bound with organic matter fraction of sediment is an important factor in determining the extent to which an element would be released and available for the uptake by bottom feeding biota. 241 Key words: Geochemical association of Am in sediment I. INTRODUCTION: The observed distribution factor (10 ) tor Am in Trombay s-diraent is higher by one order of magnitude than Kd for Pu/1/. The geochemical association o:: Pu in sediment has been studied and reported earlier/2/. The reraobilisation of Am from the sediment would depend upon its association with the sediment. According to Gibbs/3/. the association of heavy metals with aqueous solid m?y be divided into four broad types: adsorptive bonding, coprecipitation by hydrous Fe-Mn oxide, by coinplexation with organic molecules and by incorporation into crystalline solid structure. The first three of these are associated with non-lattice held sediment fraction and the last with lattice held fraction. The transuranic located in lattice positions can usually be considered/4/ to be immobile i.e. environmentally unreactive, whereas those in non-lattice sites can be considered to be at least potentially mobile, i.e. environmentally reactive, in the biological & chemical processes which occur in the sediment/interstitial water complex. Investigation on the chemical partitioning of transuranics in sediment by leaching techniques have been rather limited and dealt mainly with plutonium. Study on geochemical association of Am in Bombay harbour bay sediment by the sequential leaching techniques is reported in this paper. II. EXPERIMENTAL: Two grab sediments- were collected from the area near to the low level liquid radiowaste discharge point and at a distance of 120 m from it. A homogenised sediment slurry was made with distilled water. A known aliquot of sediment slurry was taken for wet and dry weight relation and to estimate total concentration of 1Am. Another aliquot of the same sediment slurry was taken for chemical sequential leaching similar to Pu studies carried out earlier/2/. The extraction was continued for 18 hours i ng magnetic stirrer at room temperature. Thfe fraction obtained after each extraction was separated immediately by centrifuging. The supernatant was filtered through millipore (0.22 urn) filter AE - 01.1 241 paper. Am was estimated in the filtrate using the radiochemical separation/5/. Am was used as internal tracer for recovery correction. Electroplated sample was counted by alpha spectrometry. The reagents used for sequential leaching and classification of extracted phases are given in Table 1. III. RESULTS AND DISCUSSION: The data on Am obtained ' Table 1) are compared with the partioning of plutonium in the sediment from the same area. As seen from Table 1. the association of Am in the lattice held fraction was 27 to 38% whereas Pu had shown only 4 to 13%. This means that Am is less available than Pu. Among the non lattice held, Am showed 50.0 to 59.0% bound with organic molecules, 8 to 10% with hydrous Fe-Mn oxide, 6 to 7% with carbonate while Pu association had shown entirely different pattern, that is 30 to 40% organic, 45 to 55% hydrous oxide and 0.1 to 0.6 % with carbonate fractions. However, Pu and Am showed similar behaviour as far as exchangeable fraction is concerned both showing almost nil. It is observed from the above data that association of Am in organic fraction is significantly higher than that of Pu. This is in agreement with findings reported on the relative solubility of Pu and Am in seawater in presence of humus matter/6/. IV. REFERENCES: 1. V.M. Matkar, E. Mathew, N.N. Dey, M.C. Abani and K.C. Pillai, 12 Indi, J. Mar. Sci., 72 (1983). 2. V.M. Matkar, U. Narayanan, I.S. Bhat and K.C. Pillai, J. of Rad. Anal. & Nucl. Chem. Articles ( in press). 3. Gibbs R; Science, 180, 71 (1973). 4. R. Chester and S.R. Aston, IAEA, Vienna, 173 (1981). 5. E. Mathew, V.M. Matkar and K.C. Pillai, J. Radio-Anal.Chem. 62, 265 (1981). 6. V.M. Matkar and K.C. Pillai, J. cf Rad. anal, and Nucl. Chem., Articles, 138, 93 (1990). 241 Table - 1: Extracted phases, reagents used and results on Am 241 Am in Exchangeable Carbonate Organic bound Fe-Mn oxide Residual sediment [0.05MCaCl ] [0.5MCH COOH][0.1M t.lMOxa+Amm.Oxa.] [HF+HNO ] Na4P2°7] 0.51 Bq BDL 0.037 Bq 0.3 Bq 0.04 Bq 0.14 Bq (7.2%) (59%) (8%) (27%) 0.21 Bq 0.002 Bq 0.012 Bq 0.11 Bq 0.021 Bq 0.08 Bq (1%) (5.7%) (50%) (10%) (38%) [ ] leaching reagents, ( ) % extracted in each phase, S.D. 8 to 30% AE - 01.2 URANYL UPTAKE IN CEMENT HYDRATION PHASE : 11A TOBERMORITE. N.K. Labhasetwar and O.P. Shrivastava Department of Chemistry, Dr. H.S. Gour University, SAGAR - 470 003 (M.P.) India SUMMARY- Synthetic 11A tobermorite Ca5Si£ O^-^O exhibits V0\ uptake in aqueous medium. 0.93-14.57 wt% of uranyl was taken up by tobermorite when placed it in dilute solutions of U0*+ (100-3000 ppm). The uptake appear? to be a combined effect of cation exchange with the release of calcium from tober- 2+ morite and also sorption of UO2 on the surface of synthetic-crystalline compouri. Low cost synthesis and nonreactive nature of synthetic tobermorite support its potential for uranyl removal from radioactive waste solutions. KEY WORDS- Uranyl uptake; 11A toberiPorite. I. INTRODUCTION- An important phase formed during the hydration of ceirents is identical with the IlA tobermcrite, \ calcium silicate hydrate clofe to 1 Ca5Si60,eH£ .4H20. Recently synthetic I. * "cobermorite has been recognised as effective cation exchanger in alkaline or nearly neutral medium/1-3/. We have already reported the uptake properties of synthetic 11A tobermorite towards different cations/3-5/. The aim of the present study is to esr.mine the uptake of U0£+ by tobermorite and to investigate the mechanism of uranyl uptake in this low cost silicate material. II. EXPERIMENTAL- Tobermorite was synthesized hydrothermally by the procedure described by Kalousek/6/. The reactants calcium oxide and silicon dioxide were mixed in ratio of 0.8 alongwith sufficient decarbonated water. The contents were autoclaved at 175°C for 48 hrs. The hydrothernial reaction product was characterised with the help of elemental analysis, infrared spectrum and X-ray diffraction. Electron micrographs of the compound show typical plate like crystals of tobermorite. Uranyl uptake studies were carried out following batch process. Synthetic tobermorite was equilibriated with aqueous uranyl nitrate solutions (initially tO'1-3000 ppm) at 29CC for 10 days. Analytical data were generated from analysis of both solid and solution phases. Few uranyl contain- ing tobermorite samples were also investigated by Electron Probe Microanalysl* using CORA, aa electron microscope equipped with energy dispersive analytical system. The preparations were dispersed on grids using ultrasonics. III. RESULTS AND DISCUSSION- Characterisation of the synthetic compound confirm the single phase preparation of synthetic, normal, 11.3 A tobermorite. The uranyl uptake data (Table 1) reveal a concentration dependent substitution of Ca from tobermorite for U0;>+ with a maximum 0.4806 m mols Ca release per gram of tobermorite. As much as 14.57 wt% of uranyl could be incorporated in tobermorite after equilibriatioii with 3000 ppm uranyl solution. Molar ratio of Ca release and uranyl uptake is not stoichiometric (Table 1) for higher uranyl concentrations. Thus the uranyl uptake is not only an ion exchange reac- tion but also the result of surface adsorption or other reactions with silicate material. The presence of uranyl in tobermorite was investigated by electron probe microanalysis. These results reveal that uranyl is significantly present on the surface of the tobermorite and also in the crystal. In this way the uranyl uptake is a combined effect of surfaces and broken bonds along with structural Ca displacement from tobermorite. The true exchange is followed by an irreversible reaction that takes place to attack the surface at high concentrations of uranyl in solution. The massive and complex geometry of AE - 02.1 uranyl cations is responsible for low displacement of Ca from tobermorite* X-ray diffraction analysis of uranyl containing tobermorite shows uneffected crystallinlty and structure at lower concentration of uranyl solutions. Although uranyl uptake in tobermorite is not a thermodynamically true ion exchange but the high uptake capacity of the compound is remarkable. Owing to the low cost of synthesis and nonreactive nature, • synthetic tobermorite could be useful in uranyl removal from radioactive solutions. Table 1 : URANYL UPTAKE DATA OF 11A TOBERMORITE. Initial U0|+ Uranyl in * Ca released Ca release Ky concent ratIon solid phase d from tober. U02 uptake (ppm) (wtZ) (mmols/g) (mols) 100 0.93 1500.0 0.0343 0.99 500 4.40 1024.8 0.1591 0.94 1000 7.86 524.0 0.2905 0.93 1500 10.64 332.6 0.3855 0.89 2000** 12.63 229.6 0.4366 0.83 2500 13.83 153.6 0.4717 0.81 3000 14.57 113.8 0.4806 0.77 * Kd is defined as the uranyl sorbed per gram of sample to the unsorbed uranyl per ml of solution. ** Electron probe microana.lysis was carried out on this sample. IV. REFERENCES- 1. S. Komarneni, R. Roy and D.M. Roy, Cea. Concr. Res., 16, 47 (1986). 2. O.P. Shrivastava and F.P. Glasser, React. Solids, 2, 261 (1985). 3. N. Labhasetwar and O.P. Shrivastava, Ind. J. Che*., 27A, 999 (1988). 4. N. Labhasetwar and O.P. Shrivastava, React. Solids, 7, 225 (1989). 5. N. Labhasetwar and O.P. Shrivastava, J. Mater. Scl., 24, 4359 (1989). 6. G.L. Kalousek, J. AH. Ceran. Soc, 40, 124 (1957). AE - 02.2 PREPARATION OF 236Pu BY 235U(o.3n) RBACTION R.J.Singh, S.M.Deshmukh, A.Ramaswami, S.S.Rattan and Satya Prakash Radiochemistry Division, B.A.R.C. Bombay 400 085, India SUMMARY: Trace level of Pu has been prepared by the U(a, 3n) reaction. The alpha spectrum™ of the purified sample showed tiie.presence of Pu and Np, w' .ch is the daughter product of 2_?Pu. (KEY WORDS:( Pu, Pu, Nuclear reaction, Alpha- spectrometry, Pu purification) INTRODUCTION: In recent years there has been increasing interest in the biological and environmental effects of piutonium. For these,studies__the two important isotopes of plutonium used are Pu and Pu, d->*» to their relatively short half lives. These isotopes of plutoniun.- cannot be prepared by neutron irradiation and are the products of charged particle reactions.„_ This paper reports the work on the preparation of Pu by U(a,3n) reaction using the alpha particle beam at VECV Calcutta. EXPERIMBNTAL: Fig.(ll-shows the stack foil arrangement used for the irradiation. U (~1.5mg/cm ) (~ 94% enriched) targets were prepared by electrpdeposition on 25.4um thick aluminium foil. The stack foil was irradiated by 40MeV alpha particle beam (current 5iiA) for 10 hrs. This arrangement of the stack foil was chosen based on the literature data on the reaction cross section/1/. The alpha particle energy on the three target assemblies was 37.2,35.2 and 33.1 MeV. After irradiation the targets were dissolved together and taken up in 2M nitric acid. Plutonium was extracted by the standard TTA extraction procedure. Plutonium was further purified by ion exchange method. This method of the extraction and purification of trace amount of plutonium gave the best result. RESULTS AND DISCUSSION: The alpha spectrum of the purified Plutonium is shown in the fig.(2). The alpha spectrum shows the presence of Pu and Np, which is lag daughter product of " Pu. The presence of trace amounts of Pu and Pu is also se»n in the spectrum. These isotopes of 2plutonium are produced by the alphn particle interaction on U,present in the target. The amount of Pu produced for 50 uAhour/rog of ^JDU is 90 Bq. CONCLUSION; We have besn able to presage for the first time in our country trace amountb of Pu for environmental, biological and nuclear studies. ACKNOWLEDGEMENT The authors are grateful to Dr.R.M.Iyer, Director Chemical and Isotopes Group and Dr.P.R.Natarajan Head, Radiochemibliy Division for their keen interest in this work. The help of the VECC staff is also gratefully acknowledged. REFERENCE: 1.H.Delagrange,A.Fleury and J.N.Alexander Phy.Rev.C 17.No.5 1706,(1978). AE - 03.1 ALUMINIUM FOILS (?o.4 /j.m THICK) i t 235U TARGET 2 { A/ 1.5 mo/cm ) T T & T \- ]( 2 3 TARGET ASSEMBLY Tl T2 T3 FIG.-1. STACK FOIL ARRANGEA'.ENT USED FOR THE PREPARATION OF 236Pu . 5.77M«V 4000 236 Pu 4.79MeV 237, 3000 2000 O J000 ' ' •>)*•. U00 1500 1600 1700 1800 CHANNEL NUMBER FIG.-2. ALPHA SPECTRUM OF SEPARATED PLUTONIUM. AE - 03.2 AC - Analytical Chemistry of Actinides Papers : AC - 01 to AC - 18 RADIOCHEMICAL STUDIES IN CHEMICAL SEPARATION OF RARE EARTHS IN PLUTONIUM AND THEIR ESTIMATION BY EMISSION SPECTROGRAPHIC METHOD B. A. Dhawale, B. Rajeswari, T. A, Bangia, M. D. Sastry and P. R. Natarajan Radiochemistry Division. B. A. R. C.» Trombay, Bombay- 400 065. SUMMARY- Radioactive tracers have been used for optimising extraction conditions for separation of plutonium and determining rare earths at trace levels using Emission Spectrographic methods. 0.5M Tri-n-octyl phosphine oxi* ' xylene / HC1 #ystem has been used for determining rare eacus? in the range 0.2 - 200 ppm with 100 mg plutonium sample. Key Words i Tracer Studies* Emission Spectrographic method I. INTRODUCTION The determination of rare earths in nuclear materials like V and Pu is an essential pact of the chemical characterization of these materials. Solvent extraction methods **'•** have earlier been developed using TLA and TnOA solvents for the separation of rare earths from these major matrices and their subsequent determination by ICP/AES and DC arc/AES. methods. Authors-* have eddiet deveolped TOPO/xylene/HCl extraction system for the extraction of Th. The present paper describes the use of TOPO/xylene/HCl system for quantitative separation and determination of rare earths in plutonium . The recoveries of rare earths and plutonium havr> been checked by using radioactive tracers and DC arc/AES method. II. EXPERIMENTAL Radioactive tracers of »""Ce, »•»*-> ^*Eu and »"^Gd have been used for standardising extraction conditions for extracting plutonium into organic phase leaving R. Es in the agueous phase. FOR TRACER STUDIES: A solution of 10 mg of plutonium in HC1 was shaken with 5 ml of 0.5M T0P0 / xylene system. Rart earth tracers namely Ce,Gd and Eu were added separately to a solution of 10 mg plutonium. Similar aliguots of plutonium and rare earth tracers were kept for reference. The solutions were shaken with 5 ml of 0. 5 M T0P0/xylene/6M HC1. This extraction was repeated 3-4 times. The organic and agueoi-s phases were counted for alpha and gamma radiations respectively. FOR SAMPLE ANALYSIS: The rare earths from 100 mg of plutonitim were separated and estimated by emission spectrographic method. The agueous phase was evaporated to dryness and redissolved in 250 inicrolitres of 6M HC1. 50 microlitres of this solution was loaded on graphite electrodes containing 10 mg charge of 1:1:2 LiF—AgCl-Graphite buffer. Thulium was used as internal standard. The spectra were photographed on SA-1 plates in the region 3150-3500 A in a d.c. arc carrying a 13 ampere current using 3.4M Ebert spectrograph. AC - 01.1 III. RESULTS AND DISCUSSION Both TLA and TnOA are liquid anion exchangers and extract U and Pu from acid medium forming anionic complexes. R. EB. and Th rsmain in aqueous phase as they do not complex with these extrsctants while TOPO is a neutral extractant and extracts U and Pu alongwith Th in HC1 medium. The order of extraction of Pu among neutral extractants is found to be highest in phosphine oxides as compared to phosphinates. phosphonates and phosphates. The extraction of plutonium in organic phase was checked by cK. -count ing and that of R. Es in aqueous phase by *y - counting. The results shown in Table-1 show a quantitative recovery of R. Es. No o(-counts above background were observed in aqueous phase showing quantitative extraction of plutonium. The recovery of R. Es was also checked by DC arc /AES method developed ear 1 iear *•J*. The AE1S results qiven in Table-2 show a quant itive recovery within experimental errors. The method can be used to determine the R.Es in the range 0.2 - 200 ppm using only 100 mg of sample. Lower limits can however be obtained using larger amounts of plutonium for extrqcf ion. IV. REFERENCES .1. R. K. Dhumwad, M. V. Joshi and A. B. Fatwardhan, Ana 1 . Ch i m. Acta. 42 < 2>,334 < 1968). 2. A. G. I. Dalvi,C. S. Deodhar and B. D. Joshi; Ta1anta.24,143< 1977> 3. B. A. Dhawale, B. Rajeswaci, T. R. Bangia,M.D. Sastry and P. R. Natacajan. Anal.Letters.23 TABLE - 1 Tracer Recovery Results Tracer Counts per minute % Recovery Before Extrn. After Extrn. Ce 4740 4650 98 Eu 30000 30000 100 Gd 80200 80100 99 Pu" 4085 3965 97 * — Alpha counts TABLE - 2 AES Recovery Resvrlts Element Wavelength Amount Amount Recovery Efetn. rim added(ug) recovered* ug) % Ranqe < pprr. > Ce 320. 1 0. 4 0. 47 117 2-200 Dy 340.7 0. 1 0. 09 98 0. 5-50 Eu 321.2 0. 1 0. 12 120 2-50 Er 326.4 0. 1 0.08 80 1-50 Gd 342. 2 0. 1 0. 1 100 0.5-50 La 331.2 0. 1 0.07 70 1-50 Lu 337.6 0. 1 0. 08 80 2-50 Sm 336.5 0. 4 0.4 100 S-200 Tb 332. 4 0. 4 0. 3 75 2200 Y 321.6 0.04 0.04 100 0. 2-20 Yb 328.9 0. 04 0.03 75 0.5-20 AC - 01 .2 NEW METHOD OF NON-DESTRUCTIVE ASSAY OF PLUTONIUM IN SOLUTION USING A SINGLE ISOTOPIC GAMMA-RAY SOURCB G.K.Gubbi, A.Ramaswami, R.J.Singh, Satya Prokash and P.R.Natarajan. Radiochemistry Jivision, B.A.R.C. "Bombay-400 085, India JUMMARY: A new method of non-destructive assay of plutonium in tolution, with built in matrix correction,- has been developed lsing a single isptopic gamma-ray source ( Se). ' KEY WORDS: Se.Non- Destructive Assay, Plutonium, Matrix Correction) [.-INTRODUCTION Compact K-edge densltometer based on t.-o sources [ Se and Co ) is generally used for accurate non-destructive issay of plutonium in solution. However Co is very expensive »nd needs to be,imported frequently. Hence a method based on only me source L Se) which is obtained indegenously has been leveloped. Se haB gamma-rays ( 121.1,136,264.6KeV) of good ibundance having energies just below and well above the plutonium C-edge(121.8KeV). As 136KeV energy gamma-ray is about 14 KeV above the Pu K-edge matrix corrections become pronounced and lence can not be applied for the solutions of unknown matrix. We lave developed a formalism for applying matrix correction by smperical determination of a linear correlation between the logarithm of the ratios of peak areas of 121,136 and 264Kev gamma-ray energies for pure uranium solutions. II. EXPERIMENTAL The details of the experimental set up is given alsewhere /I/. A ten mCi source of Se was used for the transmission measurements. Transmissions at 121,136 and 264 KeV aroma-ray energies were measured for the plutonium standard lolutions in the concentration range of 20-80 g/1 and for uranium tolutions in the concentration range of 0-70 g/1. II. RESULTS AND DISCUSSION Background subtracted peak areas 121' ^136' ^264 were determined from the transmitted spectrum of }ach sample. Linear correlation between log (Pi36/Pl2l) and log ^264/^121) £or uranium solutions is shown in Fig.l. Least square ltting was performed to get slope (m=0.2631) and intercept C-1.002). Concentration of plutonium is given by CO. ttiere k is a constant depending upon the mass attenuation :oefficients for Pu and path length of the solution in the sample rial. Fig.2.shows the linear variation - of the experimental >arameter with plutonium concentration. The same parameter is nsensitive to samples containing nitric acid or uranium in :oncentrations upto 70 g/1 (eqn. 1). Hence the method is Plutonium specific and insensitive to matrix. Purther experiments using mixed (U,Pu) solutions are in progress. BV REFERENCE I 1. A.Ramaswami, G.K.Gubbi, Satya Prakash and P.R.Natarajan Paper No.NI-05, presented in Radiochemistry and Radiation Chemistry Symposium, Kalpakkam, Jan.(1989). AC - 02.1 1.6 U4 - 70o/l U3 - 58 U2 • 42 g/l Ul - 28 CM CO L U0 -HNO3 2M 1.2 1.6 2.0 U 40 60 80 PLUTONIUM CONCENTRATION (g/l) FIG.-2. CALIBRATION CURVE FOR PLUTONIUM CONCENTRATION AC - 02.2 "*OLE OF URANIUM AND PLUTONIUM MATRICES IN THE ATOMIZAT1ON OF B«,Sn and Zn FROM A GRAPHITE FURNACE Neelam Goyal,Paru J. Purohit,A. G. Page and M.D.Sastry Radiochemistry Division Bhabha Atomic Research Centre*Trombay, Bombay 400 085,INDIA. SUMMARY:Atomization mechanism of Be.. Sn and Zn from a graphite furnace in the presence of U and Pu matrices has been investigated by measuring the temperature dependence of the absorbance.The activation energies and appearance temperatures obtained here,suggest that atomization of Be and Sn is preceded by carbothermic reduction of the respective oxidRB. On the other hand atomizat: ion mechanism for Zn could not be obtained conclusively.Experimental evidence, however, points towards the dissociation of ZnO as the probable mode of atomization. I INTRODUCTION: As part of the trace metal characterization of nuclear fuel materiaIs,electrothermal atomization-atomic absorption spectrometric methods have already been developed and reported from this laboratory /I,2/ for trace metal determination in U and U-Pu matrices without chemical separation of the matrix.During these studies, it has been observed that the absorbance signals for these analytes get affected to varying degrees in the presence of uranium and plutonium matrices, suggesting strong matrix effects on the atomization behaviour of the anal ytes. Invest igat i nns have, thctnf tiro, been carried out with a view to getting, insight into probable mode of atomization. These include, determination of the appearance temperature for the absorbance signal, activation energy and U/Pu concentration-dependent variation in the signal. The present paper deals with investigations of Be, Sn and Zn in U and Pu matrices. II EXPERIMENTAL: A Varian Techtron atomic absorption spectrometer AA-6 equipped with CRA-63 carbon rod atomizer and BC-6 background corrector formed the experimental facility.The absorbance signals were obtained in peak mode at different atomization temperatures. The temperatures corresponding to different voltage settings of the power supply were provided by the manufacturer. The details of solution preparation are as given in our earlier paper 121, The following studies were carried out here for each of the three analytes. KXPTI.. DATA Be Sn Zn MATRICES : Aqu. Pu Aqu. 0 Pu : Agu. U Pu App.Temp.< C) 1900 190u 0 1900 ' 1900 1900 1900 : 1400 1400 1400 Ea The absorbance signal for the two onalytes were found to enhance on the addition of xylene prior to atomization suggest ing the possible carbothermic reduction process resulting in formation of < Be, Sn>Eol id. The &G'=" value for Sn0a reduction has baen reported/5/ to bs negative at ths appearance temperature. Similsr data on BeO is not available; however*results obtained here,suggest carbothermic reduction of BeO as the probable process. Zn_: The appearance temperature of Zn was found to be matrix independent and Ea values were found to be matrix dependent. The E_ value for aqueous medium is significantly lower than reported /5/ wherein the authors have suggested dissociation of ZnO as the predominant mode of atomization. The matrix—dependent Zn aborbance signal and absence of any rise in it on addition of xylene as observed here,support the same viewpoint but we have no explanation for the lower E^ values for Zn atomization. Matrix Effect:(a) Be: The significant reduction of Be absorbance in the presence of Pu matrix has been interpreted as due to the formation of PuBe which has been reported to form under similar condit ions/4/, (b> Sn: The appearance temperature for Sn in the three matrices. is sufficiently higher than the temperature at which changes in the partial pressure of oxygen due to dissociation of []30a and PuOa in V and Pu matrices takes place. The absorbance therefore does not change when the matrix is changed from U to Pu.(c) Zn: The reduction of absorbance signal compared to aqueous solutions was found to be significant for uranium, whereas reduction is marginal for in caBe of Pu. This can be explained on the basis of respective change in partial pressure of oxygen at th« temperatures shown balqw: 0/U:3 67 0/U: 2 ii) 0/Pu:2 O/PuTl.97 These studies on the reaction mechanism involved in the atomization of an analyte from a refractory matrix are helpful in improving the sensitivity of the analytical technique and to overcome interference effects. REFERENCES: 1. N.Goya 1. P.J.Purohit, A.R.Dhobale, A. G. Page and M. D. Sastry,Fresenius Z. Anal. Chem.330, 114< 1988) 2. N. Goyal,P.J.Purohit,A. G. Page and M.D.Sastry,Proceedings of 5th ISAS Symp. on Analytical Spectroscopy including Hyfenated Techniques held at IICT,Hyderabad during January 1988 edited by M.D. Sastry and A. G. Page. p. 13. 3.CRC Handbook of Chemistry and Physics edited by R.C.Weast and published by Chemical Rubber Co.,Ohio, U. S. A. < 1970) . 4. Chemistry of plutonium by J.M.Cleveland published by Gordon and Breach Science publishers,N.Y.< 1970). 5. R. E. Sturgeon and C. L. ChaJcrabart i, Prog. in Anal. Atom. Spectrscopy 1, 147< 1978) AC - 03.2 DIRECT SPECTROGRAPHIC DETERMINATION OF SUB PPM LEVELS OF Ba, LI AND Sr AND PPM LEVELS OF Cs, K AND Na IN NUCLEAR GRADE THORIUM OXIDE CThOz) R. Venkataaubramanlan Spectroscopy Division Bhabha Atomic Research Centre Bombay-400 085, India SUMMARY - A simple ami rapid optical omission spectrographlij (OES) technique employing RbCl as carrier is described for the direct determination of trace alkalies and alkaline earths in nuclear grade ThC^. The carrier-distillation technique reported does not Involve prior chemical separation of the impurities from t .*• • sample and makes possible the following range of estimations Li:0.02 - 5.0 ppm; Ba, Sr:0.5 - 20 ppm; Cs:2 - 50 ppm; ' K, Na-5 100 ppm. The detection limit obtained by this method for Li and Sr is lower than that obtained by the earlier carrier-distillation methods reported and other direct sensitive techniques of Zeidel and Avni. [Key Words : Impurities in ThOfc, Carbon Arc OES] I. INTRODUCTION Subsequent to the use of thorium as a fuel and breeder material in nuclear reactors, direct spectrographic methods /I-3/ have been developed for the determination of trace impurities in thorium. Further to the development of a method /4/ in our laboratory for the determination of nuclear poisons such as B and Cd and 24 other trace impurities in nuclear pure ThO^., there was a need for the determination of two more impurities viz. Cs and Sr in ThCfe. Hence the present method was developed to determine Cs and Sr as well as to lower the detection limits for other alkalies and alkaline earths. The OES method reported here, which employs RbCl as carrier is simple and rapid and it also provides the required detection limits. II. EXPERIMENTAL The working standards of ThOz and the samples brought to ThCfe form were mixed with RbCl containing 5% CaCQj, in the ratio 19:1. Samples and standards loaded in UCC 1990 electrodes were excited in a d.c. arc 10A current. Spectra were photographed using Hilger F1219 filter during an exposure of 20 seconds on Kodak 1-N emulsion by employing Hilger glass prism apeotrograph. The optical densities of the line pairs presented in Table 1 were measured on a non-recording microphotometer and converted to intensity ratios and subsequently to analyte concentrations by use of ND computer. III. RESULTS AND DISCUSSION The emission spectrum of thorium is characterised by a complex array of closely spaced lines contributing to a heavy background. In order to obtain the optimum line to background ratio for the analytical lines, the matrix spectrum was suppressed by the use of the carrier-distillation technique devised by Soribner and Mullin /5/. In addition to the common AgCl carrier, RbCl was tried as a carrier material since Rb was not an analyte. From the rate of volatilisation of alkalies and alkaline earths in ThCt. with two different carriers tried at 5% level (Fig.l), it is evident that RbCl helps in enhancing the intensity r>f the analytical linos and AC - 04.1 Table 1. Analytical Data for the Determination of Alkalies and Alkaline Earths in Thorium Oxide Element Analytical Filter Internal R-jiifjo of Me.m Line(nm) Step Standard Estimation Std. % Trans- Line(nm) (ppm) Devi a mission tioo(%) Cs 852.11 10 Rb 629.92* 2-50 8.3 K 769.89 10 •• 5 100 11. 1 Li 670.78 100 •• 0.02 0.5 14.3 Li 670.78 10 •• 0.1 5.0 14 .1 Na 589.59 10 •' 5 - 100 14.2 Ba 455.40 100 Ca 558.87 0.5 20 16.7 Sr 460.73 100 *• 0.5 20 9.8 10% transmission step was used for measurement .4 ! i •A 1/ i • Fia.l- Volatilisation of Alkalies and Alkaline Earths in ThQj shows a definite improvement over AgCl in carrier action. As can be seen from Fig.l, with RbCl as carrier, five-fold increase in intensity was obtained for Cs and K lines as compared to the intensity obtained with AgCl. Hence RbCl (5X) was chosen as the carrier material for the present analysis. A survey of the literature revealed that the detection limit obtained by the present method for L.I and Sr is better than that obtained by the earlier carrier-distillation methods reported as well as other direct sensitive techniques such as Zeidel's evaporation method /2/ and Avni's dc arc cathode region technique /3/. The accuracy of the present method was assessed by analysing synthetic samples and the analysis showed good agreement between added and estimated amounts of the impurity elements. The precision of the method ranges f^oni 8% for Cs to 17% for Ba (Table 1). IV. REFERENCES 1. G.R. Blank and J.K. Reusch, USAEC Report NLCO-1078 (1970). ?,. A.N. Zeidel, S.I. Kaliteevski i , G.G. Kund and 2.F. Fratkin, Optika i Cpektroscopiya 2, 2U (1957). 3. R. Avni, Spectrochim. Aeta, Part B 24, 133 (1969). 4. S.V. Grampurohit, M.D. Si.ikaena, V.N.P. KaimaJ , S.K. Kapoor and P.S. Murty. Indian J. Tech. 19, 336 (1981). 5. B.F. Scribner and H.R. Mullin, J. Res. Nat. Bur. lit 3JT, 379 (1946).. AC - 04.2 BACK EXTRACTION OF Th(lV) FROM ITS TTA COMPLEX IN BENZBNE BY AQUEOUS FLUORIDE AND ITS APPLICATION IN THE ANALYSIS OF FLUORIDE IN NUCLEAR FUEL SAMPLBS fi.K. Rastogi, M.A. Mahajan aad N.K. Chaudhuri Fuel Chemistry Division Bhabha Atomic Research Centre, Trombay, Bombay 400 085 SUMMARY A spectrophotometric method based on Th(IV)-Arsenazo III complex formation with the Th(IV) back extracted from its TTA complex in benzene by aqueous fluoride and measurement of absorbance at 662.5 ni» has been developed. The coefficient of variation obtained are 2.4 and 1.4 percent in 11 determinations each at 1 and 10 Avg/mJ of fluoride • |Key words : Fluoride determination, spectrophotometry, Zr-Arsenazo lill, Zr-Thenoyl trif luoroacetonate, back-extraction] INTRODUCTION Formation of strong fluoride complexes leading to the back extraction of tetravalent cations from their thenojrltrifluoroacetonate {TTA) complexes in benzene medium by aqueous F was utilised to develop two sensitive methods for_the determination of fluoride .at trace level and reported earlier ' . One was a radiometric method using Hf as tracer in which the activity back extracted was proportional to the [F~J. The other was a spectrophotometric method using coloured Arsenazo III complex of the. back extracted Zr(IV) having very high extinction coefficient . In both the cases, however, a high acidity of the aqueous medium was imperative to prevent hydrolysis and polymerisation of Hf(IV) or Zr(IV) leading to somewhat high blank values. As Th(IV) is less susceptible to hydrolysis and forms strong complex with Araenazo III with high molar absorptivity (1.3 x 10 1. mol . cn~ ) at the absorption maxirausi 662.5 nm,' the possibility of using Th(IV) instead of Zr(IV) in the organic phase has been explored in the present work.. EXPERIMENTAL An organic stock solution of Th-TA complex containing 400-300 fig of Th(IV) per ml was made by extracting Th(IV) from 0.04-0.06 M nitric acid medium with a pre-equi1ibrated 0.1 M HTTA solution in benzene. 5 ml of the aqueous perchloric acid solution containing F~ having pH 1.5 WBB equlibrated with 5 ml of this organic stock solution for 15 minutes. An aliquot of the aqueous phase was transferred to a 10 ml volumetric flask and mixed with 1 ml of 1 M sulphamic acid (to destroy nitrous acid present in nitric acid), 1 ml of 1 M aluminium nitrate and about 3.8 ml of 16 M nitric acid (so as to make the final acidity AC - 05.1 6M). 1 ml of Arsenazo III solution ( 2 mg/rnl in 0.01 M NaOH) was then added and diluted upto the mark. The ahsorbance at 662.5. nm was measured against a reagent, blank. Various parameters were studied to optimise the conditions for the method of determination of fluoride at trace level. RESULTS AND DISCUSSION The pH of the aqueousiinedi uin was important as Th(IV) might be back extracted at high pH duei hydro 1 ys i H and at low pH due to exchaiigt- with H . A studstudy carried with aqueous medium with varying pH showed tha*: the ratio of the net absorbances obtained in • he presence of the same amount of F to that in its absence w«s maximum at pH 1 .5. Linearity of the absorbance with [F j was observed in the range 0 2 to 16 /jg/ml of fluoride. When 1 F solution at pH 1.5 was equilibrated with the organic stock solution containing 450 pg/ml Th(IV), the apparent molar absorptivity for _F after colour development was 6.3 x 10 l.mol cm . Below a [F~J of 0.2 ug/ml the absorbance values showed high variations presumably due to the hydrolysis and polymerisation of Th(IV) whirh may be prevented by the addition of a known quantity (if F (standard addition) to attain higher effective [F j. Colour remained almost constant at least «jpto 12 hr after mixing. In a study on the effect of diverse ions 5t was observed that the recovery was within + 5% when the solution contained 100 fold excess of C! and NO3, 16 fold excess of acetate, 8 fold excess of SO/, and PO4 , 80 fold excess of citrate, 80 fyld excess of Ca^*, Co and Cu , 16 fold excess of Cr . However Bi , Al , Zr , Hf and Th interfered seriously even at low level. Obviously cations forming strong complexes with fluoride or forming coloured Arsenazo III complexes which have any absorbance in 662.5 nm must be separated beforehand. For the analysis of fluoride.iropuri ty in nuclear fuel samples, it is separated by pyrohydrolyais and collected in acetate buffer (0.01 M HAc and 0.01 M NaAc). This solution is free from interfering ions and can be analysed by the present procedure. The results of analysis of a few solutions by the present procedure <. r- wrel 1 as by potentiometry using F ion selective electrode showed an agreement within +5%. REFERENCES 1. R.K. Rastogi, M.A. Mahajan, N.K. Chaudhuri and S.K. Patil, J. Radioanal. and Nucl. Chem. Articles. 133 (1989) 203. 2. R.K. Rastogi, M.A. Mahajan, N.K. Chaudhuri and S.K. Patil, Mikrochim. Acta (in press). 3. M.A. Mahajan, M.V.R. Prasad, H.R. Mhatre, R M. Sawant- R.K. Rastogi, G.H. Rizvi and N.K. Chaudhuri, J Radioanal. and Nucl. Chem. Articles 148 (1990) 93. AC - 0 5.2 DIRECT AND SIMULTANEOUS SPECTROPHOTOMETRIC DETERMINATION OF URANIUM AND ACIDITY USING A FLOW CELL A.D.Moorthy. P.B.Gurba, G.A.Chau«ule, N.Varadarajan, R.K.Singh, M.K.T.N^ir PREFRE Plant, Tarapur SUMMARY An attempt has been made to correlate the optical densities; oi uranyl nitrate solution, measured at three wavelengths, with uranyl (U(VI)) concentration and nitric acid roolarity. Shiroadsu UV-160 speotropholometer with a flow cell provision has been used. Computational prediction of the results is found to be with-in 4 % for U(VI) concentration and with-in 10 % for acidity. I.INTRODUCTION In the fuel reprocessing laboratory it is necessary to have quick analysis methods for the purpose of process monitoring. It is also desirable to use the methods which do not generate any waste. The conventional methods/I,2/ of free nitric acid and U(VI) determination are time consuming, generate corrosive waste and involve direct handling of radioactive liquids. To overcome these problems a computational method has been evolved. This method correlate.3 the optical densities of uranyl nitrate solution at three preselected wavelengths with 0(VI) concentration and acidity. The correlation is possible because optical density of uranyl nitrate solution is the function of U(VI) concentration as well as acidity. II.EXPERIMENTAL Uranyl nitrate solutions of 1 to 20 g/1 in 0.5 to 3.5 nitric acid molarity were prepared. The recorded spectrum revealed presence of one hump each on both the sides of wavelength maxima at 415 millimicron.(Fig.1) The variation of optical densities at wavelengths corresponding to the hump positions was observed to be depending on U{VI) concentration and acidity. Hence below mentioned relations were arrived at for the U(VI) concentration and acidily range mentioned above. (U(VI)C.O.D.)x = O.D.at 415mu - KfO.D.at 425 mu - O.D. at 405 mu) (Acidity)x = ((U(VI)C.O.D.)x - O.D.at 415 mu)/(U(VI)C.O.D.)x Where, O.D.= Optical Density, C.O.D.= Corrected Optical Density, (U(VI)C.O.D.)x = Factor proportional to U(VI) concentration, & (Acidity)x = Factor proportional to nitric acid molarity. Linear relations were observed between U(VI) concentration verses (U(VI)C.O.D.)x and acidity verses (Acidity)x. Corresponding slopes(m) and intercepts(c) were determined. Different uranyl nitrate solutions were prepared and subjected to optical density measurement using a flow cell for testing the validity of above mentioned correlations and the constants determined. U(VI) concentration and acidity in each case was determined using Davies Gray method and potassium oxalate medium acidity determination method, respectively. Comparison between the computed and analysed acidity and U(VI) concentration results is given in Table-1. AC - 06.1 Til.RESULTS & DISCUSSION On an average, computed value:-: show agreement of •/ 4% for IJ(VJ) coiiceiitratixi & ^/ 10 % for acidity with the re sed results. All the work reported in thi; paper was carried out using simulated solutions. This computational method is being rigorously tested with actual process samples for routine analysis. Developing of such ••» computational method can he considered as a step further t.uwfirdf, the inline monitoring of procers solutions using sen.si.rs like '>pt i«- F* 1 fih^rs IV ACKNOWLEDGEMENTS Author1.! tire grateful to Ghri AN F'r^sad, Director, Fuel Reprocessing and Waste Mana^eri!1";, L nratf>-:iii«-ii t given during the course of this work. Authors are also thankful to Ghri S. Venkate.".waran for hi i s help in analytical work. V.REFERENCES !. 0. J . Hodden , Anal. Chemist, ry of Hunlial l"n pi.-oier.-t , P luj, MoGraw Hill , (195.0) . 2. M.T.Kelly, ORNL - T I D 70 1 5, Selection 1.(1957). O.75" H^40s O-A5 CIS •feo +xo 44o Fig. 1 Optical Density Vs. Wavelength apuctra of ur&ny I nitrate solution in nitric acid medium. Table 1 si .• U(VI) (Joneentration g/l Acidity (M). No. ..- Analysed Computed Diff. AnalAnalysedy ! Computed! D j r 1 3.61 3 31 0 . 30 1 .03 0 97 0. 11 2 10.0 10.25 0.25 ' 0.61 0.62 0 .01 3 20.0 20.68 0. bB 1 . IB 1 . 1 0 .06 4 3.61 3. 2 b 0.36 2. 14 1 .88 (1 26 5 10.0 10.2 0.20 2.43 2.17 6 20.0 20. 02 0. 02 3.07 3.09 0.02 7 10.0 9.56 0.44 3. 45 3.33 0 12 8 4.24 4.12 0 12 2.52 2.20 0. "<2 9 7.67 7.70 ! 0.03 , 2.84 2.40 0. 44 10 18.03 17.62 0.41 3.58 3 . 20 (). 38 AC - 06 .2 SIMULTANEOUS DETERMINATION OF URANIUM ANV IRON IN H1GHLEVEL RADIOACTIVE WASTE STREAMS USING VUAL WAI/E LENGTH TECHNIQUE OF SPECTROPHOTOMETER C.P. Kauihik, R.G. Veotikat and Kamoan. Raj ii'IP, WMV, BARC, Tanapun Complex P.O. Gh4.va.li, Vist. Thane., PIN 401 502 (M.S) ABSTRACT The high level radioactive waste (HLW) is analysed fion vanious constituents by di^fienzn' methodi. Vi6fie.te.nt spe.ctKophotometn.ic mzthods ane available fion individual . 'talysU, ofi unanium and inon pn.z-ie.nt in HLW. Thiocynate is onz iach method which can be individually adopted {,01 ulanium and iKon analy&iA. In thi*- papei, detail** ojj iimultanzotu determination OjJ iion and unanium in the watte dz&ctibed which u)a& cafiiied out u&ing thiocynate Keagent with the help oft dual uuv-. length method of, ipectiophotometei. KEY WORDS: Vua.1 wave length technique, high level \adioactive wa&ie., ammonium thiocynate, iion, uranium. INTRODUCTION The high level ladioactive wa&te |HLW) geneiated during KepKoceaing oj$ ipeni nucleat fiuel ii immobilized in vitteouA matrix. A waitz immobilisation plant ai Taiapui uAei vitieoui matrix ($01 tizatmeni o£ thii type ofa wa&te [1], F01 ^inali-dation o{> vitieouJ, matrix composition, it i& eaential to know the detail* with n.e&pect to vaiioui ion& pte&ent in the waitz. Anaty&i& ofi individual element i& done on •ipecin.ophotomeiei uiing piope>i complzxing agent [2]. Ammonium thiocynate lowmb complexes with both ih.on and uranium. Al&o theiK. abiotption Apeci'ia ovztlap to -borne extent [3]. Thzie£on.z, intenfietence between thet>e two element* exi&t& dun ing ipectn.ophotomein.ic detznmination by fhii method. Dual wave length technique ha-i been selected in &uch a way that thene i& minimum inieihenence. The elemznti a'tz dzienminzd iimultaneouily. The wavz lengih a>ie 360 and 565 nm &on. U and Fe iz&peciively. Alcohol medium wa* uied to zliminate inte.nfien.encz due to anioni and to enhance the stability ofi the complexes. Intenfienzncz 6iudy ofi vaniou-t, element!, which an.e pn.e&ent in HUO has faeew cannied out. The nesults ane quite zncounaging and it is possible to detznmine U and Fe Simultaneously in HUM by thiocynate method. INSTRUMENT Shimadzu, Japan, make spectnophotometzn. Model UV-240 was uszd fion. the szudy. REAGENTS Ammonium ihiocijnate, inon powdz'i and unanyl ace-date ofi AR g\ade wzne used. Distilled ethyl alcohol was uied. EXPERIMENTAL 1. Standard solutions wen.e p'lepanzd fion unanium and inun. 2. Scanning ofi the coloured thiocynate complex was done fion unanium and inon finom 7.50 to 650 nm. 3. \fiten selection ofi wonking wavz lengths fion U and Fe, calibnaiion UMLS done fion U 10-50 ppm) and Fz 10-10 ppm). AC - 07.1 4. Sample iolutioni containing various concentrations o£ U and Fe wete and analysed under above calibration. 5. Simulated waite ioldiioni were prepared and antijied ai above. 6. Sample iolu.tioni with known concentrations oft U and Fe were prepared with o-the; ipecizi likz Ma, d, Oi, Me ta>i.e earihi zic. and analijied by ihii method !,o-x interference itudiei. RESULTS AIM DISCUSSION It wai ob-ie/ivecf ifeat (i hoi negligible, abio-xbance. ai 565 »vn aud i{ Fe it, mzoAWKid a£ Chit wave Izngih, U vjili noi have, any intzii.e.Ktnc.z. Homve.1, Fe intzilzie.i> in dz£?.>imina.iion oft It ai 360 tun. The. natiL-ie. oj above. inx.zi£z>ie.nce. i-xu> faund to be. additive. equation elimination o{, th.it, inie.t&e.ne.ftcz fan U analyiii. Working equation* both o-i U and ¥e. aw ai fiollom Cone o{, U Ippm) = 27J25 Abi. ai 360 urn - /2.60 Abi. at 565 :vn - 0.816 Cone. o& Fe (ppm) = 7.163 Abi. at 565 run - 0.1402 a{, 4ijnthe.lic U and Fe ioluiiont, aie given in Table, it can be thai the di^'\&n.encz bo.wee.n analijind and ttue concernAaUom ii in the >iange. of, 1-4%. Thai, thi4 meihod can bz u&ed fco'i iimuUamoim dete mi nation ofa Li and Fe in HLbl. Ii. wai fauna thai p>ie&ence o^ Mo doei iome. in\.e\;-e"ence on -\hi-i dual length when piie&ent in high concentration. However, e&faati a-ie continued io uie thii method, fat determination oft U, Fe and Mo uiing dual uavz length p'l.og'iamm?. fa* tiieir i i.multaneoui anaiijiii. ACKNOXLEVGEMEIIT Auihoii ai.e thankful io Shr.i CS Pwjailia'i and Shii. KT Ahir far aaiiiitig in experimental work during the couiie o^ thii iiudy. REFERENCES 1. Kantian tlaj & ! Sample Concentration Concentration Relative di^erence in % taken {ppm) determined Ippm) U Fe U Fe u Mx.1 5 2 5.1? 2.08 1.10 4.00 Mx2 JO 4 9.62 4.07 3.80 1.85 Mx3 20 5 20.07 5.12 0.35 2.4 Mx4 30 6 30.54 6.22 1.80 3.66 MxS 40 7 40.97 7.26 2.40 3.71 SU1 10 2. 36 9.65 2.18 3.49 3.38 SW2 50 4.72 49.47 4.55 1.06 3.60 SW3 20 4.72 19.39 4.52 3.05 4.23 Notei: SO! = Emulated VJa-iie AC - 07.2 SIMULTANEOUS SPECTROPHOTOMETRIC DETERMINATION OF URANIUM (VI) AND IRON (III) IN PUREX PROCESS STREAMS George Thomas, B.M.Patil, N.Varadara.Jan. R.K.Singh, D.D.Ba.jpai and M.K.T.Nair. PREFRE Plant, DARC, Tarapur. SUMMARY A Spectrophotometric method ha.s been developed for the simultaneous estimation of uranium and iron in purex process waste concentrates using ammonium thiocyanate as chromqgenic reagent. Shimadzu UV-160 3pectrophotometer with in — built provision for data processing was used. Key words : Purex, Uranium, Iron, Spectrophtometry. I.INTRODUCTION The need for rapid determination of uranium with reasonable amount of accuracy is one of the prime requirements in process control analyses. In Purex process uranium in waste streams is analysed spectrophotometrically using ammonium thiocyanate as chromogenic reagent.(1.2) But the presence of iron (III) interferes by forming colour complex with thiocyanate. In the conventional method this interference is overcome by selective reduction of ferric to ferrous. With the advancement in the sophistication of spectrophotometers the interference due to iron can not only be overcome by suitable processing method but also can be used for the simultaneous estimation of it. The additional advantage of the iron estimation along with uranium in certain waste streams is of significant importance from the vitrification point of view. Thi3 paper describes two wave length absorbance method for the simultaneous estimation of U (VI) & Fe (III) in single aliquot. II.EXPERIMENTAL Uranyl thiocyanate complex has a A max at 340 rnu and ferric thiocyanate complex has a}\ max at 474 mu. The interference of iron in uranium estimation is corrected by taking optical density at two different wave lengths,375 mu and 600 mu,where the optical density due to iron is same. For iron estimation optical density was measured at 500 mu wherein interference due to uranium is nil. Two sets of standard graphs were prepared independently for uranium (VI) and iron (III).Aliquots containing different amounts of uranium with constant amount of iron were pipetted in 2&ml Std flask and made up wi^,h ammonium thiooyanate solution. Optical densities were measured against the blank both at 375 mu and 600 rou. Difference in optical density (O.D.at 375 O.D.at POO) was plotted against concentration of uranium.(3) Similarly different amounts of iron with same amount of uranium were pipetted in 25ml Std flask and made up with ammonium thiocyanate solution. Optical density was measured at 500 mu and AC - 08.1 ^ plotted against, iron concentration. Simulated samples were analysed for uranium and Iron concentrations using these graphs by measuring optical densities at three different wave lengths. Difference in optical density of 375 mu and 600 mu was used for uranium concentration calculation from the graph and optical density measured at 500 mu was used for the iron concentration calculation from the graph. Results for the simulated process stream condition samples obtained by this method are given in Table-1. III.RESULTS AND DISCUSSION Thib method was found to be accurate and precise for the estimation of Uranium (VI) and Iron (III) in aqueous samples. The relative standard deviation of uranium and iron at 0.57 mg. uranium and 0.05 mg. of iron was found to be +/- 2%. Method is easily amenable for routine analyses in round the clock shift operation. IV.ACKNOWLEDGMENT Authors are thankful to Shri. A.N.Prasad, Director, Fuel Reprocessing & Nuclear Waste Management group, for his support and encouragement during the course of this work. V.REFERENCES 1.C.E.Crouthamal & C.E.Johnson ,Anal taem ,vol 24, 1780, 1952 2.M.T.Kelley ORNL Report, TID-7015, Section 1, 1957. 3.Instrumental manual of Shimadzu UV-160, 1989. TABLE 1 ! SI. ! URANIUM Std. \ IRON III ! URANIUM ! IRON III ! ! No. ! TAKEN in mg 1 ADDED in mg iESTIMATEDmg !ESTIMATED mg! ! 1 ! 0.2317 ! 0.025 i 0.23 I 0.024 ! 2 0.4635 0.025 0.455 i 0.0255 ! 3 0.5793 0.025 ! 0.585 ; 0.025 ! 4 1.1587 0.025 1 . 17 0.026 S 5 2.3173 0.025 2 . .'53 0.026 ! 1 6 0.5793 0.01 0. 56 0.011 ! 7 0.5793 0.02 0.58 0.023 ! 8 0.5793 0.025 0.585 0.026 ! 9 0.5793 0.05 0 . 57 : 0.055 ! 10 0.5793 0.1 0.585 ' 0.115 ! r AC - 08.2 EMPLOYMENT OF GRAPHITE ELECTRODE IN THE COULOMETRIC DETERMINATION OF URANIUM AND PLUTONIUM H.S.Sharma, R.B.Manolkar and S.G.Marathe Fuel Chemistry Division Bhabha Atomic Research Centre, Bombay 4-00 085 SUMMARY Graphite has been employed as working electrode to see its feasibility in the coulometric determination of both uranium and plutonium. Background currents at -0.325 V, and +0.700 V, the commonly required potentials for reduction of uranium and oxidation of plutonium respectively in IN sulphuric acid are only 30—40 mi 1 I i coulombs (MC) for five minutes. Experiments with stand." uranium and plutonium solutions show that both can be determine^ employing the same electrode. (Key Words: Uranium, Plutonium, Graphite Electrode, Coulometry ) I INTRODUCTION A mercury pool electrode 1B exclusively used for the coulometric determination of uranium but the same cannot be used for plutonium determination. Platinum has been extensively employed for the coulometric determination of plutonium but becomes unfit in the cathodic region for the determination of uranium. From this point of view carbonaceous electrodes such as graphite, glassy carbon may become a good choice as they can be used in a good range of potentials on the anodic and cathodic sides. Besides, these materials possess high electrical conductivity, good tensile strength, resistance to chemical attack etc. Also ease of availablity and low cost are additional attractive features. Very few reports exiat in the literature on the use of graphite or glassy carbon electrode /f II EXPERIMENTAL Graphite rods (density 1.78 g/cc) were used in fabrication of electrodes. Cylindrical, disk, well and beaker shaped electrodes of differing surface areas were fabricated and employed in the present work. Electrical contact to the electrode was made with a platinum wire. A platinum gauze was used as counter electrode and a saturated calomel as reference electrode. All these electrodes were enclosed in a glass cell of 40 mm diameter. A Coulometer giving very accurate and stable potentials and also displaying digital accumulated charge value was used. A personal computer was employed for on-line acquisition of charge and time data. IN H2SO4 was used as electrolyte. High purity Argon gas was purged to maintain oxygen-free atmosphere. Standard solutions of plutonium and uranium were prepared by dissolving purified PuOj AC - 09.1 and U3OQ and determining their concentrations by well established redox methods. Aliquots of these standard solutions were added on weight basis. Ill RESULTS AND DISCUSSION Initial feasibility experiments were carried out for studying uranium (VI) reduction by employing various graphite electrodes. The background current at -0.325 V, in IN sulphuric acid ranged between 30-40 MC for five minutes duration. First few experiments with the cylindrical and disk shaped graphite electrode (suface area •>* 20 cm ) indicated that only 10-20% reduction of uranium(VI) was possible in about 30 minutes. By providing additional horizontal and vertical holes in the cylindrical electrode the reduction of uranium to the extent of 50—70% could be enhanced. Further experiments were carried out by employing well and beaker shaped electrodes (surface area ^50 cm ). Almost quantitative reduction of uranium (VI) could be achieved in about 60-80 minutes. Reduction of plutonium to t.rivalent state and oxidation back to Pu (IV) was also studied employing the beaker shaped electrode at +0.300 V, and +0.700 V respectively, in IN H2SO4. Quantitative reduction of plutonium to Pu (III) could be achieved in about 40 minutes duration. During the course of repetitive experiments sluggishness in electrolysis was observed. This may be due to decrease in active centers. Probably the deposition of impurities from electrolyte solution could affect the activity. Dipping the electrode in nitric acid and polishing with 400 grit silicon carbide paper was found to help in regeneration of the electrode surface. Since the process of electrolysis ,for uranium and plutonium was slow, a computer prediction method was employed to calculate Q<» , coulombs corresponding to actual end point, from the initial data, stored on-line, in a personal computer. On the basis of this method, a few working standards were analysed for uranium and plutoniun contents. The results obtained agreed within + 1% to the expected value. More work is to be carried out for optimizing proper treatment to regenerate the electrode surface. Glassy carbon electrode shall also be employed in future work. IV. CONCLUSION Employment of the same graphite electrode for reduction of uranium (VI >IV) as well as oxidation of plutonium (III—>IV) appears to be feasible. Although the electrolysis is slow, proper surface conditions and treatment for regeneration of the electrode surface are needed to be optimized. ACKNOWLEDGEMENT The authors gratefully acknowledge with thanks the encouragement given by Dr.H.C.Jain, Head, Mass Spectronetry Section and Dr.D.D.Sood, Head, Fuel Chemistry Division. V REFERENCES 1. C.E.Plock and J.Vasquez, Talanta, 16, 1490 (1969). 2. N.M.Saponara and D.D.Jackson, Rep-LA-UR-83-2970 (1963). 3. F.B.Stephens, Fredi Jakob, L.P.Rigdon and J.B.Harrar, Anal.Chem. 42, 764 (1970). AC - 09.2 EXPERIMENTAL DESIGN AND STATISTICAL ANALYSIS OF DATA TO ASSIGN A UALUE TO URANIUM CONTENT IN RUBIDIUM URANIUM(IV) TRISULPHATE- A POSSIBLE STANDARD REFERENCE MATERIAL (SRM) FOR URANIUM. M.B. Yadav, Hari Singh, S. Uaidyanathan and O.O. Sood Nuclear Materials Accounting Section, Fuel Chemistry Division, Bhabha Atomic Research Centre, Trombay, Bombay-400 085. SUMMARY An experimental design based on a Randomised Block Design (RBD) was prepared to assign a value to uranium content in Rubidium Uranium(IU) Trisulphate (Key (Jords: Uranium Chemical Standard, Experimental Design and Statistical Analysis) I. INTRODUCTION Since SRMs for uranium and plutonium are not available commercially, nearly 75Og of RUS was prepared in Fuel Chemistry Division from very pure chemicals in three lots. A feasibility study was undertaken to ascertain whether this high purity material can be characterised as an SRM for uranium. As a part of this feasibility study the authors were involved in designing an experimental plan and statistical analysis and interpretation of the experimental data. II. LAYOUT OF EXPERIMENTAL DESIGN The objectives of the experiment were to study the homogenity of the lots, lot-lot variation, method-method variation, and to assign a value to uranium content in the RUS. There were three lots of RUS material and six methods coulometry 1 (C0U1), COU2, C0U3, potentiometry (POT), Amperometry (AMP), mass spectromatry (MASP) to analyse the material. Three dissolutions were prepared from each lot, In order to make three replicates of the experiment, total number of aliguots prepared were 216 ( 162 aliquots to be analysed by six methods and 48 to be kept as reserves). To make random allocation of these 216 aliguots to different methods, 216 random numbers ranging from 217 to 432 were selected from random number able, 216 labels were prepared bearing these numbers. An aliquot was weighed, a label was picked out and the aliquot vial was labelled with this label. To distinguish different lots and dissolutions two alphabetic characters, one for lot and the other for dissolution, were associated with each aliquot number. Thus before aliquoting it was not known by which method the aliquot will be analysed. AC - 10.1 -2- III. STATISTICAL LAYOUT Two way analysis of variance techniques uere used to analyse the data. The mathematical model is given by Y -p+t +B •€ , where Y - lth observation by ith treatment in the jth block, u " true mean ; t - ith treatment effect; i B - jth block effect and € - corresponding random effect. Two fold RBD was used. In the first fold, for each lot dissolutions were taken as blocks and methods as treatments. It was found that in each lot dissolution effect was insignificant. This meant all the three lots were homogeneous. Therefore in the second fold of RBD, lots themselves were taken as blocks.In each block data for the three dissolutions of the particular lot were considered. Analysis in the second fold revealed that lot-lot variation was insignificant but method-method variation was highly significant. IV. RESULTS AND DISCUSSION The two fold RBD analyses of the data showed that lots were homogeneous and lot-lot variation was insignificant. This showed that a single valuo can be assigned to the uranium content in the RUS. But method-method variation was highly significant.This divided the experimental data into three significantly different groups leading to three different values as given below: Group 1 C0U1 Mean - (34.080 + 0.043)* Group 2 AMP— -, Mean - (34.139 t 0.031)% POT-- J Group 3 COU2— n Hoart - (34.182 t 0.031)% C0U3-- J These three group-means were pooled together to give a single value to the uranium content in RUS by taking weighted mean of the three group-means, where weights are the inverses of their variances. Pooled mean for uranium content - (34.144 ± 0.020)%. Since pooling of three significantly different values is statistically not advisable, a fre%h experiment is planned to find out the reasons for the method-method variation by investigating the biases in different methods if any, analyst etc. to arrive at a more appropriate value. U. ACKNOWLEDGEMENTS Thanks are due to our colleagues in Fuel Chemistry Division for their support and cooperation in conducting the experiment and providing us with the analysis data. AC - 10.2 A TITRIMETRIC METHOD FOR THE SEQUENTIAL DETERMINATION OF THORIUM AND URANIUM Keshav Chandei, S.P.Hasilkar, A.V.Jadhav and H.C.Jain Fuel Chemistry Division, B.A.R.C., ' Bombay 400 085. SUMMARY A method for the sequential determination of thorium and uranium has been developed. In the sample solution containing thorium and uranium, thorium is first determined by complexometric titration with EDTA and then in the same solution uranium is determined by redox titration employing potentiometry. Prior to the determination of uranium, EDTA is destroyed by fuming with concentrated HCIO^. A precision and accuracy +0.5% is obtained when Th:U varies from 0.5 to 2.0 (Key Words : Sequential determination, Thorium, Uranium) I. INTRODUCTION ThC>2 - UC>2 is a promising nuclear fuel for future power- reactors due to large availability of thorium in the country' '. For the chemical quality control of the fuel, accurate determination of uranium and thorium becomes necessary. Thorium can be determined by complexometric titration .7.'. and uranium can be determined by modified Davies-Gray method' ' in a separate aliqout. Method for the sequential determination of thorium and uranium in the same aliquot is desirable to minimise the volume of analytical waste from which valuable ^^^U is to be recovered and when the availability of the sample is restricted. This paper describes the method developed for the sequential determination of thorium & uranium. II. EXPERIMENTAL Thorium and uranium solutions were prepared from high purity Th(NOj)j and UgOo. Thorium was standardised by titration against Std.EDTDTA solution. Uranium solution was standardised by potentiometric method using Fe(II) as reducing agent in H3PO4 medium employing Pt -Calomel electrode system^"" '. All reagents used were of AR/GR grade. Synthetic solutions were prepared by mixing thorium & uranium in required ratios on weight basis. III. RESULTS AND DISCUSSION Thorium at- 10 mg level was determined by the method reported^ ' earlier. The results of thorium in presence of uranium varying from 2 mg to 20 mg are presented in Table-1. It can be seen that precision and accuracy of thorium determinations are within +0.5%. Uranium upto 20 mg does not interfere in the thorium determination. Beyound this amount the intense yellow colour due to uranium masks the end point detection. For the subsequent determination of uranium (following thorium) in the same aliquot, the presence of EDTA in the solution was found to interfere in the uranium determination AC - 11.1 giving higher results. The positive bias increased from 0.4 to A% for EDTA amount ranging from 2 rng to 60 mg. Similar effect due to presence of EDTA in Fe( II)-K-jCfoOy titration has been reported^"' earlier. 1'r- eliminate'' the interference of EDTA the solution after thorium determination was fumed with HCIO^. Uranium was then determined by the modified Davies-Gray method^ ;. The results of the uranium determination are included in Table-1. It can be seen from the table that for Th:U ratio upto 1:2 the sequential determination of thorium and uranium can be performed with a precision and accuracy of +0.5%. IV. CONCLUSION The E;tudiec show that sequential determinal ion of thorium and uranium can be carried out with a precision and accuracy of better than +0.5% when thorium is at 10 mg level atid uranium upi o 20 mg. ACKNOWLEDGEMENT The authors thank Dr.D.D.Sood, Head, Fuel Chemistry Division for his constant encouragement and inspiration during this work. 1. C.Ganguly, IAEA-TECPOC 352, (1985) p 107. 2. A.I.Vogel, Text book of Quantitative Inorganic Analysis The English Language Book Society and Longmans Gre^n & Co. Ltd, 3r<1 ed. , 1961 p. 442 3. A.R.Eberle, M.W.Lerner, C. G. Goldbeck and CJ-ftodd-n, Report, MBL-252 (1970). 4. S.P.Hasilkar, N.Gopinath, Keshav Chander, £-G.Marathe and H.C.Jain, J.Radioanal. Nucl. Chern. Articles, 122(1), 69 (1988). 5. S.P.Hasiikar, Keshav Chander and S.G.Marathe, J. Radioatial. Nuol. Chern. Article 139(2), 263 (1990). TABLE 1 SEQUENTIAL DETERMINATION OF THORIUM AND URANIUM IN SYNTHETIC MIXTURE THORIUM URANIUM 5. Th/U Expec ted obtained Devi . Expected obtainsd Devi . No. (mg) (mg) (X) (rt. g) (mg) C/. t 1 . 1 . 32 9.95 9.9 4 -3. IB 5. 454 5. 466 • 0.22 1 -93 9. t,9 9. 65 0. 10 5.010 5. OOfi 0. P4 3. 1 . 98 10. 00 10. 00 •0.01 5.353 5. 059 ••0.12 4. 0.9 6 9. 7 7 9. 7 6 -0.10 ' 0 . ? 2 i0-:-2 + 0.01 j. 1.0? 18.12 10.14 < a. 2 a 9.97 9-96 -0. 10 A. 1-34 10.54 18.55 »- 0 . 0 •-> 10.17 10. 18 '0.10 7. : 0-50 I 10. 24 Iff. 26 <3\ 23 r Tiff. 5ff *e. rs 8. : 0.59 : 1 1 . 55 11.57 i a. i.' 1 9 . 5 6 • t). ! 5 9. ; 0.46 : 10.52 10. 43 •B.ja ; •"' ~) "7 "7 22. a-=; 4 11 "> c. AC - 11.2 QUANTITATIVE ESTIMATION OF PLUTONIUM IN DEGRADED ANION EXCHANGE RXSIN AND TRI BUTYL-PHOSPHATE BY NEUTRON COUNTING A.J.A.Phas, P.R.Rakshe , K.M.Michael , R.U.Yadav , V.P.Singh , K.VtJayan , N.Ramamoorthy and S.C.Kapoor Fuel Reprocessing Division, BARC. Bombay 400 085 fNDIA SUMMARY Quantitative estimation of Plutonium in degraded anion exchange resin and degraded TBP has been carried out by total neutron counting by calibrating the system in 32-1000 mg range for resin and 15-1000mg range for TBP within an accuracy of ±6 %. Spiked samples of re3in and TBP from laboratory with known quantites of Plutonium have also been estimated. (Key words : Plutonium , Degraded resin , TBP, Dowex 1X4, Total Neutron Counting , Helium (3) Counters) I. INTRODUCTION In Purex process TBP is used as an extractant for uranium and plutonium .Tailend purification of plutonium is usually carried out by anion exchange U3ing Dowex 1x4 resin /I/.After prolonged use both TBP and resin, get degraded by thermal, radiation and chemical attack and have to be replaced The discarded TBP and resin may still hold some plutonium , the estimation of which is essential before the batch is packed for safe disposal or diverted for further recovery of Plutonium. Conventional methods are generally not suitable for estimating Plutonium in such cases. Total neutron counting was considered adequate since Plutonium was isotopically well characterised Gamma spectrometry could not be employed because of hish fission product activity encountered in the degraded resin and TBP . II. EXPERIMENTAL A 250 ml calibrated glass reagent bottle was used as the standard container in which 250 ml of* Dowex 1x4 anion exchange resin was taken . For TBP a 2 litre standard glass battle was used in which 1.5 litre of 30 % TBP in Shell-Sol-T was taken The resin and TBP were separately equilibrated with different quantities of plutonium In 3tages under favourable conditions of loading . After every stage the supernatent or raffinate as the case may be was analysed from which the plutonium absorbed by the resin or TBP was estimated . The clear supernatent or raffinate was decanted , the resin or TBP in standard container was sealed in PVC bags and counted in the neutron well counter using He-3 detectors currently in use in the laboratory /2/ . At least three observations of each 1000 seconds duration were taken at each stage . The quantity of plutonium absorbed on the bed of resin varied from 32 mg to 1000 mg and in TBP from 15 mg - 1000 mg . Tha data obtained was used for calibrating the counting setup for estimating piutonium content in degraded resin and TBP under identical conditions of geometry and isotopic composition. AC - 12.1 III. RESULTS AND DISCUSSION A batch of resin was randomly checked by spiking Plutonium (33-963 mg ). Similarly a batch of TBP was checked by spiking plutonlum (101-414 mg ).Table 1 3hows the comparison of analytical results. In conclusion a method is developed for estimation ^1 Plutonium in degraded resin and TBP using total neutron counting technique . The error of estimation Is within +6 % . This method is suitable for estimating Plutonium in unknown samples under identical conditions of geometry , matrix and isotopic composition . IV. ACKNOWLEDGEMENTS The authors wish to thank Shri A.N. Prasad Director Fuel Reprocessing and Nuclear Waste Management Group, Shri M.K. Rao Head, Fuel Reprocessing Division and Dr. R.K.Dhumwad Head, Laboratory Section for their whole hearted encouragement, guidance and constructive suggestions during the course of this work. V. REFERENCES 1. J.L.Ryan and E.J.Wheelwright.The recovery, Purification and concentration of Plutonium by anion exchange in nitric acid. USAEC Report H.W.55893(1959). 2. Satyaprakash,S.P.Dange,Tarun Dutta etal Development of neutron well coincidence counter for Plutonium assay and neutron multiplicity studies. (Under publication as BARC Report.) TABLE 1 Matrix Quantity of Pu taken Pu content estimated by\ ! in the matri^t (mg) ! neutron counting (mg) • i 33.53 32.88 ! 216.65 216.75 ! Resin 614.34 646.86 963.19 944.66 ! 101.92 101.96 J 196.00 188.56 ! TBP 290.00 294.05 ! 414.60 415.32 ! AC - 12.2 EXTRACTIVE RADI0MET3IC DETERMINATION OF ORAMIUM-233 WITH 1(2- FYRIDYLAZOJ- 2 NAPHTBOL (PAN) IN TBOREX PROCESS SAMPLES K. Sreenivasa Rao.G. A. Inamdar, R. T. Kulkarni • A. Mukher .11 ,A. Ramanujam R.R.Dhumwad Fuel Reprocessing Division,B.A.R.C.BOMBAY-400 065, INDIA SUMMARY A sensitive and selective method for the determination of microgram amounts of U233 in Thorex process samples is described. It involves precipitation of U233 with PAN in alkaline medium with EDTA as the masking agent for the interfering cations and extracting it into 1,2 dichloroethane(DCE). An aliquot of the extract is planchetted for radiometric estimation of U233. Near quantitative extraction is observed with Th/U233 ratios ranging from 1000 to 2000. U233 can be estimated by the method in the range 0.5 to 50 ug.For samples of feed and raffinate composition the results agree within 2 and 10% of the expected values respectively. (Key Words: URANIUM-233, THOREX PROCESS, l-(2-PYRIDYLAZO)-2NAPHTHOL) I. INTRODUCTION: For colorimetric estimation of uranium l-(2- Pyridylazo)-2-tiaphthol (PAN) is a sensitive reagent which in alkaline solution containing strong complexing agent§ precipitates only uranium. The precipitated uranium is extracted into 1-2 dichloroethane (DCE) and the absorbance of uranium complex is determined at 560 nm/1/. PAN has also been used for the determination of uranium in thorium solutions/2/. Th^ present work extends the us% of this technique for the selective extraction of (J- 233 as PAN complex in DCE for assay by gross alpha counting or alpha spectrometry. It can be applied in THOREX process foir the analysis of U233 in irradiated thorium solution where hexone extraction of U233 from thorium is employed for this analysis. Here repeated extraction of U233 with hexone and scrubbing out thorium with aluminium nitrate salt solution results in lower U233 recovery. The present method uses a single step extraction for U233 after masking thorium with EDTA. II. RECOMMENDED PROCEDURE: An aliquot of the U-233 (0.5 ml or less), containing C.5 to 50 ug is taken. The thorium content of the solution is adjusted such that Th/U233 ratio is between 1000-2000. 1.0 mi of 0.5H EDTA is added and the pH is adjusted to 9.8 +/-0.2 (with ammonia); 1.0 ml of pH 10 (NH4CI + NH4OH) buffer is added and the volume is made to 10 ml with distilled water; 2 ml of 0.55t PAN is added, mixed and after waiting for 15 minutes, 5 ml of DCE is added and shaken for two minutes to extract U-PAN complex. The concentration of U233 in U-PAN complex extract is estimated by drying an aliqu.ot on SS planchet and by gross alpha counting. III. RESULTS AMD DISCUSSION: For the radiometric assay of 0233,a near quantitative and reproducible extraction of U233-PAN complex by DCE in a single contact is essential.The percentage extraction of U233 in presence of thorium obtained as a function of Th/U233 ratio in the aliquot (containing 14-56 ug U233) is presented in AC - 13.1 the Table.Thorium plays a vital role in enhancing the extraction of U233 PAN complex by DCE. In the absence of thorium, the U233-PAN extraction is owly about 80% and this value increases with the Th /U233 ratio and 98-100% extraction of U233 is achieved for Th/U233 ratios between 1000-2000. This is in conformity with the results reported /2/ in spectrophotometric work. The presence of EDTA lowers the extraction of U233 by competing for the metal ion. The observed enhancement in the extraction of U-PAN complex in presence of thorium could be due to the reduction in free [EDTA] in its presence. With [EDTA] levels employed here thorium upto 55mg in the aliquot could be tolerated. Extraction of thorium into solvent phase is negligible and a similar trend has been observed in the case of plutonium as well. Non interference of various cations using suitable masking agents and anions is reported/2/. The influence of nitrate ion on U-PAN extraction was checked and total nitrate content upto 0.6g/Aliq./I/ does rot interfere in the extraction. The Th/U233 ratio in the feed and raffinate samples of Thorex process vary from 1000 to 50,000 and the concentration levels of U233 also vary. The results given in the Table cover the normal range expected in feed and raffinate samples. For feed samples,it is seen that the results can be reported within +/-2% error. In the case of raffinate samples, U233 in the range of 0.5 to 2 mg/1 is to be determined in presence of large amounts of thorium.The results at this ratio and level have a positive bias but are within 10% of the expected values. This bias could probably be due to the presence of U238 in the thorium spike as impurity as it would also get extracted along with U233. For routine process control analysis of trace level U233 these errors are acceptable.The planchets prepared by drying aliquots of U-PAN-DCE phase can also be used for alpha spectrometrlc work. IV. RFERENCES •• (1 ) R. T. Kulkami , S. S. Pandit, A . Mukherji , A.Ramanujam Radiochem and Radiation Chem Symposium,Kanpur P 467 (1985) (2) K.R.Hayes, J.S.Wright ;Talanta Vol.11,607 (1964) (3) I.H.Spinner,F.C.Miller;AECL-789;Atom. Energ. Canada Ltd.(1959). TABLE Sr Th cont- U233/ U233/ Th/U ratio % extraction No. tent/Aliq Aliq taken Aliq found approx. based on alpha ( mg ) ( ug ) f ug ) counting 1 14.10 U.29 80.07 2 28.21 22.61 80. 15 3 27.38 56.42 47.96 500 85.00 4 00.55 0.56 0.56 1000 99.29 5 27.38 28.21 27. 78 1000 98.48 6 54.76 56.42 55.15 1000 97.7 5 7 27.38 14.10 14.05 2000 99.65 8 54.76 28.28 27.87 2000 98. 53 9 27.38 5.72 5.96 5000 104.20 10 27.38 2.86 3.06 10000 106.90 11 27.38 1.11 1. 20 24000 108.10 12 27.38 0.57 0.62 48000 108.90 AC - 13.2 STUDIES ON ALPHA LIQUID SCINTILLATION COUNTING FOR THE DETERMINATION OF PLUTONIUM IN SOLUTIONS R.B.Manolkar, Keshav Chander and S.G.Marathe Fuel Chemistry Division, B.A.R.C Bombay 400,085 SUMMARY Effect of some common parameters like acidity, size of the aliquot etc. on alpha liquid scintillation counting was studied. The quenching caused due to presence of high acidity and large size of the aliquot was found to be eliminated by the addition of dil. HC1, H2SO4 or HNO3. On addition of the acid, two phases were separated and decrease in acidity in the organic phase seemed to be responsible for the antiquenching. {Key words:-Flutonium, liquid scintillation counting, quenching) I.INTRODUCTION Knowledge of the concentration of plutonium in process solutions and analytical wastes becomes essential for its recovery. Assay of plutonium by alpha scintillation counting is an easy and convenient method. The solutions for counting bear various acidities and different amounts of solutions are required for the counting purpose. Studies have been carried out to see the effect of various parameters like acidity, amount of solution aliquot etc. on alpha liquid scintillation counting. Results of the studies to eliminate the effect of the above parameters with a view to develop a method for the determination of plutoniura are presented in this paper. II. EXPERIMENTAL Scintillator solution consisted of scintillation grade (Koch Light) 0.7% 2,5-Diphenyl oxazole (PPO), 0.035% 1,4 Di-[2-(5- phenyl oxazolyl)]- Benzene(POPOP), 2% Tri-n-octyl-phosphine oxide (TOPO) and 10% Naphthalene in distilled dioxane. Purified solution of plutonium in IN HNO3 (impurity content < 800 ppm) of known isotopic composition was standardised for its concentration by potentiometric method. The solution aliquots were measured on the weight basis in order 'to get precise results. 5ml of the scinti1lator solution was taken for each counting. III. KESULTS AND DISCUSSION The effect of different weights of solution aliquots and associatea nitric acid tnolarity on the counting rate was studied. The tolerence limits of these parameters were obtained. It was found that if the amount of the aliquot was more than 50rng for the solution having the acidity 7.5N or more than 100 ing for the solution having acidity more than IN of HNO3 quenching effect was observed. It is known' ' that the increase in cone. of TOPO eliminates the quenching effect due to acid when limit of acidity exceeds only to a small extent. Studies were carried out to eliminate completely the quenching effect even at higher acidity of the plutoniuia solution. It was interesting to observe that on addition of t^SO^ in the scintillation vial containing acidity and weight of the plutonium aliquot more than the tolerance AC - U.1 limit, the countratc- inc-ro-ased considerably. Inorder to find the cause of the enhancement of the countrate on addition of H9SO4 and to optimise the condition to eliminate the quenching effect completely, the effect of different amounts of I^SOA of different concentrations was studied. It was found that 0.2N H9SO4 (2- 2.5ml) is the most suitable to practically nullify the quenching effect(>9d%). The results of the studies using 0.2N H2SO4 are given in Fig.l. Similar effect was observed using E-iCl or HNOo also. By employing this method, the aliquot size as large as 500 tag in 7.5N HNO3 or lml in 4N HNO3 can be tolerated for scintillation counting. It was observed that addition of H2SO4, HC1 or HNOq caused two phases to separate out and the acidity was determined in both the phases. With the addition of the acid, the acidity was found to be carried away with aqueous phase thus decreasing the ac-idiy in the organic phase. The decreased acidity in the organic phase appeared to be responsible for eliminating the quenching effect. IV. CONCLUSION The studies carried show that plutonium can be determined 'in solution containing high acidity and in large size of the aliquot using alpha liquid scintillation counting technique. ACKNOWLEDGEMENTS The authors thank Dr. H.C.Jain, Head, Mass Spectrornetry Section and Dr.D.D.Sood, Head, Fuel Chemistry Division for their interest and constant encouragement in this work. V. REFERENCES 1. K.Joon, P.A.Deurloo and I.A.Hundere, Report, KR-100 (1965). Original amounts of Pu-solution aliquo o in 4N HNO J a 1Q0 mg * 150 mg w O 250 mg • 500 mg A 1000 mg io 1ft 20 VOLUME \r>DED (ml) Fig. 1 t Antiquenching effect of the amount of 0.2N on the alpha liquid scintillation counting. AC - 14.2 EXTRACTION CHRGMATOGRAPHIC STUDIES OF THORIUM (IV) AND ITS ANALYTICAL APPLICATIONS Uday Sankar Ray and Kalyanmoy Mitra Department of Chemistry,Visva-Bharati, Santiniketan - 731 235, India. SUMMARY k rapid and selective method has been proposed for extraction and separation of Th(IV) by reversed phase extraction chromatography with capric acid coated on silica gel. Quantitative extraction has been achieved in the pH range 4.0 - 5.0. Th(IV) has been separated from variety of metal ions present in mixture by exploi- ting the difference in extractability and stripping behaviour. Key word t Extraction Chromatography, Capric Acid, Silica gel, Exchange Capacity, Thoron. I. INTRODUCTION High molecular weight <"*rboxylic acids have been widely used for solvent extraction of metal ions. Amberlite LA-1, Amberlite LA-2, TOPO/1/ has been used for the reversed phase extraction chromato- graphic separation of metals. In this paper we have described the reversed phase extraction chromatographic studies of Th(IV) with capric acid and its selective separation. II. EXPERIMENTAL Digital pH meter (Elico-LI 120, India), spectrophotometer CECIL, India) and chromatographic column were used. Capric acid was used as liquid cation exchanger. Solution of Th(IV) was prepared from Th(N03>4. 5HgO and standardised complexometrically and diluted to 50 >ug ml*1. Silica gel was rendered hydrophobic by dimethyl dicnlorosilane and it was coated with capric acid in benzene in a ro+ary vacuum evaporator /2/. 2.5 g of gel was slurried with water to make 1 x 7 cm column. The column was washed with 4 ftHC1 . Procedure : Solution containing 50 /Ug of Th(IV) In 0.2 J£ acotic acid at pH 4.0 was passed through the column at a flow rate of 1 ml rain"1. Th(IV) was extracted with capric acid and stripped with 0.2 M, HNO3. Six fractions each of 5 ml was collected and analysed spectrophotometrically with thoron at 545 nra. III. RESULTS AND DISCUSSIONS The extraction behaviour of Th(IV) with capric acid showed that the quantitative extraction was achieved in the pH range 4.0 - 5,0 from 0.2 M acetic acid solution. Thorium(IV) was quantitatively stripped with ^O.lU HN03 ^ 0.2 ft HCl and ^ 0.2 ft H2S04. AC - 15.1 Capacity s The exchangge capacitpy y was determined by shaking 50 rcll of B]J NaCC l with 1 g exchanger for 10 hrs in hot cum cold bath au1 titrating 25 ml solution with standard alkali The exchange capacity was found to be 2.04 meq of H /g at 25°C. Separation of Th (IV) from binary and multicomponent mixtures i Th(lV) was separated from several binary ani multicomponent mix- tures by exploiting the differences In pH for extraction and elution. Metal ions not extracted from 0.2 £ acetic acid solu- tion passed through the column at pH 4.5. Th(IV) was separated from them by stripping the column with 0.2 M HN03. Th(IV) was separated from Cr(III), Mn(II), Co(II), Ni(Il), Zn(II), and Cd(II) by this process. Slements like Pb(II), 11^(11), 31(111), La(III), Ce(IV), Pr(III), Nd(III), Sm(III), Zr(lV) and U (VI) were extracted with Th(IV). The separation of Th(IV) from them was achieved by passing the mixture at pH 4,5. After extraction Pb('ll) was stripped with 0.002 & E1W3 » HgUl) with 0.01 U HWO3 Bi(III) with 1.5 iJ HNO3 , rare earth elements were stripped with a mixture of NaN03 ( 1JU ) and HNO3 ( .01 U ), U(VI) with 0.005 M HHO3 and Zr(IVJ with 4 M HN03 ; while Th(IV) was reco- vered with 0..2 iJ HNO3. Th(IV) was separated from several mix- tures containing metal Ions commonly associated with Th(IV). The results are given In Table - l. Table - 1 1 Separation of Th(IV) from multicomponent mixture. Mixture Th( IV) (Recovered ^T" 1. Th(IV) + Ce(IV) + Zjt(IV) 50.5 100.4 2. Th(IV) + Ce(IV) + U (VI) 50.4 100.2 3. Th(IV) + Zr(IV) + U (VI) 49.9 99.2 4. Th(IV) + Ce(IV) + 2r(IV) + U (VI) 49.7 93.3 Thorium taken = 50.3 /Ug { Flow rate = 1 ml min""^ ; pH = 4.5. IV., REFERENCES 1. E. Cerral and C. Testa } J. Chromatog., 9 , 215 (1962). 2, S.M.Bhosale and S.M.Khopkar ; Talanta, ^6 , 889 (1979). AC - 15.2 SIMULTANEOUS MASS SPECTROMETRIC ANALYSIS OF URANIUM ANT) PLUTONIUM FOR DETERMINATION OF CONCENTRATION OF U AND Pu IN DISSOLVER SOLUTION OF SPENT FUEL S.A.Chitarcbar, P.S Khodade, A.R.Parah and H.C.Jain Fuel Chemistry Division, B.A.R.C., Bombay 400 08'). y A method involving direct loading of spiked solution of dissolver solution for determination of concentration of U and Pu without chemicH separation is developed using a variable millticol 1 Ret or system on a mass spectrometer. (Key words: Dissoiver solution, U, Pu, Single faraday cup, Variable, multicol lector mass' spectrometer) I. INTRODUCTION Determination of U and Pu in dissolver solution of spent fuel is one of the key measurements in the nuclear fuel cycle. Isotope dilution-Thermal ionisation mass spectrometry fID-TIMS] technique is widely used for precise and accurate determination of U and Pu at the input stage of Pu reprocessing plant. These methods normally require chemical separation of U and Pu/1/. Sequential isotopic analysis of U and Pu using a) resin bead technique /2,3/ and b) a purified fraction of solution of U and Pu for single loading on sample filament have been reported in literature/4/. Some observations for simultaneous analysis of U and Pu have been reported in recent past from our laboratory/5/, yet no analytical data is reported in literature. The procedures standardised for simultaneous mass spectrometric analysis of U and Pu using a) Secondary electron Multiplier (SEM) detector, b) Single faraday cup as collector and also c) Variable multicollector system on MAT261 thermal ionisation mass spectrometer, are described in present paper. II. EXPERIMENTAL Experiments have been carried out on solution of synthetic mixtures (SM) of U and Pu having different U/P>u ratio, to establish mass spectrometric operating conditions. U and Pu were used as spikes for U and Pu in ID- TIMS technique. For SM-1 solution, SEM was used as detector, while for SM-2 and SM-3, single faraday cup on MAT261 mass spectrometer with peak jumping mode for simultaneous analysis of U and Pu was used. A variable multicollector system with nine faraday cups and a SEM, recently installed on MAT261 was used for the analysis of dissolver solution of spent fuel. For SM- 1 solution, SEH detgctor system was essentially used to monitor the ion intensities of U , Pu , UO , PuO and UO2 during the analysis at low as well as high sample filament currents. The ratio UO /U was found to be greater than A at a sample filament current of 1.6-1.8A, with no significant peaks of PuO and UO2. This observation was used to standardise pre-heating at position A and B of sample and ionisotion filanents at currents 1.6 A and 5.6A respectively for 10 minutes. The analysis of U and Pu was carried out at sample and ionisation filament currents of 2.6A and 5.8A with intensity of 5V and 0.4V for U and Pu respectively. The isotope ratios from spiked sample, obtained by analysis of direct loading of a small fraction, were used for calculation of U and Pu concentration. Results obtained on these measurements for different solutions are given in Table-1. AC - 16.1 III. RESULT AND DISCUSSION From the U and Pvi concentration values on different solutions given in Table 1, we observe that a percision of better than 0.2X. is achieved for SM-1 and SM-2, while for Pu in SM-3 is having a percision of 0.8*. All these analyses have been carried out using a peak jumping mode requiring 25-30 mins time for a block of 11 scans. This introduces the errors from time dependent variations in the ion beam. During acquisition of data, this error is eliminated in the results obtained from dissolver solution analysed by employing variable multicollector- system. The present work, thus demonstrates a method for simultaneous determination of U and Pu concentration in a solution having U/Pit ratio of 1500. The precision of better than O.lt is achieved for U and Pu values. Because of high level of radioactivity on the sample filament due to the loading of solutions ot spent fuel with high burnup will cause fieavy contamination in the ion source, the present method should be used on selected samples to cross check values obtained using usual ID-TIMS procedures. The results on SM-1 and SM-2 will improve further as is the case of SM-4 with m.-iticollector system for simultaneous analysis of U and Pu. It can be concluded that the purified Pu solution can be directly analysed for U and Pu. This can be seen from data on SM-1 and SM-2, having U/Pu ratio as 4 and 25, usually present in plutonium eluted with 0.35M nitric acid in the anion exchange separation procedures for samples from spent fuels of power and research reactors respectively. IV. ACKNOWLEDGMENT Authors are grateful to Dr.D.D.Sood, Head, Fuel Chemistry Division, B.A.R.C. for his keen interest in the work. V . REFERENCES 1. S.A.Chitambar et.al. B.A.R.C.-Report, BARC-865(1976) 2. J.A.Carter et.al. IAEA-SM-201/9, p-461(1976) 3. S.A.Chitambar et.al. Second Nat. Symp. on Mass Spectroaetry, Dec21-23 1981, BARC, Bombay paper No.P-7(1981) 4. A.D.Moorthy et.al. Third Nat.Symp. on Mass Spectro»etry, Sept22-24,1985 R.R.L. Hyderabad, paper no. B-8(1985). 5. G.Chourasiya et.al. Preprint vol of Symp. on Radio chew & Rad. Che». Jan. 4-7, 1989, Kalpakkam Paper RA-32 (1989) Table 1 Results of ID-TIMS experiments for simultaneous determination of U and Pu Saaple Detector or Concentration U/Pu No. collector used U(^g/g) PuC/Jg/g) ratio SM-1 SEK 6.939±.004 1.89741.0017 3.7 SM-2 Faraday(Single cup) 92.10 ±.10 3.71 ±.01 24.8 SM-3 Faraday(Single cup) 354.2 ±.4 0.251 ±.002 1411.0 D.S.« Variable Faraday cups(9) 13541.0 ±10 8.823 ±.004 1535.0 * D.S.=Dissolver Solution of spent fuel. AC - 16.2 DKTEKMIMATION OF N1TKOCJKN, VLUOHIHK AND CULOU1NK IN U^(.'e I'UKi. M&TKUJLALti USING tilJAKK ISOUKUK MAUiS iil'ltOTKOWiTU* K. L.Kamakuinar, V .A.Kaman, V. L.iJant, V.l).K.avimaridan and H.C.Jain IJuel Chemistry Division, U.A.K.C., Bombay 4UU Obb, India The multieiemental analysis capability of opark oouroe Mass Spectrometry (SbMii) has been exploited tor the determination ol gaseous impurities nitrogen, fluorine and chlorine in U^Oo iuel materials. The ubiquitous nature oi nitrogen and oxygen made it. imperative to have the experiments carried out at pressures less than lxlu turr so as to minimise the blank. The gases N-, I1 and (Jl were determined assuming a unit relative sensitivity iaoUi. (KEYWORDS : Spark source mass spectrometry, Analysis o.r gases, photoplate detection, U^u-j matrix; INTRODUCTION Spark source mass spectrometry EXPEH1MKNTAL The experiments were carried uul employing a JHb uibH-i; double focusing rr spark source inhss speccivmotor with an auU spaik uontruij ing unit. 1'lie instrument has both electric detection as well as photoplatc detection iacilities i»vaiiabl«. Tli-a cltstalii uf the instrument have been given elsewhere,'1/. Nuclear grade U^Uy (2UU mesh) wat. mixed with high purity graphite in J:l ratio by weight in order to make it conducting for sparking. This 1:1 ratio ot sample to graphite was found most suitable for preparing strong and compact electrodes which could sustain sparkig without breaking. The mixture was hortiogenolsed for about half an hour and moulded into cylindrical electrodes (. lumm^i'in) using a hydraulic preS3 operating at a pressure of lit. Kg/Cm*' applied for about ZO minutes. $lth a view to eliminating isobaric interferences due u^iii1^ oi jbFe 4 present if any at 14N+, the experiments were carried out using t>hotoplate detection system at an operating practical resolution of about 2UUU. This resolution, was mpre than adequate to resolve the mass number 14Nf f ron/bGi^ ^ and ^i'c'4 ibil'uKD w-2 photoplates (35Ummx5Ummx linni) ware used tor tbis purpose. exposures ranging from lxlu *C to lxlu IK^C woie taken on different stages so as to cover the entire concentration, range or the impurity as Well as the matrix element. Th-r phoi-wj-late was AC - 17.1 developed, evaluated employing Franzen et ai. method/ii/ and the concentration in ppmw was caiouiated using the equation- ^At.Wt/, MY-^ whero C. is the concentration ol i,he element ii» nd are the exposure values tor the matrix and the impurity required rur tlie same degree or darXenljig, A&3 are the atom %> abundances tor the matrix aiid the impurity clement isotopes monitored, sftt.Wt->& are the average atomic weights,and the terms involving the nias3 numbers are the correction laoturs to account for the n;ass response or the photopi&te emulsion and the line width. RESULTS AMU DISCUSSION The concentration oi N-, F and Ci determined in U^O^ is given in Table 1. The results shown in the table are the average oi the data obtained on live ditrerent photopiatea. It should bo notod that no calibration factor in the torm ot relative sensitivity factor IRSF) to account for the deviation trom the true value has been applied. The data are given assuming a unit calibration factor.With a view to estimating the minimum value oi N.. that can be obtained under the experimental conditions, a semiconductoi grade tellurium sample was analysed and the concentration was found to beabout 0.3 ppmw. in view ot tills, no correction was deemed necessary for N^. in ,... Future work is planned to prepare the trace constituent standards ol U -y l«_ ACKNOWLEDGEMENTS The authors are grateful to Dr. D. D. tiood, Head, Fuel Chemistry Division tor his keen interest. They also thank olu l J.K.. oaiuuel and his colleagues in instrumentation group, liadiochemistry Division. REFERENCES 1. K. L. liaiuakumar et ai . , Fre&enius L . Ana 1. Chem. , iiii Ii: 2. B.P.Datta et al., int. J. Mass Spootrom. Ion l £4 319,198b. 1. Concentration of N2, V and Cl in UyOy (ppmw of U) Element Concentration U ) L.O ± 1.6 14.0 1 4.L AC - 17.2 QUANTITATIVE DlCnauaJUTIO* OF DO* If UlMJOt HIXTV&E BY Z-1AT DlfFRACTIOU KETBOD K.B. Khtn, C.C. J«in and C. Ganguly Radiometallurgy Division Bhabha Atoaic Research Centre Sonbay 400 085 SUMHAKT A direct comparison X-ray diffraction method for determination of UO* in UN- UOi aixture has been standardised over the range of 4 to 10 w/o of UOt. The results obtained by this aethod ai> in flood agreement with those of chemical analysis. The accuracy of the method is id w/o (absolute) of UOt. KEYWORDS X-ray diffraction, Oraaiua aononitride, uranium dioxide, K-factor, Inert gas fusion. IIITEODOCTIOM Uraniua nononitride is considered to be a proaising nuclear fuel because of its attractive high temperature properties and high density as compared to uraniua dioxide. Carbothermic nitriding of U02 is the aost widely jised method for preparation of oononitride. Ihe amount of VOz present in the powder after carbotheraic synthesis of UN deteraines the extent cf reaction. Since UN has very Halted oxygen solubility (<0.1 w/o ), aost of the oxygen present in the system during nitriding exists in the form of unreacted oxide. Hence the results obtained by chemical analysis of oxygen in im-VOx fixtures can be correlated with UO* contents analysed by X-ray diffraction (XRD) nethod. Thus, XRD which ia a rapid non-destructive aethod of analysis could be effectively utilised for evaluation for oxygen content in UN samples prepared by carbotheraic nitriding of oxide. The paper presents quantitative determination of UOs content in UN- UOx mixture by X-ray diffraction and its correlation with results based on chemical analysis. THEORY The integrated intensity Ii of (hkl) plane of phase X is related to its volua* fractior-.C* by the following relationship*l> hkl hkl Is - c . Kx . Cs (1) where c is a constant depending on experimental parameters as well as instrument, and the K* k is given by kkl ~kl K* « j FK p. P. LP. (T/V») A (2) where Fxfcfci is structure factor of phase X for (hkli plane, 1 is •ultiplicity factor, LP is Lorentz polarisation factor, T is teaperature factor, A is absorption factor, and V is volume of the unit cell of phaso X, Is tfaie paper direct comparison method <*<3> has been followed for quantitative phase analysis in UN-UOs systea by theoretically calculating the K viluea of selected Unas of UN and U02. Following relationship has been derived theoretically for quantitative phase analysis in UN-UOZ aixture froa the reported X-ray cryetallographic data <*••> for aN and UOa. : w/o of UOs - [ 1 - l/U+(0.6812S3)(IuoV/Ioin] x 100 (3) KXPSKIMKBTAL The samples fot XKD analysis were prepared in the fora of compacted 30 tut diameter diskettes. X-ray diffraction patterns were taken with CuKa radiation, { 40 kV and 20 aA ) by continuous scanning at the rate of 1 froa 26 *>25° to 38°. The intensity data thus obtained were subjected AC - 18.1 to background correction and the integrated intensities were computed. A minimus of five test samples of UN containing UOt in the range of 4 to 10 w/o were prepared by aechanically Mixing pure UOi and pure UN powders,for verifying the theoretically deduced equation(3). The oxygen content* of some UN fabrication batches wera determined by inert gas fusion method**'. RESULTS AMD DISCUSSION The results of XRD phase analysis of the test samples,presented in Tablel show good agreement between the expected UOz content and the values obtained fro* equationO). The deviation of the values is within t 1 w/o of UO,. A large number of UN samples from different batches have been analysed. Soae of these results have been presented in Table 2. There is a good agreement between the values of UOt contents derived from chemical analysis (Inert gas- fusion) results acd those froa XRD results. It can be concluded from this wotk that UOn content in UN-UOs mixture can be determined by X-ray diffraction aethod with an accuracy of ±1 w/o in the range c* 4 to 10 w/o of UO». &EFKRENCSS 1. B.0. Cullity, Elements of X-ray diffraction. Addisoa Wesley publ. Co. 2. C. Gauguly aud D. Vollr.tfc, KFK-2049. 3. R. Conti, C.J. et al., Anal. Chea. Acta 41, (1968) 83 4. Inter. Tbl. of X-ray crystallography, Kynoch Press, Birmingham. 5. Pearson's Handbook of crystallographic data tor internetallic phases, vol.3 P. Villars and L.D. Calvert., ASM , 1985. 6. G.C. Jain and C. Ganguly., Proc. DAE Synp. Nucl. Chem. k Rad. Chem., Varanasi., 1981 Table 1: XRD results of test samples Sample U0? content (w/o) expected XRD difference A. 4.025 3.700 -0.325 B. 8.981 9.360 +0.379 C. 7.010 6.865 -0.145 D. 6.020 6.192 +0.172 E. 4.980 4.927 -0.053 Table 2 : XRD results of samples drawn from UN batches Sample Oxygen UOi content (w/o) (ppm) calculated from XRD difference oxygen content S-15/A 5197 4.384 3.857 -0.527 S-15/B 5609 4.732 4.539 -0.193 S-17-A 3991 3.367 3.362 -0.005 S-17-B 6127 5.169 5.242 +0.073 S-20-B 5213 4.400 5.280 +0.880 S-21-C 6933 5.849 5.674 -0.175 S-21-D 7597 6.410 6.687 +0.277 AC - 18.2 AUTHOR INDEX AUTHOR INDEX Abraham J.V. SSC-B5 Chaudhuri N.K. AC-05 Aqarwal R.K. SSC-21 Chauqule G.A. AC-06 Aqqarwal S.K. NC-a? Chawla K.L. SSC-31 Amalroj R.V. SST-I5 Chen S.Y. NC-05 Anbe F. NC-BS Chetty K.V. SST-05 Ambe 5. Nc-aei Ch)nnusamy A. SST-13,SST-14 Ambulkar M.N. AR-ll,AR-43 Chitanbar S.A. AC-U.NC-02 Ami am A.M. AR-B5 ChuU.s N.L. AtV40 Ani i Kunar SST-819 A n k 1 a m E . RC-0/ Dahale N.D. SSC-31 Antonysem/ S. SSC-36 Dange S.P. NC-01 Apte R.6. NC-04 Daniels E.A. AR-30 Aravindaku«ar C .T RC-17 Das A. AR-44 Arpita K. RC-13 Das N.R. AR-04,AR-lS,AR-26 Asmus K.D. RC-87' * AR-46,RC-31 Aurthur Fry AR-2'I Das S.K. AR-56,NC-S6 Dash K.C. A^-03,AF:-10 Badhefca L.P. SSC-01 Date D.B. RC-26 Bajpai D. D,. SST-B8 Datta T. IT-29,NC-81.,NC-03 Bajpai D.D. AC-08 David F. IT-08 Bajpai D.D. SST-12 De Amitava RC-31 Balasubramani an AR-36,lT-19,NT-03, Dedqaonkar V.G. MC-04,RC-39,RC-4B G.R. SST-13,SST-14,SST-18 Dehadraya J.V. SSC-30 Bandyopadhyay T,, RC-01 Cssai C.N. IT-30 Baner;i A. SSC-01 Desai G.S. SSC-10 Bangia T.R. AC-01 Deshmukh S.M. AE-03,AR-U Bapat L. RC-36 Deshpande S.G. SSC-22 B a 5 u D. AR-46 Devarajan K.C. SST-18 Basu Roy M. RC-19 Dey. G.R. RC-22 Basu S. AR-01,AR-25 Dhabolkar G.M. SSC-24 Beauvy M. IT-06 Dharai P.S. SST-07 Bhaqwat D.A. NC-04 Dhanya S. RC-09 Bhat G.S. SSC-07 Dharkas S.P. SST-12 Bhat I.S. AE-01 Dharmpuri kar G.R. SSC-24 Bhatia D.S. AR-06 Dhas A.J.A. AC-12 Bhatta D. RC-33,RC-34 Dhawale B.A. AC-01 Bhattacharyya D.K AR-18,AR-26 Dhumwad R.K. AC-13.SST-B7 Bhattacharyya P.K RC-09,RC-25 Dongre N.S. SSC-07 Bhattacharyya AR-46,RC-J8,RC-19 Dutta N.C. &R-lB,AR-26 S.N RC-2i31RC-21,RC-31 Dwivedi K.K. NT-02 Bhutan! M.M. AR-12,AR-28,AR-29 Bodas M.O. RC-38 Gabriel J. SST-10 Bprhade «A!. AR-13 Ganesan V. SSC-37 Genga;ah T. AR-35 Capdevila H. IT-12 Ganguly C. AC-18, IT-13 Ch^eita K. RC-28 Sarg A.N, AR-ll,AR-39,AR-4B Chacharkar M.P. AR-37 AR-41,AR-43,NC-B5 Chaddha A.K. NC-03 RC-29 Chakrabarti S. RC-18 George Thomas AC-08 Chakravarty N. NC-09 Ghadse D.R. SS7-01 Chakravortty V. AR-03,AR-i0,gSC-B8 Ghosh S. NT-02 Chandran K. SSC-37 Godbole A.G. SST-05 Changdar S.N. AR-4 4 Gopalakriehnan V. SST-07 Chsryulu M.I1. SSC-34,S6T-Bl,eST-06 Gopi nathar. C. RC-02.RC-B3 Chatterjee A. AR-0J Go&wami A. NC-08 AI - 1 Govindan P. SST-13,SST-H Kelkar S.S. RC-38 Goyal N AC-03 K*ch«v Chandtr AC-11,AC-14 Gubbi G.K. AC-02 Kctkar M. SSC-32 Giii 11aumont R. IT-22 Khan K.B. AC-18 Supta K.V.. Kh»n 2.ft. ftR-32 Gurba P.B. AC-06,SST-BB,SST-12 Khodade P.S., AC-1A Gurin Y. 1T-02 Knight G.S. AR-48 Konashi K. IT-15 Haqenlocher J. IT-28 Korah Prince C. RC-32 Hareendran K,N. SSC-0S,SST-B4 Korde Aruna AR-21 Hari Mohan RC-07,RC-88,RC-12 Kothari K.K. AR-19 Hari Singh AC-IB Krishnanoorthy K.R. AR-38 Harifh Chand»r AR-23 Kshirrtaqar T.V.S.P. AR-45.AR 4/ Hasilkar S.P. AC-11 Kulkarni R.I. AC. IS Hill B.I . AR-43 Kulyako Vu.M. SSL 3 a Kumar S.V. lla P. AR-47 Kumar aguru V.. SSC .'4 Inamdar G.A. AC-13 Kumra M.S. IT ii Inqle M.N. RC-23 Kuriacoss J.C RC 32 Iwamoco N. NC-05 Iyer R.H. NC-B7 Labhasetwar N.K. AE-02 Iyer R.K. AR-38,SST-02 Lahiri S. AR-04 Iyer V.S. SSC-25,SSC-27 Lai K.B. SST-15 LarjeMar R.B. AR-40 Jadhav A.V. AC-11,SSC-B2 Lian C.T. Shbutt RC-11 Jadhav S,D. AR-05 Lohithakshan K.V. SSC-02 Jai Prakash SSC-21 Lokhande R.S. RC-23,RC-38 Jain G.C. AC-18 Jain H.C. AC-11,AC-16.AC-17 Nahajan G.R. GSC-12,SSC-2B IT-26,NC-02,SSC-B2 Hahajan M.A. AC-35 SSC-23,SSC-34,SST-B3 Mahal H.S. RC-65 Jain S.C. SSC-22 Maity D.K. RC-16 Jayadevan H.C. S5C-29,SSC-31,SSC-32 Malleswar Riddy A. SSC-09 Jayakumar V. AR-23 Manchanda V.K. SSC-17.SSC-18 Jha S.K. RC-35 Mandal P.C. RC-18,RC-19,RC-2fc Joshi A.R. SSC-03SSC-11,BST-B1 RC-21 Joshi N.G. RC-29 Kanmohan Kunar RC-06 Jyotsna N. AR-50 Hanohar Lai RC-05 Katiohar S.B. NC-08 Kadam A.V. SST-01 Hanolkar R.B. AC-B9;AC-14 Kalkar CD. RC-26,RC-3B,RC-37 tfapara P.M. SST-05 Kailiappan !. SSC-35 narathe S.G. AC-09,AC-14 KaU). P.C. NC-07 Marian £. AR-06 K^sal Kishora RC-10.RC-22 Mary Xavier SST-B3 Kansde K.G. AR-49 Hath-kar A.R. AR-17 Kanw^r Raj AC-07,SST-11 Hit hew K.A. SST-03 Kapadi> 0. fiR-07 Mathews C.K. IT-14 Kapoor S.C. AC-12.SBC-24 Mathur J.N. SSC-01 Kapoor S.K. RC-03 Matkar V.M. AE-01 Kasar U.M. SSC-03,SSC-li,SST-8l Michael K.N. AC-12.SSC-24 Kasar O.M. SST-01 Miihra G.K. AR-08,AR-i9,AR-27 Kaushik C.P. AC-07.&ST-11 Mithra P.K. AR-03 Kiuetiik C.P. SST-il Miihra S. RC-33 Kavimandan V.C AC-17 Kedari C.S. SSC-16 AI - 2 Mishra S.P. AR-7*4,NT-81,RC-35 Padatanabhan P.K . AR-15 RC-4i,RC-42,RC-43 Page A.6. AC-B3 RC-44-RC-45,RC-46 Pai 5.A. SSC-19 Mithapara P.D. SSC-34 Pal H. RC-15 Mithlesh Kunar SSC-26 Palanalai A. SST-13,SST-14 Mitra A.K. AR-12,AR-28,AR-29 Palit D.K. RC-15 Mitra K. AC-15 Pandey A.K. NC-07 Mitra S. AR-48 Pant H.J. AR-33 Mittal J.P. RC-12,RC-13,RC-15 Parab A.R. AC-16.NC-B2 RC-17 Parthasarat>iy R . AR-47 Mohan S.V. SST-13,SST-14 Pathare M.N. SSC-07 Mohanty R.N. AR-10 Patil B.M. AC-08 Mohapatra P.K, SSC-17,SSC-18 Patil B.N, SST-B3 Moorthy A.D. AC-06 Patil R.M. RC-27 Moorthy P.N. RC-04,RC-I0,RC-22 Patil S.r. AR-13RC-27,RC-2B Mudaliar M. RC-12 Patil S.K. IT-11 Mukerjee S.K. SSC-38 Patil V.B. AR-07 Mukherjee T. RC-15 Patil V.J. RC-30 Mukherji A. AC-13 Pawar M.B. AR-43 Muller H. IT-28 Pawar S.M. SSC-03,SSC-n,BST-»l Murali M.S. SSC-B1 Pawar S.S. RC-28 Muralidharan P. SSC-37 Pawaskar D.B. AR-47 Murty G.S. RC-35 Perevalov S.A. SSC-08 Murty G.S. AR-15 Periaswami G. SSC-37 Musikas C. IT-li Phal D.G. SSC-06 Myasoedov B.F. IT-07 Phulkar S. RC-14 Pikaev A.K. IT-2/ Naqatnal1eswararao B. AR-45,AR-47 Pillai K.C. AE-01,IT-23 Naqar M.S. SSC-13.SSC-15 Pillsi M.ft.A. AR-19.AR-22 Naidu G.R.K. AR-35 Pinole R.T. RC-40 Naidu R.T. AR-45 Pintu Sen ST-10 Naik D.B. RC-04 Pius I.C. SSC-04,SST-ai,SST-a6 Naik H. NC-B1 Plank L.«D. AR-48 Nair A.G.C. AR-16 Porwal N.K. SSC-26 Nair G.M. SSC-i2,SSC-2B Prabhu D.R. SSC-12,SSC-2fl Hair M.K.T. AC-0o,AC-B8.SST-B8 Prssad R. SSC-25.SSC-27 BST-09,SST-1B,SST-12 prashant Dutt SSC-21 Nair P.R. SST-03 Pujar M.A. SSC-B9 Nair P.S. SSC-02 Pujs.-i P.K. NC-03 Naito K. IT-04 Purohit P.J. AC-B3 Narasimha Moorthy B. SSC-28 Pushparaja SSC-05,SST-B4 Narayanan C.V. SSC-24 Narayanan U. AE-01 Rail Anted A.6. AR-3/>,NT-83 Naronha D.M. SST-01 Raghupathy S. AR-02 Natarajan P.R. AC-01,AC-a2,RC-24 Raghuraman K. SSC-B2 SSC-B1 Raj an S.K. SST-13 Natarajan R. SST-18 ftajandr.a Kunar SST-08 Natu G.N. RC-36 Rajendra Swarup SST-B5 Nawada H.P. SSC-35,8SC-36 Rajesh N. AR-42 Neeta Lala RC-3B,RC-37 RajssMari B. AC-01 Newton Nathaniel T. SSC-33 Rajput D.U. RC-39.RC-4B Raju J. NT-«I2 Obergfell P. IT-28 Rajurkar N.S. AR-13.AR-49 Ohkubo Y. NC-05 Rakshp P.R. AC-1.2 Ram K.D. AR-:U AI - 3 Ram L.C. NT-0!,RC-35 Satya Brat AR-32 Rama Rao 6.A. SSC-30 Satya Prakash AC-02,AE-03,AR-16 Ramadevi P. AR-35 IT-10,NC-01,NC-03, Ramakri shna M.J. AR-45 NC-0ii>,NC-08,NC-B9 Ramakrishna V.V. SSC-06 Sawant A.D. SSC-07 f'^makrishnan V. RC-32 Saxena M.K. SSC-23 Ramakumar K.L. AC-17,SS:-23 Seenivasan 6. SSC-35,SSC-36 fiamamoorthy N. SSC-24 Shanbhaq B.S. AR-14 RaraamoDrthy N. AC-12 Sharma H.S. AC-09 Ramarourty C.K. SSC-28 Sharma R.C. NC-07 Ramamurthy T. V. AR-23,AR-24 Shastri L.V. RC-13 Raman V.A. AC-17,SSC-23 Shinde V.M. SSC-10 Raman V.R. SST-!3,SST-14,SST-18 Shivarudrappa V. SSC-34 Ram an u jam A. AC-13,IT-2B,SST-87 Sf-.ri vastava O.P. AE-02 Ramaswami A. AC-02,AE-03,NC-06, Shukla J.P. SSC-13,SSC-16,S3C-2B NC-09 SST-09 Ramesh Kumari AR-J2,AR-28,AR-29 Shulka N.K. SSC-25,SSC-27 Ramji Lai AR-21 Shukla V.D SSC-24 Raronani S.P. RC-09 Siddappa K. SSC-09 Rao B.S.M. RC-12,RC-!4,RC-17 Singh D.N. RC-39 Rao M.H. RC-06 Si nqh J. RC-41 Rao S.V.S. SST-15 Singh Mudher K.D SSC-29 Rao V.R.3. RC-32 Singh N.P. IT-24 Rastoqi R.K. AC-05 Singh N.S.B. AR-36,NT-03 Rathinaswamy A. AR-51 Singh R.J. AC-02,AE-03,NC--0<> Rattan S.S. AE-03,NC-06,NC-09 NC-09 Raut V.M. RC-30 Singh R.K. AC-06,AC-08,SST-08 Ravi Sankar A. NT-03 SST-09,SST-12 Ravi T.N. SST-18 Singh R.N. AR-08,AR-09.AR-27 Ravi ndran P.V. SSC-05 Singh R.R. SST-12 F Unni P.R. AR-17 Upedhyay* V. AR-34 Vaidya N.D. AR-17 Vaidya V.N. SSC-27,SSC-38 Vaidyanathan S. Ac-ia Var 4\i»r • jan N. AC-BD,AC-fl8,8ST-aB SST-1B Vasudeva Rao P.R. IT-18,SST-16,SST-17 Veeraraqhavan R. SSC-14 Venkatasubramanian R. AC-B4 Venkateean M. SST-18 Venkate&h M. AR-20,AR-21 Venuqopal A.K. SST-12 Venuqopel V. SSC-25,SSC-27 Vijayan K. AC-12 Vitorqe P. IT-12 Weginwar R.S. AR-40,AR-41 Uestrun E.F.Jr. IT-(81 Wrenn Mc.E. IT-24 yadav M.B. AC-IB Yadav R.B. SSC-28 Vadav R.U AC-12 Yanaviaki n. IT-15 Yeotikar R.6. AC-B7,SST-11 Yeotikar R.G. SST-11 Zaman M.R. RC-43 Zvara I. IT-B5 AI - 5i 1 'J ^ . Also tlie values of br^nu-ll j ch oi.-rf.stcmts f-'-i Jii-Ti/^aii'r.O'i for b---t.lt the 5-:'il.x, are greater than thvs^ for '.rlii- omi nm iiid i • •-.« t Liijg that. m-nn^'iies'.- will be m--IVJ :j...r-l •••••1 •.!> t.Jie / • • • i Is •;• inifi.'.red to .•iirv.'iii ium. ['r-:>\jZibly. it i £• -JII-.-J t •.• i-.ui'j .iiz.-^ anj i--ni.j t>'..i. en tin 1 ' '. The aroa •-f s t. n,l j. •.-, r-^f-r 1 ed t... in the-