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FUEL CHEMISTRY DIVISION ANNUAL PROGRESS REPORT FOR 1988
S. Vaidyanathan
1991 GOVERNMENT OF INDIA ATOMIC ENERGY COMMISSION
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FUEL CHEMISTRY DIVISION ANNUAL PROGRESS REPORT FOR 1908 EdiLecl by S. VaidyanaLhan
BHADHA ATOMIC RLSLAROI CLN7KI HOMHAY, INDIA 199 I BARC/1991/P/002
BIBLIOGRAPHIC DESCRIPTION SHEET FOR TECHNICAL REPORT (as per IS : 9400 - 1980)
01 Security classification : Uncl assi-f ied
02 Distribution : External
03 Report status : New
04 Series : BARC External
05 Report type : Progress Report
06 Report ND. : BARC/1991/P/002
07 Part No. or Volume No. :
08 Contract No. ;
10 Title and subtitle : Fuel Chemistry Division : annual progress report for 19BB
11 Collation : 158 p., 61 tabs., 13 figs. 13 Project No. : 20 Personal author(s) : S. Vaidyanathan 21 Affiliation Df author(s) :Fuel Chemistry Division , Bhabha Atomic Research Centre, Bombay 22 Corporate author(s) : Bhabha Atomic Research Centre, Bombay - 400 0B5 23 Originating unit : Fuel Chemistry Division, BARC, BDHIIJU 24 Sponsor(s) Name ; Department of Atomic Energy Type : Government 30 Date of submission : July 1991 31 Publication/Issue date : August 1991 Contd... 40 Publisher/Distributor : Head, Library and Information Division, Bhabha Atomic Research Centre, Bombay 42 Form of distribution : Hard Copy 50 Language of text : Engli sh 51 Language of summary : English 52 No. of references : 53 Gives data on : 60 Abstract : The progress report gives the brief descriptions of various activities of the Fuel Chemistry Division of Bhabha Atomic Research Centre, Bombay for the year 1988. The descriptions of activities arm arranged under the headings :Fuel Development Chemistry, Chemistry of Actinides, Quality Control of Fuel, and Studies related to Nuclear Material Accounting. At the end of report, a list of publications published in journals and papers presented at various conferences/symposia during 1988 is given. 73 Keywords/Descriptors : PROGRESS REPORT; RESEARCH PROGRAMS; MICRDSPHERES; THORIUM OXIDES; URANIUM OXIDES; PLUTONIUM; URANIUM; SOLVENT EXTRACTION; ION EXCHANGE MATERIALS; RESINS; QUANTITATIVE CHEMICAL ANALYSIS; NUCLEAR FUELS; ACCOUNTNG; QUALITY CONTROL; BARC; CALIBRATION STANDARDS 71 Class No. : INIS Subject Category : B16.10 99 Supplementary elements : The previous progress report was published as BARC-1516 covering the period 19B7 ( i ) PREFACE The activities of the Fuel Chemistry Division during 1988 are presented in this Annual Report in four sections. The first section, Fuel Development Chemistry, deals with sol-gel processes for the synthesis of microspheres such as DC, (U,Ce)C and ThO2 and non-nuclear ceramics. Also included in this section are the measurement of some of the thermodynamic parameters of materials like nickel telluride, sodium zirconate, caesium chromate, caesium molybdate etc. which are likely to be encountered in operating reactors. The second section, Chemistry of Actinides, deals with i,he solid state, sol'ition and process chemistry of actinides. Solid state reactions and phase studies are the areas of work covered in solid state chemistry. In solution chemistry, solvent- extraction studies with Di-2 ethylhexyl phosphoric acid (D2EHPA), long chain secondary amine Amtaerlite LA-1 and MOFPA and ion-exchange studies from mixed solvent media have been covered. Studies in Process Chemistry include evaluation of different anion exchange resins for plutonium processing, recovery of U-233 from phosphate containing aqueous waste, recovery and purification of plutonium from fuel fabrication scrap etc. i i ) The third section, Chemical Quality Control of Nuclear Fuels deals with analytical methods, primary ciemica) standards and analytical services. Analytical methods cover the developmental work on electrochemical, titrimetric and mass spectrometric methods for the chemical quality control of nuclear materials. Preliminary studies on the preparation and characterisation of primary chemical assay standards for uranium and plutonium are reported. In the fourth section, Nuclear Materials Account ing,an overview of NUMAC database maintained for the nuclear materials accounting ir. all the DAE Facilities has been given. The report also includes a list of papers published in Journals and presented at various Con f er ences -'Sy rr.pos i a.. The Editor expresses his sincere thanks to Dr. D.D.Sood, Head, Fuel Chemistry Division for his valuable suggestions in the preparation and organisation of this report. He also highly appreciates the cooperation and help rendered by Sri Satya Jyothi in compilation of this report. ( ii i CONTENTS PREFACE ( i ) 1. FUEL DEVELOPMENT CHEMISTRY 1 1.1 Soi-Gel Process Development. 1 1.2 High Temperature Thermodynamics. 20 2. CHEMISTRY OF ACT IN IDES 35 2.1 Solid State Chemistry. 35 2.2 Solution Chemistry. 47 2.3 Process Chemistry. 76 3. CHEMICAL QUALITY CONTROL OF FUELS 99 3.1 Analytical Methods. 99 3.2 Primary Chemical Assay Standards. 131 3.3 Analytical Services. 139 4. NUCLEAR MATERIALS ACCOUNTING 142 5. PUBLICATIONS 146 FUEL DEVELOPMENT CHEMISTRY 1. 1 SOL-GEL PROCESS DEVELOPMENT 1.1.1 SYNTHESIS OF URANIUM MONOCARBIDE MICROSPHERES BY SOL-GEL PROCESS. S.K. Mukerjee, J.V.Dehadraya, Y.R.Bamankar, V.N.Vaidya and D.D. Sood. Introduction The present trend in fast reactors towards higher operational temperatures, higher linear power rating, and higher breeding ratio points towards the possible use of carbide as nuclear fuel for future systems. Presently carbothermic reduction of UO2 + PuO2 powders followed by grinding and pe11etization of carbide powders is the established route for mixed carbide fuel fabrication. Sol-ge! method of preparation of carbides has several advantages over the conventional pellet route; as it uses a minimum number of steps with carbothermic reduction itself leading to dense particles and no grinding or milling of reactive and pyrophoric powders is required. Several reports^"4-' have been written concerning the carbo- thermic reduction of carbon containing oxide gel particles to carbide. Al1 these reports deal with the reduction carried out in two steps, first the reduction of UO3 to UO2 by hydrogen followed by carbotherinic reduction. This was necessary as the C/M ratio, approximately 3.5, required for direct carbothermic reduction of the particles could not be obtained because of difficulties in dispersion of large amounts of carbon. The hydrogen reduction of UO3 itself is complicated due to side reaction of hydrogen and moisture with carbon and thus fixing of the carbide s to i ch i oine try becomes difficult. The present work overcomes this problem as sufficient quantity of carbon can be dispersed in the gel particles using proper metai ion concentration by the use of gelation field diagram^'' and C/M ratio around 3.5 can be obtained easily in the gel par t ic1es. Ex per i menta1 Internal gelation process and the gelation assembly used for the preparation of gel particles have been described elsewhere^-', Maximum C/M ratio obtainable with standard feed composition is 3.3 as uranium molarity in such a solution is 1.2M and (HMTA, urea)/uranium mole ratio is 1.4. Thus in the present investigations the uranium molarity in the feed solutions was lowered. Uranyl nitrate solution (3M) was mixed with a scliitior, of HMTA and urea containing finely dispersed carbon powder (United KAF) in cold condition (0°C) to obtain feed solution 1.1M in uranium and having (HMTA, ur ea ) / u ran i. urn mole ratio of 1.5. The droplets of this solution were contacted with hot silicone oil to obtain UO3 gel partic! ','F with homogeneously dispersed carbon. Total 9 batches were prepared and C/M mole ratio in the feed was varied between 3.45 to 3.50. The gel particles were washed, dried ami heated in argon upto 300°C to remove moisture, ammonia and residua! gelation agents. Experiments on heat treatment were done on 50 g per bai-rh scale in a tantalum carbide crucible. Heating was carried out in high temperature high vacuum tungsten heater furnace. Initially UO3 was reduced to UUo by reaction at 700°C. Carbothermic reduction of UO2 was initiated around 1200° C and soak time at various temperatures varied from 2 to 10 hrs. Sintering was done for 2 hours at 1700°C in argon. The product was analysed chemically for oxygen uranium and carbon. X-ray diffraction analysis was done to identify the phases present. The density was determined by stereo pycnometer. Results and Discussion The results are summarised in Table 1. In all cases silvery shining and crack free microspheres were obtained. Reduction and sintering carried out at lower temperature resulted in high oxygen content. Thi? could have resulted because of slow rate of reaction at low temperatures as the particles sinter and close the pores and do not allow the reaction to go to completion. Similar results were obtained when the reaction was carried out at high temperature as the sintering became faster than the reduction. A compromise was made by increasing the temperature from 1300 to 1500°C at a very slow rate. This helped the reaction to proceed at a steady rate and the product contained around 1000 ppm of oxygen. Further work is in progress to lower oxygen content. This work has shown the possibility of using direct carbothermic reduction of gel particles to obtain UC microspheres using internal gelation process without resorting to a hydrogen reduction step. References 1. J.I. Federer and V.J. Tenner, ORNUTM-6089119781. 2. K. Bischoff, V. Scherer and H. Schumacher, Symposium on Sol-Gel Processes and Reactor Fuel Cycl e, CONF-700502, U5M 1 970 ) 3. K.Bischoff, M.H. Lloyd and H, Schumacher, So I-Gel Processes for Fuel Fabrication 1AEA-161, International Atomic Energy Agency, Vienna, 95 (1974i 4. A.Facchini and P. Gerontopulos, Sol-Gel Processes for Fuel Fabrication, IAEA-161, Internationa I Atomic Energy Agency,Vienna,95.11974 ). 5. V.N. Vaidya, S.K. MukQrjee, J.K. Joshi, R.V. Kamat and D. D. Sood, J. Nucl. Mat., 148, 324 ( 1987). 6. V.N. Vaidya, R.V. Kamat and D.D. Sood, Radiochemistry Division Annual Progress Report for 1978, BARC-1114 (1981). Table - 1 Preparation of UC raicrospheres from (UO3 + C) gel Particles by Carbo-thermic Reduction alone. Batch c/u Temp Reaction Product X ilo. in gel °C timelhrsI U On C Phase CB-2 3.56 1450 2.0 94.5 0. 18 3. 12 uc,uc2 CB-4 3.54 1450 2.0 94.3 0.21 5.01 l)C,VC2 CB-6 3.51 1450 94.8 0. 16 4.88 2.0 «uc,uc2 CB-6 3.51 1300 6.0 91.9 1.20 7. 04 UC,U02 CB-6 3.51 1300 10.0 92.9 0.68 6. 49 UC,UC2 CB-8 3.49 1300-1500 in 4 hrs 95.0 0. 11 4.83 UC CB-9 3.50 1300-1500 in 4 hrs 94.9 0. 10 4.81 UC * UC is the main phase. 1.1.2. SYNTHESIS OF (U,Ce)C MICROSPHERES CONTAINING 20% OF Ce BY SOL-GEL PROCESS. S. K. Mukerjee, J . V. Dehadraya, Y.R. Bamankar, V.N. Vaidys and D.D. Sood. This discussion presents the studies on the direct reduction of (U,Ce)02+x with carbon in vacuum for the preparation of (U,Ce) monocarbide microspheres. This work was carried out with a view to obtain experimental conditions suitable for the preparation of (U,Pu)C, which is a potential nuclear fuel for fast reactors. UO3+ CeO2 microsphores containing 5-20% of Ce and 3.35 to 3.55 moles of carbon per mole of metal prepared by internal gelation process. Cerium nitrate solutions having NO^/Ce = 4.0F> us? prepared by first precipitating Ce as Ce(OH)^ from solution of 1.1.3. KINETIC STUDY OF THE CARBOTHERMIC SYNTHESIS OF URANIUM MONOCARBIDE MICROSPKERES. • S.K.Mukerjee, J.V.Dehadraya, V.N.Va.dya and D.D. Sood. Isothermal kinetics of the carbothermic reduction of porous uranium oxide microspheres, having carbon black uniformly dispersed in them was studied under vacuum and flowing gas from 1250 to 1550°C. The quantity of carbon monoxide gas evolved with tin"- during reduction was used to determine the rate of formation of the carbide. The results of these studies were found to be useful in understanding the mechanism of the carbothermic reaction involved, and in defining the heat treatment scheme for preparation of good quality uranium monocsrbide microspheres. Ex per imenta1 The 'UO3 + C) gel particles were prepared by the process described earlier. These particles were heated at 700°C under vacuum to obtain (UO2 + C) particles. The k ' 1 etic studies were carried out in a high temperatura/high vacuum controlled atmosphere furnace. Experiments were done on 50g per batch scale in tantalum carbide crucible. Carbothermic reduction of UO2 takes place according to the following reaction. UO2 + 3C > UC + 2 CO - (1) The carbon monoxide measurements in the flowing gas conditions; was measured by its oxidation to CO2 and absorption in alkali. However under vacuum the rate of CO gas released from the sample, Q(t), was calculated from the pressure inside the reaction vessel at time 't', P(t) and the pumping speed 'S' of the system using the following relation: Q(t) = S . P(t) Reaction ratio a was calculated by integrating Q(t) with respect to 't'. The reaction ratio a is defined as, « = (W£ - W^)/(Wj - Wf) where wj , w^ and Wf are initial weight, weight «t timo 't', and final weight of the sample. The intermediate and the final products were analysed chemically for carbon, oxygen and nitrogen content. X-ray diffraction analysis was done to identify the phases present. Results Typical data for the reduction carried out under flowing purified argon and vacuum at 1450°C are given in Table-2. At lower temperatures, the time required for completion of more than 95 percent of the reaction was more than 20 hours. Two experiments were carried out at each temperature. The data could be fitted in the equation -ln(l-o) = kt (diffusion controlled). The energy of activation for tho reaction under flowing gas was found to be 330 kJ.mol" . For experiments carried out under vacuum,the best fit for the results was obtained for the rate equation: = kt (interface controlled). Temperature dependence of the rate constants gave an activation energy of 330 kJ . mo I ~ *•. Discuss i on The possible rate controlling mechanisms for the oarbothermic reduction of spherical UO2 particles to 1>C has been suggested by Lindemer et al t3^. Based on this model the following rate controlling steps can be considered operative for the conversion of UO2 to UC. 1) C > tC]yC at the surface. 2) IClyQ diffusion from surface to the UO2 - UC interface. 3) The reaction CC](jc + UO2 > UC + 2[0]UC at the interface. 4) tOD^c diffusion to the surface of the sphere. 5) tO^uc + c > co (8> at the surface of the sphere. 6) Diffusion of CO gas through porous UC layer. in which [X] denotes species X in solid solution. Holmes et al^**] studied the conversion of 1.5 to 2mm agglomerates (made from submicron carbon and UO2) to UC in fluidised bed. L i ndemer ^5 •* analysed his data and compared total reaction time with the value theoretically calculated on the basis of reaction governed by steps 2 and 4 or step 6. He reported that the reaction appeared to be controlled by solid state diffusion process (step 4) as the increase in CO pressure decreased the reaction rate . Ainsley et a 1 *• ° ^ who carried out reaction in vacuum also reported solid state diffusion as the rate controlling step. They are of the opinion that step 2 is rate controlling, since if step 4 were rate controlling this would result in formation of sesquicarbide or dicarbide under vacuum as step 2 would be relatively rapid in the opposite d i rect ion. The results of the present work indicate that the reaction rate is diffusion controlled under flowing gas conditions but it changes to interface controlled under vacuum . Under flowing gas conditions diffusion of CO gas through porous UC layer (step 6) appears to be rate controlling since for solid state diffusion controlled reaction the mechanism will not change with change of reaction condition from flowing gas to vacuum. References. 1. V.N. Vaidya et al; J. Nucl. Mater; 148 (19871 324. 2. T.B. Lindemer et al; J. Amer. Ceram. Soc; 52. ,233(1969). 3. J,T. Holmes et al; USAEC report ANL - 7482, Argonne National Laboratories, USA (1968). 4. T.B. Lindemer, Nucl. Appl. Technol; 9,711(1970). 5. R. Ainsley et al; New Nuclear Material including Nonmetallic fuels, Proc. IAEA, Prague,34911963). 6. B. Serin and R.T. Ellickson, J. Chem. Phys; 9,742(1941). 7. W.D. Spencer and B. Topley, J. Chem. Soc;2633 (1929). 8. D.P. Stinton et a! ; J. Amer. Ceram. Socj 6£ ,596(1979). Table - 2 Carbothermic reduction of UO2 particles undar flowing gas and vacuum at 1450°C. C/U = 3.5,(C = 12.48% and U = 70.22%). Pumping speed of the system = 1250 1it./min. Under flowing gas Under vacuum sample weight = 47.50g sample weight = 54.35g Time Amount of CO2 Time Pit) Q(t) Amount of CO2 trapped g a min. mbar moles/min. ovolved IX 102 103 moles on 0.75 0.06 - 11 0.65 0.36 50 2.23 0.18 - 6 10.0 5.58 0.023 0.08 100 4.10 0.33 0* 20.0 11. 16 0.037 0. 13 150 5.46 0.44 12 15.0 8.37 0.120 0.42 200 6.70 0.54 23 10.0 5.58 0.177 0.62 250 7.70 0.62 27 8.5 4.74 0.188 0.66 300 8.45 0.66 30 7.0 3.90 0.202 0.71 350 9.19 0.74 38 4.5 2.51 0.225 0.79 400 9.69 0.78 42 3.5 1.95 0.234 0..82 450 10.06 0.81 50 2.0 1. 11 0.251 0..88 500 10.56 0.85 63 1.0 0.55 0.271 0..95 550 10.93 0.88 80 0.6 0.36 0.276 0..97 600 11.18 0.90 650 11.30 0.91 700 11.55 0.93 750 11.80 0.95 800 11.92 0.96 Q(t) = (P(tl . 10'3 /1.03J (1250/22.4) tt From integration of Q(t), t curve. I Temperature H50»C. 10 1.1.4. PREPARATION OF ThO2 MICROSPHERES BY INTERNAL GELATION PROCESS. N.Kumar, V.R. Ganatra, S.K. Mukerjee, V.N. Vaidya and D.D. Sood. The work on development of internal gelation process for the preparation of ThO2 microspheres was continued. Som^ experiments were carried out to study the gelation behaviour of formaldehyde denitrated thorium nitrate solution. Feed solution (200 ml) was taken in a jacketted beaker and viscosity of the feed solution was monitored as a function of temperature. The change in appearance of the solution was also noted. The temperature at which gelation of the solution occured was termed as gelation temperature. For various feed compositions (CTh3 = 1.1 to 1.4 b. R = 1.2 to 1.5) the gelation temperature varied between 60 ± 10°C. Three types of gels were obtained, namely, soft opaque (some cases resulted into phase separation! hard opaque and trans 1uscent. A "gelation field diagram* has been constructed, which can be used in selecting feed compositions suitable for the preparatirn of good quality microspheres. Few batches were made with thorium nitrate solution preneutraIized with NH4 OH solution. Feed compositions were selected from opaque hard and transparent region of gelation field diagram. The results indicated that for the same thorium concentration the amount of gelation agents required was more compared to the corresponding composition from gelation field diagram. This is expected as ammonium ion being present in the preneutralized feed solution would retard the gelation reaction due to common ion effect. 11 1.1.5. NON-NUCLEAR CERAMICS BY SOL-GEL ROUTE. R.V. Kamat, K.T. Pillai, N. Raghu, V.N. Vaidya and D.D. Sood. A. Preparation of Metal Alkoxides Metal alkoxides readily hydrolise to form a sol and henoe serve as the starting materials for obtaining a wide variety of high technology ceramics such as PZT.PLZT, superconducting ceramics, silica and alumina based ceramics of superior quality in the form of bulk bodies as well as thin-films, coatings, f i bres etc. Commissioning of Karl Fischer Instrument Realising the adverse effect of moisture in alkoxide synthesis an old Karl Fischer instrument was recommissioned for the routine moisture analysis of alcohols. Using anhydrous K2CO3 & CaO as desiccants the water content of the alcohols could be kept below 3%. For complete drying of alcohols treatment with calcium hydride will ba done shortly. Preparation of Aluminium Iso-Propoxide 150 gms of aluminium iso-propoxide was obtained by refluxing aluminium metal turnings with propanol at 78° C for 6 hours in presence of HgCl2 catalyst. After distilling out the propanol at 80°C, the alkoxide was distilled out at 115~120°C at 50 mbar pressure. The yield was seen to improve from 45% to greater than 90% by using small size Aluminium turnings and mosture-free HgCl2 catalyst. 12 Preparation of Copper Iso-Propoxide About 50g of copper iso-propoxide was prepared by the reaction of Li iso-propoxide with anhydrous cupric chloride obtained by careful heating of hydrated CuCl2 at 200°C in an atmosphere of dry HC1 gas. By refluxing the 2:1 molar mixture in moisture-free isopropanol, bluish green precipitate of cupric iso-propoKide was obtained. This was freed from the soluble salt of LiCl by repeated washing with iso-propanol and then dried in vacuum (0.5 mm) at room temperature. Preparation of Iso-Propoxides of Ca. Ba & Sr With a view to prepare high Tc ceramic superconducting materials by sol-gel route, the alkoxides of Ca, Ba and Sr were made on a trial basis with a batch size of 15g. The metal rods were cleaned and cut into pieces inside argon box and allowed to react with moisture-free propanol by refluxing for 10 to 20 hours at suitable temperature till the reaction was complete. The alcohol was then boiled off to get the respective metal alkoxide in the form of powder with 70 percent yield. The quality of the powder was not very good as the refluxing etc. could not be done in argon box. Improved lots will be made. B. Prepration of Sols and Gels Preparation of Alumina and Cupric Oxide Sols These were made by the controlled hydrolysis of the respective metal iso-propoxides. Nearly one litre of 0.05 molar stable sol of cupric oxide was obtained which was then concentrated to 0.2M. Further concentration appeared to destabilise the sol. One litre of 1.3 molar alumina sol was made by the procedure described in the previous report. 13 Preparation of Thoria and Yttria Sols The metal hydroxides were precipitated by adding ammonia solution to the metal nitrate solution. The precipitate was then washed repeatedly with dilute ammonia solution to wash out the nitrate ions. The washing was then continued with distilled water till the precipitate was free from ammonium ions as tested by Nessler's reagent. The slurry was then heated with vigorous stirring after adding a calculated amount of nitric acid to cause peptisation. The dilute sols were ready after about 30 to 40 hours of peptisation. Three litres of yttria sol (0.1 M) and then ten litres of thoria sol (Q.1M) were initially made. After curing at 100°C for about 100 hours, these could be concentrated to 0.4M yttria sol and 2.5M transparent thoria sol. Burium Carbonate Sol Attempts were mads to peptise BaC03 by vigorous stirring at 90°C in presence of baryta solution (0.1 mole ratio) and 0.1% gelatine. There was no sol formation even after stirring for 24 hours. Gels from Alumina, Thoria and Yttria Sols Gel preparation studies described in previous report were continued for alumina sol and these were extended to thoria and yttria sols. Several gel monoliths in the form of discs cf various sizes (10 to 20 mm dia and 2 to 12 mm height) were obtained by chemical gelation of alumina sol in Tpflon molJs in ammonia atmosphere followed by ambient drying. Unlike alumina the thoria & thoria-yttria gels were found to crack extensively during drying. Glycerol addition (1%) to the sol was tried to achieve flexibility to the gels but without success. 14 The dried alumina gels were crack-free and were having density of around 3.5 gm/cc as determined by stereo pychnometer. Heat treatment of these gels is planned to obtain high grade ceramics. C.Preparat ion of YBa2 CU3 0y_x Ceramics With the excitement over the discovery of high temperature superconductivity, it was decided to prepare YBCO ceramics by sol-gel route. As this route involves considerable effort and time, work was initiated simultaneously by oxalate route to gain some experience in this new field. Preparation of YBCD by Qxalate Route The precipitation pH,- of the oxlates of Y, Ba and Cu were separately found to be 0.16, 0.53 and 0.31 respectively. The metal salts were mixed In the mole ratio of 1:2:3 and precipitated with oxalic acid. Th2 precip.tate was washed and calcined at 500°C for 12 hours to get the oxide. This powder was cold pressed into 9mm dia pellets and sintered at 950°C for 20 hours in flowing oxygen. After slow cooling, (rate :l°C/min) the pellets exhibited superconductivity at liquid nitrogen temperature as shown by the ESR technique. The 4-probe measurement revealed the sample to show zero resistance at 81°K (Tc>. The XRD pattern showed the sample to be multiphasic having several undesirable peaks which could not be identified. More accurate control of precipitating conditions seems to be necessary for obtaining a single phase compound with higher Tc. 15 Preparation of YBCO by Sol-gel Route The sols of yttria & hydrous cupric oxide were made as described earlier, whereas the attempt to peptise BaC03 did not succeed. Hence a slurry was prepared by mixing barium iso-propoxide with sols of copper and yttrium in calculated amounts needed for 1:2:3 composition. Part of the slurry was coated on different substrates (Cu.Ni, SS and alumina) to get thin films. Remaining slurry was dried at roar.' temperature to get a powder, which was pressed into 9mm dia pellets. The pellets and coated substrates were heat trf3?,tr>d upto 860°C for various durations. The coating peeled off during heating. The pellets showed high resistance (mega ohms). Another attempt was done by refluxing the propoxides of Cu i, Bs with yttria sol and drying the slurry. The pellets were fired at 950°C for 20 hours in oxygen but the X-ray diffraction pattern could not be matched with any of the 9 known phases of Y.Ba.Cu and 0. Low temperature firing of the fresh pellets seemed to be promising with the resistance level in the kilo ohm range. The firing temperature was systematically changed between 250 and 550°C. A minimum resistance of 10 kilo ohms (by 2-probe) was noticed for the pellet fired at 470°C. Improvements were not observed with prolonged heating in oxygen. ESR technique showed a weak signal at 77°K. Further work has been postponed awaiting the commissioning of the argon dry box facility. D. Preparation of PL2T Ceramics. Lead lanthanum zirconate titanate route, trial experiments are bBing done by the oxalate route in ethanol medium. Starting with the titanium tetra chloride, titanium nitrate solution was obtained by precipitating the hydrous oxide with ammonia, washing it free from chloride and then dissolving in nitric acid. The solution was assayed to be O.2Z6M and was not very stable. Calculated amounts cf the nitrate salts of Pb,La,2r and Ti were mixed in the aqueous form to have the ratio X:65:35 where X is La(8%) and 65:35 is the PbZrG"3 : PbTiO3 ratio. Ethanol ic solution of oxalic acid was added to the nitrate mixture under controlled conditions to get a co-precipitate of PL2T oxalate. This precipitate was repeatedly washed with ethanol, dried at 120°C and then calcined at 800°C for 2 hours to get the PLZT oxide, (batch size 25g). The surface area of the oxalate and oxide powders was found to be 11.7m^/g and 3.37m2/g respectively. The cold pressed 9 mm dia button, after sintering at 1100°C for 6 hours, was not transparent. The crystalline nature of the PLZT oxalate powder, as indicated by its XRD pattern, may be due to the higher water content of ethanol. Work has been planned to obtain amorphous oxalate which will be hot pressed between 800-1100"C to obtain the transparent ceramic. E. Preparation of Thallium based Superconducting Ceramic. Several superconducting phases have been reported in the Tl-Ca-Ba-Cu-0 system. These include the 2021, 2122 and 2223 phases where the numbers refer to Tl, Ca, Ba and Cu atoms respective1y,in the formula unit. The work on the preparation of TCBCO compounds by solid state reaction of CaC03, BaCC>3, CuO and TI7O3 was taken up in collaboration with Metallurgy Division. The difficulties arose because of the volatility and high toxicity of T1203 and the 17 high reactivity of the final product with the cruciblu (alumina, silica and even silver) thereby affecting the superconducting properties. After several trial runs, these problems have been largely overcome to get the superconducting material showing an onset at 120°K and zero resistance at 9B°K, as determined by 4-probe technique. Several nominal compositions were tried and the heat treatment scheme is being changed carefully in order to push up the T to 125K. Partial substitution of thallium with lead, has been planned and trial runs are in progress. 1.1.6. PROCESS DEVELOPMENT FOR TCE GELATION. S. Suryanarayana , N. Kumar, V.N.Vaidya and D.D.Sood With a view to minimize the process steps and to limit the amount of liquids to be handled in glove box, work on modifying the process parameters and equipment for internal gelation process using Trichloroethylene as gelation medium has been taken up. Introduction of Trichloroethylene as gelation medium in place of silicone oil is advantageous as it eliminates the carbon tetrach1 oride washing step and its recovery by distillation. Following are the details of work carried out pertaining to process development for TCE gelation. Several trial runs were carried out to arrive at optimum values for [II] in feed broth, gelation time, gelation temperature, gelation column flow parameters, curing time etc. for TCE gelation. The optimized parameters are as follows: 18 ( i ) C'J] in feed broth Droplets from broths containing [U] less than 1.2 tend to float on TCE and hence not practicable in TCE gelation. Gelation time required for higher [IN is slightly higher than that of 1ower C U ] . (i i) Gelation medium temperature A gelation medium temperature of 60°C is found to be safe and practical in TCE gelation for both avoiding boiling of TCE as well as for achieving gelation temperature for complete gelation within reasonable length of the column. TCE being a low viscosity liquid has comparatively better heat transfer characteriFtics at 60°C than silicone oil at same temperature and hence could help achieve gelation temperature within short residence time of * 12 seconds in the column. (i i i) Ge1 at i on t ime Smaller droplets resulting from 0.5 mm capillary completely gel within the length of the column but bigger droplets of 2 mm and above require a curing time of 20-30 minutes in hot TCE for complete gelation. Gelation time for droplets with higher LU1 is achieved in one meter column as compared to shorter column of 75 cm suitable for lower [U], ( i v) Gelation column flow parameters As TCE is a very low viscous liquid compared to silicons oil, the column flows required for smooth operation of column had to be readjusted by modifying the gelation column side limb. In the gelation column used for silicone o i11 the height difference between main limb overflow and side limb discharge point was mai tained as iiigh as 20cm, by keeping the side limb discharge at a lower height, to overcome the resistance of high viscosity silicone oil for proper flows in main limb and side limb. TCE, being a low viscous liquid discharges a* a 19 higher rate in such a set up. The diameter of the side limb was decreased from 2.5cm to 1.5cm and the height increased to 2.5cm below the main limb overflow. This modification ensured a proper flow in the main limb for maintaining temperature and smooth removal of gelled product from the side limb. Dried UO3 microspheres from TCE gelation are pale yellow in appearance and these microspheres upon reduction and sintering resulted in crack free dense UO2 microspheres. Further work on modification of 10 kg/day assembly to suit TCE gelation process is in progrers. 1.1.7. OPTIMISATION OF HEAT TREATMENT SCHEME FOR OBTAINING U02 PARTICLES SUITABLE FOR GEL PELLETISATION S. Suryanarayana , N. Kumar, V.N.Vaidya and D.D.Sood UO2 particles suitable for gel pe11etization should be soft for oasy crushing and also should retain their free flowing characteristics with particle integrity intact during transfers. Pel Its obtained from such UO2 microspheres are expected to be free from berry structure and should sinter to required T.D. Addition of carbon as pore former and its subsequent removal during heat treatment to create necessary porosity for easy crushing is one of the ways known hitherto to produce UO2 suitable for gel pe11etization. In the present heat treatment scheme, it is aimed to exploit the lattice expansion properties of uranium oxide in its phase transformation during heat treatment to produce soft UO2 microspheres without resorting to pore forming additives. 20 Reduc t i on of UO3 to UO2 fol lowed by oxidation and reduction. In these experiments. UO3 microspheres are first heated upto 500°C,in argon atmosphere,reduced for 1 hour between 500-600°C in A1VH2 , subjected to a soak in oxygen at 650-700°C for one hour and then again reduced in Ar/H2 between 700-500°C, followed by stabilization in C02 at 300°C. The UO2 product obtained had high 0/M, reoxidised and heated up upon exposure to atmosphere, gave a high tap density and was found to be hard. Green and sintered pellets made from this product had retained berry structure. In a later run of this redox treatment, the stabilization at the end was done at 600°C. This prevented reoxidation and heating up of product on exposure to atmosphere but the quality of UO2 microspheres obtained from this heat treatment was poor in that they are hard and pel lets made from these microspheres retained berry structure in green as well as sintered stages. It is planned to modify the heating scheme in future runs and pursue the efforts to obtain soft UO2 microspheres suitable for gel pe11etisation. 1.2 HIGH TEMPERATURE THERMODYNAMICS. 1.2.1. STANDARD MOLAR ENTHALPY OF FORMATION OF NICKEL TELLURIDE Te00.405» N.K Shukla, R. Agarwal, R. Prasad, K.N. Roy and D.D. Sood. In an operating fast reactor, tellurium is a major fission product which can combine with the SS-316 cladding to form compounds like FeTeQ.9. CrTej_i and Ni3Te2 and thus can cause te11uriurn-induced liquid embrittlement of the SS-316 21 c I add ingl-1» 2]. A recent review on tellurides of first row transition metals by Kleykamp and Chattopadhyayl-3^ reveals lack of information on standard molar enthalpy of formation of these tellurides. From the point of view of cladding corrosion, the thermodynamic parameters of the compounds in equilibrium with metal are of importance. Hence studies have been undertaken to measure thermodynamic properties of these metal-rich te1 1ur ides. To start w;th, nickel telluride (Ni3Te2> has been chosen. This compound is called 8-phase and co-exists with metal(Ni) and has a homogeneity range of 40.0 to 40.9 atom percent tellurium at 853K. At higher temperatures more nickel dissolves and the homogeneity range extends from 37.5 to 42 atom percent te 1 ur ium'-^ ^. Whatever cal orimetric data exists in literature on molar enthalpy of formation of NixTe^_x is limited to te 1 1 ur ium-r ich compounds'-'*'. Geiderikh et al.^', have estimated standard molar enthalpy of formation of NixTe^_x (x=0.333 to x=0.565J from EMF data. More recently Prasad et a I . *• ^ 3 and Vishwanathan et al.^'-* have estimated standard molar enthalpy of formation of metal rich compounds < 6 -phase) using vapour pressure data. In the present study, the standard molar enthalpy of formation has been determined by direct method using solution calorimetry. The alloy was prepared by mixing high purity tellurium and nickel. The mixture was sealed in an evacuated quartz ampoule and heated at 1300K for 4 hours and then annealed at 1100K for 200 hours. The X-ray pattern of the alIoy agreed with the reported pattern and the chemical analysis of the alloy indicated the composition Nio.595^e0.405 • 22 Enthalpy of solution of alloy ( Nig^gsTeg 405' ancl synthetic mixture (0.595Ni + 0.405Te) were measured in 7M HNO3 ^ 5% H2SG4 acid mixture using an isoperibol solution calorimeter^^ . The standard molar enthalpy of formation of the alloy can. be given by Te0.405 ' —/\S_ o 1,H m {(0.595 Ni + 0.405 Te ) , S, 298.15°K)} Te0.405 'S' The enthalpy of solution of synthetic mixture and that of alloy were found to he -(221.026 ± 0. 157)kj.mo 1" and - (194.982 + 0.335)kJ. mo 1 respectively. The standani molar enthalpy of formation /_\.fHm - • : PRASAD er al. - A : K. C. MILLS o - •: PRESENT WORK V £ 20 — O : GEIDERIKH et al. V X I - a • 25 - o A A - A 30 i . . 1 ... I , . 1 0.3 0.4 0.5 0.6 -1 Fig 1. VARIATION OF - /\Hf(NiyTe).„)/kJ.mol WITH XN. 23 References. 1. Adamson.M.G., Aitken,E.A., Vaidyanathan,S., Nature.295,49(1962). 2. Adamson, M.G., Aitken, E.A., J. Nucl. Mater., 130,375(1985). 3. Kleykamp, H., Chattopadhyay, G., PSB-Ber 1549 (KI-II), Institut fur Material und Festkorperforschung,KFK,GaibH, 1982. 4. Mills, K.C., Thermodynamic data for inorganic Sulphides, Selenides and Tellurides 1974. 5. Geiderikh,V.A., Shevelva.S.N., Kutsenok,I.B., Krivosheya,N.S., Zh. Fiz. Khim 54,1068(1980). 6. Prasad.R., Iyer,V.S., Venugopal, V., Sundaresh, V., Singh,2. Sood, D.D., J. Chem. Thermodynamics , 19.,891 (19871. 7. Vishwanathan.R.,Sai Baba, M., Darwin Albert Raj, D.Balsubramanian, Saha, B., Mathews, C.K., J.Nucl.Mater, 149, 302(1987). 8. Venugopal,V., Shukla.N.K., Sundaresh,V., Prasad.R., Roy.K.N., Sood.D.D., J. Chem. Thermodynamics, .18, 735(1986). 24 Table - 3. 0 Enthalpy of solution of alloy, /V, ,H (Ni..cnt ,Ten .nc(S, 29B. 15K), enthalpy —iJal in 0.595 0.405 of solution of synthetic mixture, A_ ,H (0.595Ni+0.405Te,S.298.15K) in bo 1 in 7M HNO^ + 5% H-SO^ and standard molar enthalpy of formation of alloy, 1 (Ni0.595Te0.405'S'298-15K1 SI. Mass of 0 Mass of syn. 0 No. al 1 oy mixture —Sol m —Sol m g J kJ.mo! g J kJ.mol 1. 0.04990 112.6 195.437 0.05086 129.9 221.209 2. 0.05050 113.5 194.659 0.04974 126.7 220.618 3 0.04961 111.4 194.484 0.05033 128.3 220.785 4. 0.05073 114.1 194.800 0.05118 130.5 220.941 5. 0.05044 113.8 195.406 0.50848 129.8 221.126 6. 0.05044 112.9 193.860 0.04984 127.4 221.392 7. 0.05000 112.7 195.220 0.05019 127.9 221.063 8. 0.05035 113.4 195.067 0.05056 128.9 220.808 9. 0.04980 112.4 195.482 0.05144 131.4 221.240 10. 0.04982 112.4 195.404 0.04981 127.2 221.177 C - = (194.982 ± 0.335' ) -1 kJ.Mol H (0.595 Ni +0.405 Te, S, 298. 15°K ) ID = (221 026 + 0. 157 ) kJ.Mol -1 - (26.044 ± 0.369 ) kJ.Mol 1 Molar mass of alloy was taken as 86.6104. In each run amount of solvent (7M HNO + 5% H SO I was kept 500 g.* ± uncertainty is twice the standard deviation of the mean. 25 Table - 4 °(Ni Te, ,S,298.15°K)values along with lNi,, .Te . , S, 298. 15°K) m x 1-x m 0.59 COI5 n0.40 nc 5 Reference Composition (Nio Ten ..,., S, 298.15°K) Nl0.476 Te0.524 Nl0.400 Te0.600 Ni0.333 Te0.667 13 ± Ni0.565 Te0.435 Ni0.612 Te0.388 Ni0.629 Te0.371 (19-22 * 2"021 Ni0.630 Te0.370 Present work NiQ 5gg TeQ 4Qg (26.044 ± 0.582) 1.2.2 STANDARD MOLAR GIBBS FREE ENERGY OF FORMATION OF Na2Zr03ls) V.S. Iyer, V.Venugopal, ZiIey Singh, Smruti Mohapatra, K.N.Roy,R.Prasad and D.D.Sood Thermochemica1 information on Na2Zr03 is of importance as it is formed by the reaction of fission-product zirconium with coolant sodium during the clad breach of an operating fast reactor fuel pin. There is a paucity of thermodynainic information on sodium zirconate. In the present study, the standard molar Gibbs free energy of formation of Na2ZrD3(s) has been obtained by measuring equilibrium pressure of CO2 for the react ion Ns2C03(s) + Zr02(s) = Na22r03(sl + C02(g) by a static manometric method in the temperature range 878 to 1107K. Analytical grade chemicals were used for the experiments. X-ray analyses were done on the mixture before and after experiment for confirming the absence of any new phase. The C02 pressures were least squares analysed and can be given by log (p/kPa) = 7.470 - 7385.2 Using the standard molar Gibbs free energies of formation for Na2C03(s), Zr02(s) and C02(g ) from JANAF thermochemica1 Tables^-' and C02 pressure from equation (1), the standard molar Gibbs free energy of formation of Na2Zr03(s) was calculated and can be given by l rL\_fGm(Na2Zr03. S, T) /kJ . mol " = - 1676.27 + 0.348 (T/K) ±1.1 27 Table - 5 gives a comparison of these values with those of Maier and WarhusC2^. ft is seen that the two sets of data are in good agreement. Table - 6 gives the standard molar enthalpy of formation of Na2ZrO3 values calculated using the second law and third law method and the available literature values. It shows reasonably good agreement. Table - 5. Dependence of Standard Gibbs free energy of formation of on temperature. T/K , S.Tf/W.mol"1 Maier and Uarhus Present study 900 - 1363.9 -1363 .9 1000 - 1360. 1 - 1329.2 Table - 6. Standard molar enthalpy of formation of Na2Zr03(s) at 298. 15K. Authors / \H°,,/kJ . mol "^ Present study (Second law) - 1662.9 (Third law > - 1654.9 Bayer et al - 1686.3 Maier t* Uarhus'^-' - 1667.8 I4] Kohli - 1700.0 Reference;. 1. JANAF Thermocheraical Table NBRDSS, NBS, U.S. Dept. of Commerce, Supplement in J. Phys. Chem. Ref. Data, 4 ,1(1975). 2. Maier, J., Uarhus, U., J. Ch . Thermodynamics ,16,309(1986 1. 3. Bayer, R. P. , Bennigton, K. 0. ,Br . n.P.R. , J. Chem. Thermodynamics, 17., 11 <1985). 4. Kohli, R. , Thermochemica Act » , 65., 28511983 I. 28 1.2.3. THERMOCHEMISTRY OF Cs2Cr207(s,1). V.Venugopa1, V.S. Iyer, K.N. Roy, Renu Agarwal, R.Frasad and D.D. Sood. Ciiesium is a high yield fission product which participates in a number of interactions with fuel and stainless steel components of the cladding tube in mixed oxide fuel pins of fast reactors. The compounds of interest in the reaction of caesium with S.S. cladding have been suggested to be CsxCr 04(x=2 to 5) and their formation depends upon the oxygen and caesium potentials . Although Cs2Cr207 may not be formed in a normally operating oxide fuel pin, a study of this oxygen rich chromate will be of help in understanding the relative stabilities of caesium- rich chromates. The standard molar Gibbs free energy of formation and heat capacity data of (^s2^'r2l-)7(s,l) are not available in the literature. Hence work was undertaken to determine the free energy of formation of Cs2^r2'-'7(s''' by e.m.f method and heat capacity by drop calorimetry using Calvet micro-calorimeter. •_\ f Gm ( Cs2Cr207 , I,T) has been determined in the temperature range 797 to 87aK in the phase field: ( Cs2 Cr2O7(l) + Cs2Cr04(s) *• Cr203ls) ) using the cell Cr0 (s) + Pt/"Jr-;2Cr207 ( 1 ) + Cs2 A C r203 ( s )/0. 852 rO2 +0.15CaO/air/Pt where p(0r.) in air is taken as 21.21 KPa. The detailed Rxporimenta 1 procedure is given elsewhere. The relation of r?.ir..f. w i t-h temperature could be represented by E/mV ± 0.5 = 160 - 0.08104(T/K) - (1) 29 Using equation (1) and the reported thermodynamic values for 0 Cs2CrO4 and 0^03, the /_\_f Gm (Cs2Cr2C>7. 1,T) was calculated in the temperature range 797 to 874K and can be given by: -1 /AfGm The enthalpy increment measurements on Cs2Cr207(s,1) were carried out using a high temperature Calvet microcaI orimeter in the temperature range 335 to 826K. The details of the [4 ] experimental assembly and procedure are given elsewhere . The of 99.95 mass percent purity was used for the experiments. A solid-solid transition has been observed at (620.5 ± 1.5)K and the melting temperature was found to be (657 ± 1.0)K. The corresponding enthalpy values are (15.6 ± O^lkJ.mol"1 and (17 ± 0. 22 ) kJ . mo I ~1 . The enthalpy increment data were fitted in the form of polynominal and can be given by: ;)J.mol = -6.410x10 +1.939x10 T +7.441x10 T 335 to 820.5K 2- HT"H298. 15ICs2Cr2O7.slJ.mor1 = 1.313K106-4. 137K103T +3.517T2 620.5 to 650K 1 5 2 3. HT-"298 i5(Cs2Cr207,slJ.raor - 1. 218x 10 *3. 890x 10 T 657 to 826K The molar heat capacity value obtained in the study by extrapolation to 298. 15K is 230.3 J.K'/mol"1. 30 Using the molar enthalpy increment of Cs2Cr207 Cs2Cr2O7(l) were calculated and are given in Table -7. Using the presently obtained Cp> m ( Cs2C ^07 , s ) , cp, m(Cs2Cr2°7' ') • /_\J98' 15Sm from literature , the /A^S^ and (free energy function) 0mIT,298.15K) were calculated for Cs2Cr207(l). The enthalpy of formation of Cs2Cr207(s) at 298.15K was obtained by summation of the equations: Cs(s) = Cs( 1 ) Cs2Cr2°7(i' = Cs2Cr2°7(s J 2Cs(1)+2Cr(s)+l/2 02(g' = Free energy functions 0°(T,29B.i5K) for Csls), r" 1 Cr(s) and C^tg) were taken from literature. Values of /_\_f Gm (Cs2Cr207, 1,T) were taken from the present study. Table - 7 gives the thermodynainic functions for CS2C r^O? ' s ' at 700,eOO and 900K.The enthalpy of formation values at 298. 15K are constant showing no systematic error in the present investigation. Table - 7. Thermodynainic tunct'.on of 032^207* I) /lfH°(S,298.15K) K J.K ^mol ' J.K'^mol"1 J.K '00 589.25 216.41 372.8/4 -2099 WOO 6'il.29 238.08 403.21 -2101 00 607.19 254.92 432.27 -2100 Re (ere/ices : 1. M.G. Adamson, E.A. Aitken, J. Nucl. Hater. 130.3751 1985). 2. V.Venugopa1,V.S.Iyer,Renu Agarwal, K.N. Roy, R. Prasarl and D.D. Sood, J, Nucl. Hater. 11,1047(1987 1. 3. JANAF Thermocheroical Tables, J. Phys. Chem. Ref. Data (1975). 4. R. Prasad, Renu Agarwal,K.N. Roy, V.S. Iyer,V.Venugopal 3nd D.D. Sood, Presented at the International Symposium on Thermodynamics of Nuclear Materials, Chicago, Sept. 1988. 31 1.2.4. OXIDATION BEHAVIOUR OF UN AND U2N3 Jayanthi Kulkarni, G.A.Rama Rao, V . Venugopal and D.D. Sood Oxidation of UN Oxidation kinetics of UNIwith and without U02) wpre studied at 2x10 at in. oxygen pressure as well as in moisture with varying pressures ranging from 0.03 to 0.12 stm. The end product in all the cases was U3O3. U02 and U2N3 were observed as intermediates during the reaction. Uhen U02 content in UN was 12 weight percent, large concentration of ^2.^3 nas been observed as an intermediate phase compared to UN containing 0.8 '•'eight percent U02. Conversion of UN to U 3 0 g was faster with sample containing 12 weight percent I'Op. Diffusion seems to be the mechanism of oxidation with an activation energy of ths order of 170 kJ.mol . Sample containing 0.8 weight percent UO2 showed larger concentration of UO9 at intermediate stages. Mechanism observed was nucleation growth and 'be activation energies were less than 100 kJ.mol Oxidation behaviour of U 2 N3 Oxidation studies on U2N3 were continued. Formation of UN as an intermediate product with samples containing U02 as separate phase has been observed on two mere samples. Nucleation growth m^ohanisi.i was cbs «= r v <:, j cl •: r i r. g the initial stages followed by diffusio;; during n or. -isothermal oxidation studies. Diffusion becomes the only mccha. 11 s ni when oxidation was carried out in isothermal healing inodo. 32 1.2.5. VAPOUR PRESSURE OF Pd MEASURED BY KNUDSEN EFFUSION CELL MASS SPECTROMETRY. S.G. Kulkarni, C.S. Subbanna, S. Venkiteswaran , V. Venugopal and D.D. Sood * Radiochemistry Division Vapour pressures of Pd(g) over Pd(s) have been measured in the temperature range 1237 to 1826K by Knudsen Effusion Cell Mass Spectrometry Pi = k. I where, Pj = vapour pressure of gaseous ions of atom i. 1^ = ion current to gaseous ions of atom i. o' i = ionisation cross section of gaseous atom i. Hj •- isotopic abundance of atom i. t[ - multiplier efficiency of ion i. T = absolute temperature in K. k = pressure calibration constant. Values of k were obtained by measuring nickel ion intensities over Nils) before and after each palladium run . lonisation cross section and isotopic abundances were taken from standard Tables. Tj is experimentally determined. 33 The detector consisted of a secondary electron multipljpr and a Faraday cup coupled to an electrometer /liirn is capal-it/ of measuring currents down to 10 A. Operating f.HM at 10 - 1 9 gain, reliable ion intensities down to 10 A c~.. a J • 1 •. "j: y fr; :: i i ;,• be measured. The average enthalpy of vaporisation of Ni(418. 12±1.57)kJ.mo 1 compares well with literature value (422.99 ± 2.23) kJ.mol . Vapour pressures of Pd(g) over Pd(s) can be represented by a least squares analysed equation : log(p/kPa) = (10.91 ± 0.01) - (19069 ± 159)/T Second and third law enthalpies of vaporisation are (374.28 ± 5.42) and(381.71 ± 1.53) kJ.mcl"1 respectively. In the present study vapour pressure of Pd has been measured over a large temperature range of 50OK znd the lowest temperature of measurement is 1237K. Such a low temperature measurement has been made for the first time. 1.2.6. THERMODYNAMICS OF VAPORISATION OF CA£ L" ; i.:M MOLYBTATi]. R.P. Tangri*, V. Venugopa1 * *, D.K. Posf, H. S-uu<: a i am» » « * Metallurgy Division, ** fuel Chemistry Division and *** Analytical Chemistry tivis.oi;. Vaporisation behaviour of Cs9iv|o04 ( ! .' ims he^n i ;\:< • .•-; i i •;. « t: e J in the temperature range 1230 tr llii'K' i;y -t i ,.\s r. i ;• n t i o:t technique using oxygen as carrier pas. CP--MO",. •?•-•<-po.-s ; •: <; congruently without any decomposition in thi.- jLn-.ve t empp i a t'.:re range. This has been confirmed by X-ray diffraction and chemical analysis of condensate sncj boat resi.iua after 34 carrying out transpiration experiment. Experimental procedure and data are given elsewhere . As there is no information available on the existence of polymeric species in the vapour phase, vapour phase was considered to be monomeric. Vapour pressures have been least squares analysed and are given by: log (p/kPa) = (6.37 ± 0.29) -(11452 ± 370>/T(K> The enthalpy and entropy of vaporisation at the mean temperature of the present study are (219.26 ± 7.08) kJ.mo I and (83.56 ± 5.5) J.K .mo I respectively. The second law and 0 third law enthalpies of vaporisation (/_SH (vap,298.15K) ai e (317.3 ± 13.8) and (335.8 ± 2.1)kJ.mo 1"1 . In view of the small temperature range of the present study, the second law value is less re 1iabIe. Reference 1. R.P.Tangri,V.Venugopal,D.K.Bose, and M.Sundaresan, a paper presented at the International Conference on the Thermodynamics of Nuclear materials held at Chicago,U.S.A, September,1988. Fig 2. TG & DTG CURVES OF Cs2Cr04 & Rb2Cr04 IN Ar-7% H2. A: TG & B: DTG OF Cs2Cr04 (54.9 mg) C: TG & D: DTG OF Rb2Cr04 (60.2 mg) 35 2. CHEMISTRY OF ACT IN IDES 2.1 SOLID STATE CHEMISTRY 2.1.1. SOLID STATE REACTIONS OF (U,Th)02 WITH CARBONATES OF SODIUM AND POTASSIUM K.L. Chawla, N.L. Misra and N.C. Jayadevan Several ternary compounds in the Na-U-0 and K-'J-O systems are ana< well known. These are a-Na2U04 , C-Na2U04 , Na2U2G7 , NafjU^O^ Na20.2.5U03 for sodium and K2UO4,K2U207.K2U4013 and K2U7022 for potassium, all compounds of U(VI). The preparation of and K2U/j0|2, uranates in which some of the uranium atoms are in a lower oxidation state indicated the possibility of replacing these atoms with thorium atoms. The results of the reactions of caesium carbonate and rubidium t 2 ] carbonate with (UQ gThQ i'O2 were reported earlier . As a continuation of this study Na2C03 and K2C03 were heated with equimolar quantities of (UQ_yThg.3)02 solid solutions. The products formed at 900°C were examined by means of their X~ray powder diffraction. The data are presented in Table - 8. Na?C03 gave a mixture of and the carbonates are heated in air, the products were the uranates and References. 1. K.L.Chawla, N.L.Misra and N.C.Jayadevan, J. Nucl.Mater., 154.181(1988). 2. K.L. Chawla, N.L.Misra and N.C.Jayadevan, Proc. Radiochera.and Radiation Chem. Symposium, Bombay(19881, PaperNo.CT-27. Table - 8 X-ray Diffraction Data * Na2C03 + M02** a-Na2U04 K2C03 900 °C 900 °C 900°C 900°C d(A> I"° d(A) i/io d(A) i/io d(A) i/io 4.92 80 4.95 100 6.56 15 4.16 10 3.230* 100 3.430 15 3.408 10 3.280 50 3.164* 100 2.848 100 2.857 83 3.230* 100 2.741 40 2.793K 40 2.839 10 2.637 10 2.466 10 2.471 16 2.801* 45 1.942* 40 2.210 10 2.218 12 2.180 10 1.653 35 2.017 40 2.019 27 2.013 25 1.581 15 1.970K 50 1.970* 10 1.683 50 1.926 10 1.647 40 1.649 21 1.714 15 1.688* 65 (U,Th)02 lines * a = 5.509 A 37 2.1.2. PHASE STUDIES IN THE Na-U-0 SYSTEM A.K. Chadha and N.C. Jayadevan Knowledge of the formation of different phases in the Na-U-0 system are important from the point of fuel coolant interaction in fast reactors. Following the observation that the addition of small amounts of alkaline earth ions into the K-U-0 system leads to the formation of new cublic solid phases, 2+ 2 + a study of the Na-U-0 system in which Sr ,Ba or Ca ions are added was taken up as a continuation. (a) Na-U-Ba-0 system. Mixtures of NaN03,U02 and Ba(N03)£ in different molar proportions were heated in the temperature range of 900-1050°C and the reaction products examined by means of their X-ray powder diffraction patterns. The compositions studied and the phases identified are shown in Table - 9. in all the experiments either the amount of Na(N03> or Ba(N03>2 was varied maintaining the concentration of UO2 between 1.0 and 1.1M. When the amount of NaN03 was varied from 1.80 to 3.50 keeping Da(N03>2 constant at 0.1M, the products gave X-ray lines of a new cubic phase (Phase 1) alongwith those of either Ua.2 Ba(N03)2 between 1.40M and 1.70M. Yet another phase designated as Phase 111 was obtained for Ba(N03>2 .4.0M or above. All the three phases (I, II & 111) isolated in this study were found to be new phases unknown in the Na-U-0 or Ba-U-0 systems. 38 The X-ray powder patterns of phase-1 and phase-I1 have been O 0 indexed on cubic systems with 'a' values of B.401A and 15.00 A respectively. The indexed data are given in Tables 10 and 11. (b) Na-U-Ca-0 System. The results of experiments similar to that of the Ma-U-Ba-Q system conducted by substituting CaCO.g in place of BalNO^)? are shown in Table -12 along with the solid phases identified as reaction products. For most of the compositions, the products were identified to be mixtures of Na2UC>4 and Table - 9. Phases identified in Na-Ba-U-0 System. Composition of Starting Mixture Heating Conditions Phases identified NaN0,3 UO2 Ba(N03)2 Temp Time from X-ray pattern (°CI (hrsl 1 80 1 0. 1 1050 3 B-Na2U04 * Phase 1 2 00 1 0 . 1 870 4 oc-NanU( + Phase 1 2 50 1 0. 1 880 8 + Phase I 3.00 1 0 1 930 3 Phase + Na?U?07 3.30 1 0 1 980 3 Phase + J3-Na2UOA 3.' J 1 0 1 950 3 Phase + Add i tona 1 1ines " 25 1. 10 0 25 950 3 Phase only i.52 t.10 0 575 950 3 Phase 1 + Add i tona 1 1ines 3.50 1. 10 1 10 950 3 Phase I 3.30 1.10 1 43 950 2 Phase 1 1 3.30 1.10 1.60 950 2 Phase 1 1 3.30 1. 10 1.69 950 2 Phase 1 1 3.30 1. 10 2.50 950 2 Phase 1 1 + Phase 1 1 I 3.30 1. 10 3.00 950 2 Phase 1 ! + Phase III 3.30 1.10 3.50 950 2 Phase 1 ! • Phase 111 3.30 1. 10 4.00 950 2 Phase 1 I 1 3.30 1.10 4.50 950 2 Phase i I I 39 Table - 10 Indexed Pattern of Phase I in the Na-Ba-U-0 System Corr.26 Observed Calculated H (hkl) s inJ0 sin5 9 22 15.01 .01706 01684 2 (110) 35 21.26 .03403 03368 4 (2001 22 26.01 .05064 05052 6 (211) 100 30. 14 .06760 06730 8 (220) 14 33.74 .08422 08420 10 (310) 11 37. 14 .10142 10104 12 (222) 16 40.24 .11833 11788 14 (321) 39 43. 10 .13495 13470 16 (400) 11 45.89 .15198 15156 18 (411!;((330) 33 48.54 .16895 16840 20 (420) 60 53.48 .20245 20208 24 (422) 12 55.83 .21917 21892 26 (510;(431) 8 60.38 .25288 25260 30 (521! 22 62.63 .27013 26944 32 (4401 5 64.63 .28577 28628 34 (530);(4331 15 66.83 .30327 30312 36 (600);(442) 5 68.88 .31984 31996 38 (6111 ; (5321 22 70.93 .33664 33680 4. (620) 8 75.03 .37084 37048 44 (622) 8 78.98 .4044 40416 48 (444) 5 BO. 93 .4218 .4210 50 (7101;(543) 0 Body Centred Cubicz :: a = 8.401 A 40 Table - 11 Indexed Pattern of Phase II in the Na-Ba-U-0 System Corr.29 Observed Calculated (hkl > sin19 sina0 17 17.82 0.02399 0.02114 (220) 22 26.70 0.05292 0.05284 (420) 100 29. 17 0.06341 0.063408 (422) 14 34.32 0.08705 0.08720 (5221;(441) 36 41.695 0.12665 0.12682 (444) 7 45.67 0.15061 0.15060 (722);(544) 46 51.67 0.18991 0.19022 (660);(822) 6 55. 12 0.21407 0.21400 { 900) ; (841) 17 60.47 0.25356 0.25363 (844) 6 63.57 0.27745 0.27741 (854) ; (10,'', 1) 17 68.57 0.31732 0.31704 (10,4,2) 6 76.22 0.38090 0.38050 (12,2,1) (982);(876) (10,7,0) Cubic a = 15.00 A 41 Table - 12 Phases ident i f ied in Na-Ca-U-0 System Compos ition of starting Heating Condit ions Phases mixture NaN03 U02 CaC03 Temp Time CO (hrs) 1.46 1 0., 11 930 2.5 Na2U20? + fi-Na2U0, n 1.59 1 0..1 930 2 Na2U2 7 + new phase 1..694 1 0.7 930 2 Na2U207 • 8-N32U0/, + new phase ;_ 1.,75 0.11 800 34I a-Na2U0/( • new phase + 1..765 1 0. 19 930 2 Na2U207 4 B-NanUO/j new phase 1.77 0.80 0.3 920 3 new phase n 2 1 0. 11 930 2.5 Na2U2 7 + 6-Na2U04 + new phase 1 0. 11 fl-Na U0 2.7 930 3 Na2U2°7 + 2 4 3.2 1 0. 1 950 3 Na2U207 •• 6-Na2U04 3.5 1 0. 1 900 3 8-Na2L/0,(, + new phc.se However some additional lines were observed for 1. 50M Table - 13 X-ray data for Na-U-Ca-0 System NaN03: U02 : CaC03 1.77 : 0.80 : 0.314 2 I/I0 26 sin 9 sin'G (hkl) obseived calculated 24 15.,09 0.01726 0.01700 (100) 94 21.,54 0.03492 0.03460 (002) 0.03480 (110) 11 26.29 0.05172 0.05164 (012) 13 26.69 0.5328 0.05254 (102) 29 30.24 0.06804 0.0680 (020) 100 30.40 0.06936 0.06954 (112) 26 31.09 0.07182 0.0716 (200) 8 32.44 0.07802 0,.7794 (0031 13 34.34 0.8715 0.,0858 (120) 19 37.34 0.10247 0.. 1029 (113) 24 36.04 0.10621 0.,10624 (202) 11 *.}.54 0.12002 0.,1202 (212) 48 43.84 0.13936 0.1396 (220) 10 46.56 0.15608 24 48.84 0.17092 0.1709 (130) 48 49.34 0.17422 0. 17424 (222) 15 50.04 0.17887 0. 17956 (131) 37 53.89 0.20533 0.20554 (132) 35 55.04 0.21350 0.21446 (1241 11 56.84 0,22651 0.22716 (214) 24 63.65 0.2741 0.2720 (040) rthorhombic 5.762 A , b 5.913 A and 8.284 A 43 c) Na-U-Sr-0 System. For the studies in this system UD2iNaN03 and Sr(NC3>2 were heated together. By taking 1. Oh U02 and 0.2-0.3'M Sr(N03)2 and increasing the amounts of NaNOg from 1 to 4.0M, a new phase was noticed at a concentration of 3.30M of NaNCl3. The X-ray lines of this new phase could he i n"K?v. ed on a cubic cell of a = 8.327 A*. The indexed pattern given iji Table -14 that this phase is present along with some other pha S3 The value of 8.327 A for the new phase i •; in ag r rjcmuii t W i th the smal ler ionic radius of Ca^+ compared to that of Ba2 + i n which case the cubic cell has a dimension of 5. 'iOi A. Addition of more amounts of Sr(N03)7 to the above • gave Sr3U0g phase. Table - 14 X-ray data for Na-U-Sr-0 System NaN03 :U02 :Sr{N03!? 3.20 :1.10 : 0.3 1/ 26 Corr.19 H Sin- 3 Observed Calculi i •-<} 14 35 14,,90 15 .00 0.0W037 2 0 .0171305 30 75 26..20 26 .30 0.051757 5 0 .05139 40 100 30.,25 30 .35 O.O6:<522 8 0 .068522 5 13 33. 85 33 .95 0.85237 10 0 .0865 10 25 37.55 37 .65 0. 1C4J2 4 11 38. 80 36 .90 0.11086 5 14 40. 525 40 .535 0.119967 14 0 .1199 5 13 42. 50 42,.60 0. 13H/5 19 48 43. 30 43,.40 0.136" 1 16 0.. 1370 7 10 45. 90 •'(6.,00 0. 15;:67 16 0.. 15417 5 13 47. 55 4 7. 0 1631 ? 19 0. 16274 18 45 48. 70 48. 80 0.J70655 20 0. 171305 17 43 63. 60 53. 70 C.2O2P3 15 38 54. 20 54. 30 0.2082? 12 30 55. 10 55. 20 0.21A64 0. 2K13 6 56. r\ 2 15 56. 25 0.222:-i5 ro \J - 22263 8 20 62. 90 63. 00 0.273G0 32 0.2740 6 15 67. 05 6 7. 15 0.30564 7 18 71. 45 71. 55 0.3117P 40 0. 3426 2 6 75. 20 75. 30 0.373121 Cubic: a = 8.327 K 44 2.1.3.THERMOCHEMICAL STUDIES ON Cs-Cr-0 AND Rb-Cr-0 SYSTEMS. S.Sampath, S.K. Sali, N.K. Kulkarni and N.C. Jayadevan. For the mixed uraniurn-p1utoniurn oxide fuels, chemical interaction between fission products and stainless steol cladding is recognized to be the result of the oxidation of chromium. Caesium is a reactive fission product which attacks the cladding. Thermochemica1 studies on the Cs-Cr-0 and the Rb-Cr-0 systems were taken up to understand the basic chemistry of these compounds under different conditions. The compounds known in these systems are Cs^r^l • Cs2CrO4, Cs3CrO4 , Cs4CrO4 and CSC^OQ for caesium and Rb2Cr20y ,Rb2CrO4 and RbCr3Q3 for rubid ium' * •*. The techniques used in this study are thermogravimetry(TG) and X-ray diffraction. The monochromates, M2CrO4 and the dichromates, (M = Cs or Rb) were heated in a thermoana1yzer in a flowing stream of argon-7% hydrogen gas upto 1200°C. The thermogravimetric (TG) and the differential thermogravimetric (DTG) curves recorded are shown in Fig.2. The weight losses recorded are found to be 4% upto 65O°C, 67.5% upto 1000°C and 80% upto 1200°C in the case of Cs2Cr04 . The final product obtained was identified to be only Cr203 . The weight loss also corresponds to the complete loss of caesium. The first intermediate reaction product formed at 650°C has a composition Cs2CrO3 on the basis of the 4% weight loss. This product did not give any X-ray diffraction pattern. However, this hygroscopic product was converted to Cs2CrO/4 on storing. The other intermediate product formed at 1000°C with a weight loss of 67.5% gave the XRD pattern shown in Table - 15. These lines do not correspond to that of Cr203 or any other known compounds in the Cs-Cr-0 system. Oxidation of this product by heating in air showed an exothermic reaction between 45 400 and 500°C with a weight gain of 3% resulting in . i.,i.:i\>ie of Cs2CrO7 and Cr203, thereby indicating, Uiat i .i the; intermediate only part of tho caesium is lost. The results from the TG analysis oi' RboCrO/t i •• nuite a i ;JI i . " • that of Cs2CrO^ described earlier. Th; 1. r. I i..: r corresponds to the formation of Rb2CrO,3 , -.he in':i.-iiY: ., i;»• i; formed at 1150°C is a compound of unknown com] jii iu;i a;i-i '..li' final product above 1200°C is identified as C.'/H-j •,/i * N ; ;:;:• loss of all the rubidium (weight loss 74%) as in .'..;.e c^.;.' of The intermediate gained 3.5% weight on oxidation in .? i i •:i\;, the formation of a mixture of Rb2Cr207 and C^U^. Tt• r? .-., y?cn content was analyzed to give values of 22.36 arid ;-.:.. GC: respectively for the caesium and rubidium intermediates. Th;-se results point to a composition fijjCrO^ for the ;r-i L •; -i-'.st. i h•' ••;; with x * 0.4. The X-ray powder pattern of the intermediates ar^ given in Table - 15. In the K-Cr-0 sy .= t%;nr t .JO non- s to ich iometr ic oxides KxCr02 with x values co-. r-Ti-.poi-i'iin;; >* <_• 0.7 < x < 0.77 and 0.5 < x < 0.6 have been reported^' Icrdin;; support to the above composition. The results of thn rfiriij-rlic1 and oxidation studies are summarised in Table? - i ^. Table - 15 X-ray Powder Diffraction Data on CsxCr0"2 3tv' ^bxC;O, CsxCr02 Rb.-C-.O,, Intensity d Intensity c m 3.479 m 3. ,93 m 2.549 m J.S'^.1 m 2.270 v 2.392 in 2.29 w 2.14) m 2.085 m 2r.:bi w 1.563 46 Table - 16 Thermal Behaviour of Cs2Cr04 and Rb2Cr04 Starting Atra. Temp. Overal1 Wt loss/ Ut loss/ Material °C composi tion gain gain of sol id observed calculated products (%l Cs2Cr04 "2 650 CS2Cr03 - 4 - 4.2 Cs2Cr04 H2 800 - 67.5 - 66.7 (x "0 0.4) CS2Cr04 H2 1200 Cr203 - 80 - 80 Air 500 Cs2Cr207 + 3 + 3.2 (x * 0.4) + Cr203 Rb2Cr04 H2 675 Rb2Cr03 - 6 - 5.6 Rb2Cr04 "2 900 * 60 * 61 (x % 0.4) 1200 - 74 Rb2Cr04 H2 Cr203 - 73.5 Air 500 + 3.5 + 3.5 IK** 0.4) + Cr203 TG of Cs2Cr207 in Ar-736 H2 showed that weight loss of around was observed above 500°C followed by rapid volatilisation. 207 behaved similarly. XRD of reduction products showed presence of Cs2Cr04 and 0^03 and Rb2CrO4 and Cr2C>3 • No new phases could be identified. Heating at higher temperatures gave only References 1. T.B.Lindemer, T.M.Bumann and C.E.Johnson,J.Nucl.Mater. 100,178(1981). 2. C. Delmas, M. Dovalette, C. Fouassier and P. HagenmulIer, Mater.Res. Bulletin 10,393(1975). 47 2.2 SOLUTION CHEMISTRY 2.2.1. STUDIES ON SOLVENT EXTRACTION OF Pu(IV) BY D2EHPA D.G. Phal, S. Kannan, V.V. Ramakrishna and S.K. Patil A. The nature of Pu(IV) complex extracted by D2EHPA from sulphuric-nitric acid solutions. Previous studies have shown^•^'that the composition of the Pu(IV) species extracted by D2EHPA in toluene from sulphuric and nitric acids was Pu'r^Yg and Pu(N03)2H2Y^ respectively. In the presence of controlled amounts of nitrate ion the species Pu(N03)H2Y5 was identified in the extraction of Pu(IV) from sulphuric acid , using different diluents viz. toluene, dodecane and chloroform. With a view to find out the maximum number of Y groupings in PUH2Y5 that are replaceable by nitrate or TTA anions, experiments on the extraction of Pu(lV) from sulphuric acid were carried out (i) in presence of higher concentration of nitrate ions (ii) by the addition of thenoyItrif1uoro acetone (HTTA) to the organic phase, and (iii) by completely replacing sulphuric acid with nitric acid. The D values of Put IV) obtained using 1M sulphuric acid and higher nitrate concentrations are given in Table -17. It is seen that the values of (D.F/DO - 1)/[NO§] (where F is the aqueous nitrate complex ing factor of Put IV)) remained almost constant with [NO3] with dodecane as the diluent and increase linearly using toluene and chloroform as the diluents. This suggests that predominantly, only mononitrate species is extracted in dodecane whereas the dinitrate species are also 48 involved in the other two diluents. The data were also obtained on the extraction of Put IV) as a function of the D2EHPA from nitric acid concentrations of 1,3 and 6M. They suggested the dominance of the mononitrate species at 1 M HNO3 and that of the dinitrate species at high HNO3 concentration when dodecane was the diluent. In case of toluene and chloroform diluents dominant presence of the dinitrato species was observed under all the conditions studied. It is significant to note that in no case, the extraction of Put IV) species with more than two nitrate ions is indicated. B. Synsrgism in the extraction of Put IV) by mixtures of D2EHPA and HTTA. It was felt that if the Y groupings in PuY2(HY2'2 are replaceable by nitrate ions, such a substitution may also be possible with anion such as TTA~. The extraction of Put IV) from aqueous sulphuric acid by mixtures of D2EHPA and HTTA was studied. The data obtained on the variation of sulphuric acid concentration with a constant concentration of D2EPHA, HTTA and a mixture of them using the diluents dodecane, toluene and chloroform revealed that in each case, appreciable synergism was observed in the Put IV) extraction with the mixture. If the rynergism is due to the replacement of the Y by the TTA groupings, according to + HTTA > PuH2Y5(TTA) + 1/2 H2Y2 - (1) and PuH2Y6 + 2HTTA — > PuH2Y4(TTA)2 + H2Y2 - (2) it can be shown that the plot of (D/DQ - 1)/[HTTA] Vs. CHTTA] where D and DQ are the distribution ratios of Put IV) obtained with a fixed concentration of H2Y2 and aqueous medium in the presence and absence of varying concentration of HTTA, 49 respectively, would be a straight line with positive values for both the intercept and the slope . Data obtained with dodecane and toluene as diluents are given in Table - 18. The plots of (D/Do - 1)/[HTTA3 Vs CHTTA3) for all the three diluents studied were found to be straight lines. Equilibrium constants for equations (1) and (2) were calculated and the values obtained are summarised in Table - 19. The fact that only two Y groupings in Pu^Yg are replaceable by either is N0§ or TTA" anions suggests that the formula PuY2 C. Antagonistic extraction of Put IV) by mixtures of D2EHPA and TQPO from sulphuric acid. It was shown earlier^-' that Pu(IV) species extracted from H2SO4 medium by D2EHPA taken in three diluents dodecane, toluene and chloroform was PuH2Yg . Probable structural formulae for this composition are: (i) PuY4.2HY or (ii) PuY2(HY2)2 Synergism in the extraction of Put IV) by D2EHPA in the presence of TOPO is expected due to the formation of species like (i) PuY4.HY.T0P0 or tii) PuY2 References. 1. D.G.Phal, S.Kannan and V.V.Ramakrishna, DAE Radiochem. Radiation Chem. Symposium,Bombay!1988), paper CT- 18. 2. Ibid. Paper CT- 20 3. Ibid Paper CT- 19 Table -17 Extraction of Pu(IV) by D2EHPA(HY) from 1M sulphuric acid and nitrate ion concentration. CNO3] D of Pu(IV) with (D.F/DO-1)+ [NO3] with M Dodecane Toluene Chloroform [HY] CHY] CHY] F Dodecane Toluene Chloroform = 0.4 F = 0.2 F = 0.2 F xlO xlO xlO 0 0.0720 0.0121 0.00417 1.00 0.1 15. 1 2.55 0.0879 1.33 2.74 2.79 2.75 0.2 26.2 5.00 0.205 1.56 2.80 3.22 3.85 0.3 34.4 7.14 0.340 1.77 2.78 3.47 4.86 0.4 43. 1 9.43 0.438 1.87 2.76 3.63 4.98 0.5 52.5 11.5 0.543 1.95 2.81 3.70 5.15 51 Table -18 Variation of the distribution ratio D of Pu(IV) with concentration of HTTA Aqueous medium - 0.224(1 sulphuric acid. Organic medium - D2EHPA + HTTA in various diluents. CHY)=O.005F in dodecane [HY]=0.04F in toluene [HY]=0.05F in chloroform D (D/Do-1) D (D/Do-1) D (D/Do-1) [HTTA] IHTTA] [HTTA] -4 CHTTA1 M X xlO xlO xlO 0 0.0723 - 0.0355 - 0.0116 - 2 0.276 1.41 - 0.265 1.09 4 0.746 2.33 0.,157 0.856 0.902 1.92 6 1.39 3.05 - 1.95 2.79 8 2.07 3.45 0.486 1.59 3.45 3.70 10 3.17 4.28 - 5.40 4.64 12 0.869 1.96 16 1.53 2.63 20 2.37 3.29 52 Table -19 Equilibrium constant data determined for different equiibria encountered in the Pu(IV) extraction by D2EHPA+HTTA. Aqueous medium - Sulphuric acid. log equilibrium constant Equi1ibrium dode- tolu- chloro- cane ene fora Pu, , • ""z-z, > —>PuY2(HY2)2l .+ 4 7.99 5.18 4.29 (aq) z Mol c ^ Mo) PuY2(HY2))2(o)+ "TTA(o) —>PuY(TTA)(HY2)2 + 0.5H2Y2(o) 2.57 2.55 2.20 PuY2(HY2)2(oJ+ 2HTTA(Q) ->Pu(HY2)2(TTA)2{•, H2Y2(o) 4.95 5.48 5.06 + 4 Pu, ,+2.5H Y ,+ HTTA . —>PuY(TTA) 10.56 7.73 6.49 (aq) z 2 ^(o2 )1 I(o) '(aq) +4 + Pu (aq) 2H2Y2(o)+ 2HTTA(o)->Pu(TTA.2(HY2.2 12.94 10.66 9.35 53 Table -20 Variation of the distribution ratio D of Put IV! Aqueous phase : 0.224 to 1.02 M Sulphuric acid Organic phase : D2EHPA + TOPO in various diluents. D(A+B> D CH2S04] DA (DA+DB) (A+B) c - Di iuent - Dodecane, A - ID2EHPAJ = 0.02F; B - 1[TOPO] = 0.02M 0.224 2.31 0.0710 2.38 0.616 0.26 0.423 0.242 0.0579 0.300 0.145 0.48 0.622 0.0598 0.0572 0.117 0.0815 0.70 0.821 0.0211 0.0706 0.0917 0.0566 0.62 1.02 0.00931 0.0597 0.0690 0.0507 0.73 Diluent - Toluene, A - [D2EHPA] = 0.2F; B - [TOPO] = 0.1M 0.224 2.93 0.208 3.14 0.996 0.32 0.423 0.294 0.163 0.457 0.278 0.61 0.622 0.0661 0.145 0.211 0.150 0.71 0.821 0.0223 0.153 0.175 0.115 0.66 1.020 0.00998 0.160 0.170 0.0999 0.59 Oi1uent - Chloroform, A - CD2EHPAJ = 0.4F; B-IT0P01 = 0. 1M 0.224 4.05 0.101 4. 15 3.66 0.88 0.423 0.411 0.0869 0.498 0.493 0.99 0.622 0.0911 0.0689 0.160 0. 179 1.12 0.821 0.0300 0.0580 0.0880 0.0957 1.09 1 J2 0.0145 0.0563 0.0708 0.0648 0.92 2.2.2. SOME STUDIES ON THE EXTRACTION BEHAVIOUR OF ACTINIDES BY LONG-CHAIN SECONDARY AMINES. A.G. Godbole, Rajendra Swarup and S.K. Patil. Extraction of actinides by long chain secondary amines is considerably weaker as compared to the tertiary and quarternary amines ^1,2]. However, it has two major advantages: (i) a high decontamination factor for the more common fission products and (ii) an easy back-extraction. A programme on the extraction of actinides by Jong-chain secondary amines has been initiated with a view to explore the possibility of using it for the recovery of uranium and plutonium from the scrap obtained during fuel fabrication, their separation from each other and from other impurities if any. In the present work, experiments were carried out on the extraction of Pu(IV) and U(V1) by Amberlite LA- 1 ( N -Dodecany 1 trialkyl- rnethy 1 ami ne ) in benzene from aqueous nitric acid and the results obtained are reported. Put IV) was prepared by extracting Pu into TTA from iM HNO3 and back extracting into 8M HNO3 . Sodium nitrite was used as the holding oxidant. Amine was preequiIibra ted with corresponding nitric acid concentration. An equilibration time of 30 minutes was chosen. 233|j was used as tracer for the experiments with U ( V [ ) . Preliminary experiments showed that the distribution ratios (D> obtained by taking the metal ion initially in either phase were same showing thereby the reversibility of the extraction system. Distribution ratio data of Pu(lV) and U Table -21. It is seen from the table that D values for Put IV) increase initially with a maximum around *7M HNO3 . In the case of U(V1) the D values obtained were almost constant with the increase in HNO3 concentration. Distribution data for Pu(IV) and U(V I ) were also obtained at higher temperatures which indicated decrease in D of PuilV) with increase in temperature. However, there was no sifnificant change in D of U(V I ) at higher temperatures. Attempts were made to back-extract Pu from the amine phase by using very dilute nitric acid. It was observed that the stripping was not complete. However, quantitative stripping was obtained when ammonium oxalate was used as the stripping agent. From these results, it is indicated that separation betwen Pu and U is more favourable around 7M ;iitric acid. However, more work is required to substantiate this. References. 1. P.R. Danesi, F. Orlandini, G. Scibona, RT/CH I 65. 21(1965). 2. Boyd Weaver and D.E. Horner, J. Chem. Eng. Data 5_ ,266(1960) 56 Table - 21 Variation of Distribution Ratio of Pu(lV) and U(VI! with nitric acid concentration. Organic phase LA-IX in benzene HN03 Distribution Ratio !P) (Ml Pu( I V) I" (VI - 5.5 - 8.9 - 14.2 - 14.7 0.61 11.0* o.r,3» 9. 4«« 0.52** 6 0.7V. 7 16.75 0. 74 8 - 0.67 9 9.2 0.70 10 6.4 0.64 * At 40 °C ** At 50°C 2.2.3. POLVFN1' F'.'. TRACT I ON STUDltS Oi- '("l /-NO F'l. ("1) BY Di-2 ETHYLHEXYL PHOSPHORIC ACID iO r,. V . Chet ty , P . M. Mapara, R.Swarup, V . V . Ramakr i shna aiui S.K.Pati1 1}_1 ion by D2EHPA The solvent extraction studies of hexavalent plutonium and uranium ions were carried out rising the extractant D2EHPA(H2Y2) from 0.1M and 0.2M aqueous sulphuric acid. The distribution ratio (D) data obtained as a function of [H2Y2;) and CH+] were plotted on a log-log plot which gave slopes of + 2 and - 2, respectiveIy and the slope values were independent of diluents used viz. dodecane, toluene and chloroform. From 57 the data (Table 22) it may be inferred that the extraction mechanism appears to be similar to that reported earlier for HCIO4 medium. Extraction by mixture of H2Y2 and TOPU The extraction studies of both Pu(Vl) and U ( V I ) were carried out from 0.2M H2SO4 using mixtures of H2Y2 and TOPO at constant TOPO. The D values were obtained as a function of [H2SO4] by keeping the concentrations of [H2Y2] and [TOPO] constant (Tables 23 and 24). The log-log plots of /\_D Vs. [H + ] gave straight lines with slopes close to -2. The log-log plots ] and Vs of /_\D Vs. CH2V2 £AP - CTOPO] (Tabl es 25 and 26) gave straight lines with slopes of 1.8 and 0.6 respectively. The interaction between H2Y2 anc' TOPO is probably responsible for the lower slopes than the expected values of 2 and 1 respectively. From the data it may be concluded that the species responsible for synergism probably is ] M02[H Studies on absorption spectra The absorption spectral work was carried out for U(V1) with a view to infer the nature of species extracted into the organic ph.de. The absorption spectra of U(V1) extracted into H2V2 in different diluents from different acids were recorded in the range of 350-500 nm. The extraction was carried out from two different concentrations of HCIO4 , HNO3 or HC1. In case of toluene as the di1uent.sharp peaks were observed for the U(VI) species extracted from higher acidities than those obtained from lower acidities suggesting the species extracted in these two cases are probably different. However, the spectra of U(VI) solutions using dodecane as diluent extracted from HC1 medium and chloroform as diluent extracted from HNO3 were independent of acid concentrations. 58 Table -22 Variation of distribution ratios of Pu(VI) and U (V1 ) with Cl^^ in different diluents. Aq. phase = 0.1M H2SO4 . (D2EHPA) D Values C _ Pu(VI) U(V1) 1. Dodecane 0.001 0.0178 0.0463 0.002 0.0705 0.201 0.003 0. 157 0.463 0.004 0.287 0.850 0.005 0.459 1.46 II. Toluene 0.004 0.0422 0.008 0. 144 0. 171 0.010 0.212 0.012 0.271 0.371 0.016 0.553 0.683 0.020 0.905 1.01 I I!.Chioroform 0.010 0.0678 0. 141 0.020 0.264 0.563 0.030 0.588 1.25 0.040 1.01 2.20 0.050 1.66 3.39 59 Table - 23 Variation of D of PulVI> with [H+] Organic Phase 0.020 F H2Y2 [H2SO4] DAB /\D 0.198 0.940 0.299 0.00585 0.635 0.394 0.245 0.0736 - 0. 171 0.590 0.110 0.0319 0.00613 0.0720 0.786 0.0651 0.0204 0.00492 0.0398 0.982 0.0406 0.0108 0.00329 0.0265 Table - 24 Variation of D of U(VI) with CH+I Organic Phase : 0.020 F H2Y2 [H2SO4) /\D DAB DA ( M 0. 198 18.0 0.737 0.0235 17.2 0.394 5.70 0. 180 0.0154 5.50 0.590 2.90 0.0790 0.0120 2.81 0.786 1.71 0.0440 0.00971 1.66 0.982 1.08 0.0263 0.00926 1.04 60 Table -25 Variation in D of Pu(VI) and U(VI > with CH2Y2] in presence of TOPO Aq.Phase: 0.2M H2SO4 Diluent: Toluene (TOPO) : 0.0025 M £H2Y2J Pu(VI) U< VI ) F AD AD DAB DA DAB DA 0.000 _ _ _ 0.000747 _ _ 0.002 0.0108 0.00326 0.00754 0.180 0.00734 0.172 0.004 0.0368 0.0127 0.0241 0.666 0.0284 0.637 0.006 0.0777 0.0280 0.0497 1.50 0.0651 1.43 0.008 0.112 0.0455 0.0665 2.36 0.111 2.25 0.009 0.158 0.0703 0.0877 2.91 0. 145 2.77 Table -26 Variation in D of PulVI) and U(VI) with [TOPO] in presence of H2Y2 Aq.Phase: 0.2H ff2SO4 Diluent: Toluene CH2V23 : 0.0025 H CTOPO] Pu(Vl) IHV DAB DA AD DAB AD 0.000 - 0.292 1.08 r '.002 0. 505 0.213 7. 46 6.38 0.004 0. 693 0.401 11 .5 10.4 0.006 0. 787 0.495 14 .3 13.2 0.008 0. 892 0.600 16 . 1 15.0 0.010 0. 962 0.670 17 .8 16.7 61 2.2.4. ION-EXCHANGE STUDIES OF URANIUM AND PLUTONIUM FROM MIXED SOLVENT MEDIA P. M. Hapara, K. V. Chetty, Rajendra Swamp and S.K.Patil Ion-exchange in solutions containing water miscible organic solvents such as alcohols, ethers, ketones etc. offer several advantages in the separation of metal ions^1^. With a view to find out the possibility of separation of plutonium and uranium using macroporous (MP) anion exchange resins from the mixed aqueous-organic solvent media, a systematic study for uranium and plutonium was undertaken. These studies were carried out using two quaternary-amine-type MP resins Tulsion A-27 (TA-27) and Amberlyst A-26 (AA 26) and a tertiary-amine-type Amberlite XE-270 (AXE-27O>. Out of these, TA-27 is manufactured locally. Because of the larger surface area, these MP resins combine the advantage of selectivity offered by high cross linkage with fast exchange rates therby making them superior to classical gel-type resins. The batch data obtained for U(V I ) and Pu(lV) using these resins under different conditions are reported here. Preliminary studies carried out on the determination of the distribution ratio of Pu(IV) using the resins and mixtures of nitric acid and a number of organic solvents such as alcohol, acetone and dioxane revealed a high adsorption of Pu(IV) on these resins with the methanol and acetone. Subsequent experiments were carried out from HNO3 + methanol mixtures. The test experiments showed that the plutonium remained as Pu(IV) in the mixed media. For experiments with U(V I ) , U-233 was used as tracer. 62 The time of equilibration studies were carried out by determining the distribution ratio as a function of time. It was observed that equilibrium was not attained even after six hours. However, an equilibration time of 3 hours was arbitrarily chosen for all the experiments. Variation of nitric acid concentration at a fixed methanol concentration. The distribution ratios for Pu(IV) and U(VI) were determined as a function of nitric acid concentration at 50% methanol concentration. From the data it was observed that the D values increased with the increasing acidity upto 3M HNO3 and then decreased in the case of all the resins used. Variation of niethanol concentration at a fixed HNO3 concentration. The distribution data for Put IV) and U Separation Factor. From the results it was revealed that the distribution ratio of U(VI ) was very low as compared to that Pu(IV). Based on the data separation factors for Pu/U were calculated at different methano1/HNO3 concentrations for all the resins and are given in Tables 27 and 28 along with the distribution ratio* data. High separation factors were obtained at 1M HNO3 towards the higher concentration of methanol. The separation factors in 63 the case of Amberlyts XE-270 were found to be the lowest as compared to the other two resins. The results indicate that a reasonably good separation of plutonium from uranium could be achieved. Column experiments are being planned to substantiate this work. Reference. 1. J.Korkisch.Progr.Nucl.Energy Ser.IX Eds.D.C.Stewert and H.A.Elion, Pergamon Press, 6., 1 (1966 I. 64 Table - 27 Variation of Distribution ratios of Pu(IV) and U(V1) between ion-exchange resins and (HNO3 + CH3OH) as a function of methanol concentration. Methanol 1.0M HNO3 3.OM HNO3 X D values for D values for Pu(lV) U(V1) S.F Pu(lV) U(VI) S.F TULSION A-27 0 53.9 1.08 50 575 2.73 211 10 54.5 0.754 72 890 3.93 226 20 47.6 0.768 62 1390 4.43 313 30 135 0.506 267 2223 6.30 353 40 147 2.32 63 3596 8.30 433 50 252 6.49 39 5620 11.0 511 60 452 2.64 171 6485 12.7 511 70 1471 5.48 268 5319 15.3 248 80 6531 9. 12 716 5545 15.3 362 90 11468 19.6 585 - - - AMBERLYST A-26 0 37.0 2.14 17 376 3.02 124 10 30.0 2.10 14 616 3.37 183 20 28.7 2.94 10 631 4. 18 199 30 82.6 0.732 113 1424 6.46 220 40 88.8 1.16 77 2118 8.96 236 50 173 1.59 109 3613 10.4 347 60 399 2.86 139 4573 12.9 354 70 1177 8.29 142 4657 14.2 328 80 4716 9.32 506 6437 14.4 447 90 13334 18.4 725 - - AMBERLYTE !KE-270 0 1.82 0.942 1.9 68 2.60 26 10 0.92 0.773 1.2 33 3.32 9.9 20 2.41 0.406 5.9 91 3.30 27 30 9. 11 0.739 12.3 131 4.73 28 40 13.0 0.666 19.5 339 5.51 61 50 26.3 0.418 63 741 6.68 111 60 61.0 1.72 35 1266 7.61 166 70 193 6.78 28 1593 7.77 205 80 760 7.81 97 2073 6.97 297 90 3375 8.02 421 _ _ „ 65 Table - 28 Variation of Distribution ratios of Pu(IV) and U(VI) between ion exchange resins and (HNO3 + CH3OH) as a function of methanol concentration. fiethanol 5.0M HNO3 7.on HNO3 X D values for D values for Pu(IVI1 UIVII S.F Pu(lV> U(VI) S.F TULSION A-27 0 2054 8.08 254 4417 11.4 387 10 2849 8.91 320 5662 11.8 460 20 4239 10.6 400 4398 10.1 435 30 5291 11.6 456 3686 - - 40 5165 12.7 407 2377 9.75 244 50 4057 13.3 305 2039 9.84 207 57 - - - 1826 9.01 203 60 3300 - - - - - 70 3076 _ _ _ AHBERLYST A-26 0 1175 7.67 153 2756 10. 265 10 2205 9. 13 241 3342 11.5 290 20 2874 9.27 310 3251 9.6 338 30 3537 100. 354 2420 - - 40 3510 12.1 290 1668 8.78 190 50 2931 12.1 242 1259 8.40 150 57 - - - 1086 8.40 129 60 2240 10. 1 222 - - - 70 2378 10. 1 235 - - - AMBERLYTE XE-270 0 1375 5.73 240 1881 7.02 268 10 1749 5.95 294 1969 7.86 250 20 2033 7.45 273 1613 6.27 257 30 2161 6.89 314 1322 - - 40 1787 7.76 230 922 6.02 153 50 1651 7.43 222 691 5.60 123 57 - - - 647 5.96 108 60 1188 5.96 199 _ 70 1150 6.59 174 _ 66 2.2.5. STUDY ON THE FLUORIDE COMPLEXES OF ACTiNIDES: MEASUREMENT OF THE STABILITY CONSTANTS OF THE FLUORIDE COMPLEXES OF U ( I V ) AND Pu(IV) USING FLUORIDE ION-SELECTIVE ELECTRODE R.M. Sawant, N.K. Chaudhuri and S.K. Patil The development of a procedure for the measurement of stability constants of the fluoride complexes of actinides using a fluoride ion selective electrode and the values obtained for pentavalent, hexavalent and trivalent actinides were reported earl ier^"31, It was also demonstrated'41 using Th ( 1 V )-f 1 uor i de system, that the high acidity required for preventing polymerisation and maintaining the oxidation states of some tetravalent actinides would not have any adverse effect on the measurement of the stability constants. The study was then extended to U(IV) and PulIV) fluoride complexes. U(IV) perchlorate was prepared by cathodic reduction. Plutonium was purified by anion exchange and Pu(IV)- perchlorate was prepared by coulometric reduction. Concentrations of U ( I V ) and Put IV) were determined by potentiometry and coulometry, respectively. For the determination of stability constants, potentials of the i luoride electrode were measured in the solutions prepared by adding suitable aliquots of standard fluoride solution to the mixture containing NaC10/4 , HCIO4 and U(IV) or Pu(IV) perchlorates using all plastic labwares. Free F~ and n-bar values, i.e. the average number of F~ ions attached to metal ion were calculated from the experimental data following the procedure reported earl ier"-' . The stability constants were obtained by nonlinear least square fitting to Bjerrum's function. When the number of constants in 67 the fitting program was varied the best fit, as measured by least value of chi-square, was obtained by four constants. The concentration stability constant; obtained in different sets are shown in Table -29 along with the values reported in literature. The 6* values reported in the literature are converted to 6 values using dj = 889 for comparison. References 1. R.M. Sawant, G.H. Rizvi, N.K. Chaudhuri and S.K. Patil, J. Radioanal. and Nucl. Chem. 89,373(1985). 2. R.M. Sawant, G.H. Rizvi, N.K. Chaudhuri and S.K. Patil, J. Radioanal. and Nucl. Chem. 9± ,41(1985). 3. M.A. Mahajan, R.M. Sawant, N.K. Chaudhuri and S.K. Patil, Radiochem. and Radiation Chem. Symp., Tlrupati, Dec. 1986. 4. R.M. Sawant and N.K. Chaudhuri, Radiochem. and Radiation Chem. Symp. Bombay, Feb (1988). 5. B. Noren, Acta Chem.Scand., 21,931(1969). 6. I. Grenthe and J. Varfeldf, Acta Chem. Scand.,23,988(1969) 7. V.N. Krylov and E.V. Komarov, Radiokhimiya, U_, 101(1969). 8. S.V. Bagwade, V.V. Ramkrishna and S.K. Patil, J. lnorg. Nucl. Chera. 38_, 1339(1976). 68 Table •- 29 Stability constants of Ui1 IV) fc Pu(lV) Fluoride complexes in perchlorate medium. Metal Ionic Strength log 6]L log B2 log fi3 log 64 Ref ion /method* U(IV) 1M (H,NaC104)/Fe 8.48 14.,66 19.51 23.92 This work • 1M (H,NaCI04)/Fe 8.45 14. 59 19.34 23.50 1 1M (H,NaCI04l/Fe 8.50 14. 72 19.59 23.75 II 1M (H,NaC104)/Fe 8.40 14. 69 19.59 23.75 f 1M (H,NaClO4)/Fe 8.47 14. 52 19.37 23.88 4M HC104/Fe 8.49 14. 62 19.53 5 4M Put(V) 1M • 1H (H,MaCI04l/Fe 7.64 14. 69 20.23 25. 94 2H HCI04/Dis 7.59 13. 51 - 8 2M HCI04/Cix 7.40 - - 7 1M HC104/Cix 7. 15 - - - 7 *Fe-F!uoride electrode, Red-Redox,Dis-Distribution, Cix-Cation exchange. 69 2.2.6 STUDIES ON THE EFFECT OF TBP AND n-OCTANOL ON THE EXTRACTION OF Pu(IV) BY MOPPA P.D.Mithapara, V.Shivarudrappa,S.G. Marathe and H.C.Jain During the studies on the recovery of plutonium from phosphoric acid waste solutions using MOPPA, back extraction of Plutonium was achieved by the addition of TBP or n-octanol and contacting with H2SO4 . Studies were therefore, carried out to understand the effect of TBP and n-octanol on the extraction of Pu(IV) by MOPPA from H2SO4 solutions. Distribution ratios (D-values) were measured as a function of TBP or n-octanol concentrations at different MOPPA concentrations. The results are given in Fig 3. It was observed that at low concentration of the synergists (when the ratio of the synergist/MOPPA was about 0.4), a maximum in the D-values was observed which decreases with further increase in the concentration. The effect of variation of D-values as a function of H2SO4 concentration at different synergist/MOPPA ratios for both TBP and n-octanol is given in Fig 4. A minimum in the D-values was observed at about 2M H2SO4 in the case of TBP and at about 5M H2SO4 in the case of n-octanol when ratio 3f the synergist/MOPPA was 100. A large decrease in the D- vaiues of the order of lO^-lO4 indicated a large negative synergism which is responsible for the back extraction of plutonium. It is reported that at high concentrations of the synergist,deactivat ion of the active hydrogen atoms by increased hydrogen bonded interaction between the synergist and the extractant is responsible for the observed negative synerg i sm. O ' A 0.00i)5FMOn-A *; I.-J'IANCL 1.1 D 0.005F.V,O?vA - TBP . " , 4 O 0.005FMCPPA + n-CCTANO:. ^ I1'2 -r—- 1.0 " • JJ,0.8 i_ J_ JO. 6 SYNERGIST CONC. 10 Fig 3. VARIATION OT L(JU U f;f- l'w(lV) WiTIi SYNERGIST'S CONCliNTHAT J'.)N . AQUEUUEUUS IMIABE : 11MM H^JOA, ORGAANIN C I'HAiiE: M()p:'A IN XYLFNH. O 0.00 5FMCTA • ^ .•:.•..r ;-r O 0.005^ MOf'I'A + 0..M- IHi- | • 0.005F MOPPA + 0 5 f r.-GCTANQI 0.005F MOPPA + G.5F n -OC1A' ;QL ! .. ._ j CONC.(M) 0. VARIATION PI- LOT, 1) ;JI- i'u(IV) W I1\i ii^^O/i fONcr.NTKAT i > JN A'i DIF;:LKL-NT SYNERCi I.'T/riOrPA RAT I OH. 70 References. 1. V. Shivarudrappa, P.D. Mithapara, S.G. Marathe and H.C.Jain, CT-7, Preprint volume, Radiochem. and Rad. Chem. Symp., Bombay(1988). 2. G.W. Mason, S. McCarty and Peppard,D.F., J.Inorg. Nucl.Chem. 24, 967<1962>. 3. T.K.S. Murthy, V.N. Pai and R.A. Nagle, Rep.IAEA-SM- 135/11(1971). 2.2.7 Solvent Extraction Studies of U ( V 1 ) by MOPPA. P.D. Mithapara, V. Shivarudrappa and H.C. Jain. Work has been initiated to study the extraction behaviour of U(VI) by MOPPA from different acidic media using 233U tracer. Distribution ratios (D-values) of U(V1) were measured as a function of extractant concentration and H+ concentration using a solution of MOPPA in toluene as the diluent. The plot of log D Vs log MOPPA concentration is shown in Fig 5. A slope of +3 was obtained with ail the acids indicating a direct third order dependancy of D values on the extractant concentration. The dependancy of D on H+ concentration is shown in Fig 6. A slope -2 indicated an inverse second order dependancy of -values on H+ concentration. Based on these results, an expression for the equilibria involved may be written as foilows : 1 + UO^" " + 3H2Y > U02(HY)^H?Y + 2H where H2Y = MOPPA. Further work is in progress in order to study the extraction in different diluents. 10 A : HCIO4 ( 1.0 M ) : HCl (1.0 M) o : HNO3 • Q 1.0 SLOPE= ~3.0 0.1 J L J ! I 1 4 6 8 10 20 Fig 5. VARIATION OF D WITH MOPPA CONCENTRATION, 20 a : HCLO4 o : HCl 10 A:HNOo : H2SO4 SLOPE = 2.0 j L 0.1 0.2 0.4 0.6 1.0 2.0 jt. G. VARIATION OP D WITH II CUHC1-NTKAT I ON . 71 2.2.8. RECOVERY AND PURIFICATION OF PLUTONIUM FROM PLUTONIUM DIOXIDE SCRAP AND MIXED OXIDES AND CARBIDES. H.S. Sharma, J.V. Kamat, N. Gopinath, N.B. Khedekar, R.K. Duggal and H.C. Jain. Plutonium dioxide scrap (280 g) was generated during the course of dissolution of large quantities of sintered plutonium dioxide in nitric- hydrofluoric acid mixture and contained 50% by weight of plutonium. Analyses indicated that the scrap material also contained 12% by weight of carbon and 1200 ppm of f1uor ide. Dissolution experiments indicated that it was not possible to dissolve the scrap material directly in boiling sulphuric acid- nitric acid mixture (1:1). Carbon was removed by heating the scrap at 900°C. Further dissolution was tried in sulphuric- nitric-perchloric acids mixture. The scrap in lots of 10 g was dissolved in boiling 15M nitric acid,9M perchloric acid,and IBM sulphuric acid mixture (5 ml + 5 ml + 40 ml) under reflux. Plutonium thus recovered was precipitated as oxalate in lots of 50g. Plutonium oxalate was finally converted to oxide by heating at 500°C. About 160 g of plutonium from scrap was recovered. About 30 g of plutonium was recovered by dissolving mixed oxide and oxidised mixed carbide material in 15M nitric acid and nitric - perchloric - sulphuric acids mixture respectively. The recovered plutonium was converted to oxalate and finally to oxide form. 72 2.2.9. RECOVERY AND PURIFICATION OF PLUTONIUM FROM ANALYTICAL WASTE. R.B. Manolkar, S.P. Hasilkar, Keshav Chander and S.G. Marathe. 20g of plutonium was recovered from different types of analytical wastes. The plutonium present in about 20 litres of waste was precipitated as hydroxide. After dissolution of the hydroxide in HN03; plutonium was purified by anion exchange technique. Plutonium in the purified solution was precipitated as oxalate which was heated at 500°C to get pure 2.2. 10. SEPARATION OF PLATINUM FROM PLUTONIUM-PLATINUM SOLUTIONS S.P. Hasilkar, Keshav Chander and S.G. Marathe. It was felt necessary to have a simple procedure by which platinum associated with plutonium in solution could be separated in pure form without significant alpha contamination. Very little is known about the separation of platinum from plutonium and other radioactive solutions. Studies were therefore carried out to see whether separation, of platinum could be achieved through its precipitation in a suitable med ium. The redox potentials data indicated that Tit III) can reduce platinum present in any oxidation state, to platinum metal. Therefore to the Pt-Pu solution, Ti(II I) was added in HC1 medium. Platinum was precipitated as metal. The experiments have shown that the metal precipitation is quantitative and very small amount of plutonium is associated with the precipitate. Quantitativeness of the metal precipitation as 73 well as decontamination from plutonium was checked as follows: about a milligram of platinum was irradiated in the reactor APSARA under the neutron flux of 1012/cm2/sec. Known activity (107dpm) of 197Pt(18.3h> was taken and added in solution containing varying amounts of platinum (5-25mg). After the precipitation, activity in the supernate as well as in the precipitate was assayed by the peak area measurements corresponding to 191.4 kev gamma ray energy. Negligible counts were obtained in the supernate and nearly all the activity added was carried in the precipitate. Also metal precipitate was weighed on a preweighed Whatman filter paper No. 541. The observed weights of platinum metal precipitate compared well with the expected value. The platinum metal precipitate was washed repeatedly (4-5 times) with 1M - 2M HNO3 and then employing ultrasonic cleaning. The decontamination factor of 107, as observed by alpha- scintillation counting, could be obtained. This scheme of separation is proposed to be employed for the solutions containing gram levels of Pt-Pu. Dissolution of the precipitate in HC1 and reprecipitation of the same was observed to give high decontamination from plutonium. 2.2.11. RECOVERY OF HEXAMETHYLENE TETRAM1NE AND UREA FROM WASTE SOLUTION GENERATED FROM THE AMMONIA WASHING OF U03 GEL PARTICLES PREPARED BY INTERNAL GELATION PROCESS. S.K.Mukerjee, J.V. Dehadraya, T. U.Vithairao, V.N.Vaidya and O.D. Sood. Work on extraction of HMTA and urea from wash solution containing HMTA, urea, NH4NO3 , HCHO and methylol urea has been initiated. Since NH4NO3 and HMTA, when present in high concentration react vigorously (explosive reaction) on hsating, 74 waste volume cannot be reduced just by distillation. Hence it was decided to remove nitrate by ion exchange,using DOWEX 1x4 anion exchange resin. Resin in the OH form was used to avoid contamination from any more constituent. Resin was regenerated by IN NaOH. The eluent containing HMTA, urea, HCHO, methylol urea and NH4OH was distilled under vacuum. Distillate contained NH4OH and some HCHO. The concentrated solution was analysed for free and bound formaldehyde by Iodometric estimation. From the total amount of formaldehyde present in the wash solution it was found that about 94% of HMTA that was initially present in the feed solution remains unused. About 2% of HCHO was present in the concentrated wash solution. Efforts to remove HCHO by vacuum distillation failed,as removal of HCHO was accompanied by hydrolytic decomposition of HMTA and the amount of HCHO in solution remained about 2%. Work is being carried out. to bring down HCHO further. Estimation of HMTA and Urea in Solution. HMTA solution in water gives an analytical peak in UV region at 197nm, but urea peak at 191nm causes interference. However, at 205 nm, HMTA can be easily estimated in presence of urea. Urea is estimated from the knowledge of total nitrogen and nitrogen from HMTA. 75 2.2.12. STUDIES ON THE BEHAVIOUR OF PLUTONIUM IN ALKALINE MEDIA A.V. Kadam, M. Ray, I.C. Pius, M.M. Charyulu, C.K. Sivaramakrishnan and S.K. Patil. The data on the distribution of Put IV) and Pu(VI) between AJ2O3 or Amberlyst A-26 (MP) and carbonate medium had indicated feasibility of removal of plutonium from aqueous alkaline waste streams. Preliminary column experiments were carried out to oi explore the feasibi1ity^removal of plutonium from carbonate medium. Two ion exchange columns were prepared, one with 5ml of chromatographic grade A 1 2O3 and the other with 5ml of a strong base anion exchange macroporous resin Amberlyst A-26(MP). Feed solution containing 9.7ng/ml of piutonium(IV) in 0.25M carbonate was prepared and its pH was adjusted to 9.5. This feed solution was passed through above two columns separately at a flow rate of 0.5 ml/min. Samples from the column effluents were collected at regular intervals and plutonium content was determined radiometrically. Loading was continued till 10% plutonium breakthrough was obtained in the effluents. The 10 percent plutonium break-through capacity was found to be 2.5 mg/ml for A12O3 and 0.6 mg/ml for Amberlyst A- 26 (MP) under the experimental conditions used. Similar experiments were carried out with Pu(VI) in carbonate medium with Amberlyst A-26 (MP). A feed solution of 100ng/mI of plutonium (VI) in 0.25M Na£C03 at pH-12 was passed through the anion exchange column of 5 ml bed volume at the flow rate of 0.5 ml/min until 25% plutonium breakthrough was reached. Potassium persulphate was used as holding oxidan.t for Pu(VI). The 25% plutonium breakthrough capacity was found to be 12 mg/ml of resin bed indicating the feasibility of removal of plutonium from carbonate medium by using Amberlyst A-26 res i n. 76 2.3 PROCESS CHEMISTRY 2.3.1. STUDIES ON THE EVALUATION OF DIFFERENT ANION EXCHANGE RESINS FOR PLUTONIUM PROCESSING D.G. Phal, S. Kannan, Kum. N.N. Mirashi, R.D. Bhanushali, V.V. Ramakrishna and S.K. Patil. In continuation of the work on the evaluation of different resins for plutonium processing1 *-1, some more resins were procured and the results of the investigations carried out with these are reported here. Measurement of Distribution Ratios The experimental procedure adopted for measuring the distribution ratio (D) values is the same as described earlier. Among the macroporous (MP) resins studied earlier, the indigenously available Tulsion A-27 (MP) with the particle size in the range of 0.3 to 1.2 mm, (15-50 mesh) gave the best performance with reference to plutonium breakthrough capacity as well as separation of plutonium from uranium. Since this resin gave a higher plutonium breakthrough capacity when the loading was done at a higher temperature and/or a longer residence time; it was decided to study the variation of D values of Put IV) with time and temperature. The results obtained are given in Table -30. It is seen that the D values for similar equilibration times at two temperatures are quite close to each other suggesting that the better kinetics is responsible for a better loading at high temperature rather than any change in equilibrium conditions. Anticipating that the same resin (Tulsion A-27, MP) with a smaller particle size may give an overall improved performance, the resin in 50-100 mesh size was procured and studies were conducted with it. The U(VI) D values were obtained for this 77 resin as well as the normally used Dowex 1x4 gel type resin and they are compared in Table - 31 with those of Tulsion A-27, MP <15-50 mesh) obtained earlier. It is seen that practically there is no change in the LI (VI ) D values with tha change in the particle size of the resin. Though the gel type resin gave slightly lower U(V1) D values, its performance from uranium separation point of view was already demonstrated^1^ to be poor as compared with the macroporous resins. The results on a comparison of the distribution ratios of Put IV) obtained with A-27(MP) of two mesh sizes are given in Table - 32. It is seen that there is a marked improvement in the equilibrium D values with smaller particle size resin. The distribution ratio data for Pu(IV) obtained with some more resins are given in Tables 33 and 34. All these resins seem to have a good absorption power for Pu(lV) except for Tulsion A-12X. However the local gel-type resin Tulsion, A-35 (gel) gave the highest D values and further study is planned with the same. Study of the Plutonium breakthrough capacity and elution behaviour of Tulsion A-27, MP (50-100 mesh) resin. Using a 10 ml resin bed of this resin and a feed solution with the composition of 1 g/1 of plutonium and 7 g/l of uranium in 7M nitric acid the 10% plutonium breakthrough capacity was found to be 102 g/l with a residence time of 15 minutes and 110 g/l with a residence time of 30 minutes. These values are much higher than the corresponding figures of 37 g/I and 58 g/l with this resin with a bigger particle size[1^. With 0.5 M nitric acid 99% of the ptutonium held on the resin could be eluted in 5 bed volumes. Though these features were found to be very satisfactory for plutonium recovery the studies revealed that the resin with a smaller particle size crumbled 78 after a couple of runs to become like powder, thus preventing the flow of liquid through the resin bed. Efforts are on to procure a physically and more stable product otherwise equivalent to this resin. Studies on Plutonium recovery by anion exchange from HF containing nitric acid solutions. Addition of HF to nitric acid was found to improve the leaching plutonium from the gloves and PVC sheet used in the boxes used for fabrication of piutonium-bearing fuels. It is known that the presence of fluoride ion decreases the distribution ratio of Pu(IV) between nitric acid and anion exchange resin thereby reducing the plutonium breakthrough capacities of the resin columns. Addition of Alt I I I), which complexes flouride ion effectively, is reported to help in arresting the fall in the plutonium breakthrough capacities. Using three macroporous and one gel-type anion exchange resins experiments were conducted to study the feasibility of plutonium recovery the results of which are presented here. Distribution ratios of Put IV) from 7M nitric acid, in the absence and the presence of HF were measured with alI the four resins and the data obtained are given in Table - 35. As expected, the presence of HF decreased the values due to the complexing of Put IV) by fluoride ions in aqueous solution. Addition of Alt I II) is seen to improve the D values of Put IV). From the data it is seen that all the four resins studied may, more or less, behave similar to one another for plutonium recovery. The breakthrough capacities for Put IV) for these resins were obtained using 10 ml beds and a feed solution containing 7.5M nitric acid, 0.31 mg/ml of plutoniuni and 0.03M each of HF and aluminium nitrate with residence time 15 minutes. The plutonium breakthrough (bt) capacities obtained under the experimental conditions were: 14.7mg of Pu/ml resintlOX bt) for 79 Amberlite XE-270 MP.weak base,27.2mg of Pu/ml resin (2% bt) for Amberlyst A-26 MP strong base,3O.5mg of Pu/ml resin (2.5% bt) for Tulsion A-27 MP strong base, and 53.9 mg of Pu/ml resin (0.5% bt) for Biorad AG-1X4, strong base, gel type resin. The loadings for the last three beds coutd not be continued upto 10% bt due to the problem of choking of the resin beds, apparently due to the presence of some fine solids in the feed solution. Such a problem was not encountered with the resin AXE-270 and this may be due to its different packing characteristics. As a result this resin was chosesn for the recovery purpose inspite of its poor capacity as compared with others. Using a 500 ml resin bed of AXE-270, about 4.3 g of plutonium in 15 litres of nitric acid, containing HF, was quantitatively recoverd. Reference. 1. FCD Annual Report 1986. 80 Table -30 Variation of the distribution ratio (D) of Put IV) with equilibration time [HN03 ] = 7.5M Initial Aq [PuUVI] = Spg/ml Resin : Tulsion A-27 IMP), 15-50 mesh Equi1ibrium D of Pu ( IV) at Time, min. 25° C* 40°C 10 331 241 20 646 564 30 899 7 94 40 - 965 50 1090 1063 60 1213 1140 * Reference [1]: Table 30. Table - 31 Variation of the distribution ratio(D) of UIVI; with nitric acid concentration. Initial Aq CUIVN = lOjjg/m! Time of equilibration = 3 hours. CHNO3] D of U(VI) with Tuls ion A-271MP) Tulsion A-27IMP) Dowex 1x4(gel) (15-50 mesh) * (50-100 mesh) (50-100 raesh) 2 1. 2 0.97 _ 3 3. 1 3.6 - 4 5. 6 4.9 - 5 8. 6 8.8 6.3 6 11. 0 11.0 6.8 7 11. 0 13.0 10.0 8 11. 0 11.0 8.2 9 9. 2 9.2 5.9 10 7. 6 10.0 5.9 « Ref. El], Table 27. 81 Table - 32 Variation of the distribution ratio (D) of Pu(IV) with nitric acid concentration Initial Aq [PutlVIl = 5>»g/ml Time of equilibration = 3 hours. Resin-Tulsion A-27(MP) £HN03] D of Pu(lV) with the mesh size (15 - 50)« (50-100) 1 13 35 2 58 211 3 135 531 4 297 960 5 605 1510 6 1040 2280 7 1610 2700 8 1860 2860 9 1610 2690 10 1150 2130 * Ref. [11, Table 28. Table - 33 Variation of the distribution ratio (D) of Pu(lV) with nitric acid concentration Initial Aq tPu(IV)] = 5jtg/ml Resin: Strong base, Time of equilibration = 3 hrs. Type I* D values of Pu(IV) with the resins [HN03] M Duolite Duolite Duolite Amberlyst Tulsion A-113 A-161 A-101D A-26 A-35 (MP) (MP) IMP) (HP) (Gel) 1 14 30 15 36 26 2 46 87 47 116 93 3 109 291 145 449 332 4 262 390 330 530 659 5 592 1050 586 832 1310 6 1230 1290 1150 1050 1980 7 1420 1550 1690 1780 2430 8 1600 1630 1850 2210 4830 9 1090 1470 1790 1950 3690 10 734 923 846 1580 1280 + Resin Type 1* _> R-N (CH3>3 82 Table - 34 Variation of the distribution ratio D values of Pu(IV) with the resins CHNO3] Strong base, Type II Ueak-base . n - - Duolite Duolite Tulsion Duolite A-162 A-102D A-12X A-368 1 13 11 3.7 32 2 68 42 6.5 119 3 153 102 19.0 244 4 230 205 25.0 310 5 381 448 37.0 536 6 565 742 56.0 585 7 723 U40 79,0 703 8 838 1330 79.0 818 9 684 982 73.0 535 10 . 578 408 53.0 424 + Resin Type 11 —> R - N (CH3>2CH20H Table - 35 Distribution ratios of Pu(IV) from 7M nitric acid in the presence of HF and AHN03)3 . Time of equilibration = 3 hours; Temperature = 25*C Resin [HF1=O tHF]=0.025M tHFJ=0.05M CHFJ=0.05M £HF)=0.025M {AI(NO3)3J £A1(NO3)3 j =0.05M =0.025H Amber 1ite 1120 738 726 792 1010 XE-270CMP) Amberlyst 1680 880 823 934 1250 A-26(MP) Tulsion 2140 1240 1160 1180 1930 A-27(MP> Biorad 2560 1140 948 1340 1330 AG 1x4(GEL) 83 2.3.2. RECOVERY AND PURIFICATION OF PLUTONIUM FROM FUEL FABRICATION SCRAP FOR RECYCLING M.M.Charyulu, N.K.Chaudhury, D.R.Ghadse, A.R.Joshi, A.V.Kadam, U.M.Kasar, M.A.Mahajan, M.S.Oak, S.M.Pawar, I.C.Pius, R.K.Rastogi, M.Ray, V.B.Sagar, R.M.Sawant, C.K. Sivaramakrishnan and S.K. Patil. Scrap generated during the fabrication of plutonium bearing nuclear fuels has to be necessarily processed for recovery and recycling of valuable plutonium. During the fabrication of uranium - plutonium mixed carbide fuels by carbothermic reduction, part of the plutonium volatilises and gets deposited on various components of the furnace. The quantities of plutonium volatilised are significant, particularly in case of FBTR fuel, which has high (70%) plutonium content. The plutonium thus deposited being highly reactive, it is normally leached with a non-oxidising medium like HC1. As a part of the plutonium recovery and recycle operation for fuel fabrication, large volumes of HC1 solution containing 4-50 g/1 of plutonium along with U, W, Cr, Mo etc. received from Radiometa11urgy Division was processed for plutonium recovery/purification. Such large volumes of HC1 containing significant quantities of Plutonium have not been encountered in our laboratory in the past, and a suitable method had to be evolved for recovery and purification of this plutonium. Large volumes and the corrosive nature of the medium demanded that the method to be developed should aim at (a) minimum operations (b) minimum waste generation and !c) the product of required purity. BA 2.3.2.1. Preliminary trial experiments. Direct precipitation of oxalate from HC1 medium. Preliminary trial experiments were carried out using ThMV) as a stand-in for plutonium, to develop a suitable method for processing the large volumes of HC1 medium. Since the final product required for recycling is PuO£ produced through the oxalate route, direct precipitation of Th(IV) as oxalate from HC1 medium and its subsequent conversion to oxide was attempted. Analysis of ThO2 thus produced indicated the presence of 1000-1200 ppm of chloride in the product which is rather too high to accept. A hydroxide precipitation and dissolution of hydroxide precipitate in nitric acid was considered desirable to remove the bulk of chloride prior to oxalate precipitation. Precipitation of Plutonium oxalate from HNO3 med ium. A trial experiment with 2 litres of HCI solution containing about 20g of plutonium was therefore carried out. In this, the plutonium was precipitated as hydroxide using sodium hydroxide,, and the resulting plutonium hydroxide was dissolved in nitric acid after several washings of the precipitate with dilute ammonium nitrate solution. Having thus eliminated the bulk of chloride, oxalate precipitation was carried out in a stainless steel container. Prior to addition of H2O2 for adjusting the oxidation state of plutonium and addition of oxalic acid for precipitation, the solution was heated to 55- 60°C. During this heating process, the solution started reacting with the container, as seen by the effervescence, probably due to the presence of chloride ions, indicating the non- compatibility of stainless steel vessel with the plutonium solution. 85 Precipitation of oxalate from HNO3 medium after double hydrox ide preci pi tat ion. In view of this, a third trial experiment was carried out wherein after the initial conversion of the medium from HCi to HNO3, plutonium was precipitated as oxalate in a plastic container kept in a waterbath which was heated to have the temperature of plutonium solution to «* 55-60°C during oxalate precipitation. The oxalate produced was washed, dried and finally converted to PuG"2 by heating at 500°C in a stream of oxygen. Analysis of the PuO2 still showed the presence of 260 ppm of chloride. In yet another experiment, one more hydroxide precipitation of plutonium from the HNO3 medium was incorporated for further elimination of chloride impurity before oxalate precipitation in a plastic container. The oxide obtained from this oxalate still showed the presence of 150 ppm of chloride and also 5 ppm of boron, indicating need for further purification to meet the specification of PUO2 powder used for fuel fabrication. In view of this, an anion exchange purification process for plutonium was incorpoated prior to oxalate precipitation. 2.3.2.2 Anion exchange purification of plutonium. For this, the entire HC1 solutions were converted to HNO3 medium through plutonium hydroxide precipitation and redissolution of the resulting hydroxide in HNO3 in several batches. Concentration of plutonium in 7M HNO3 medium, varied between 30-60g/litre along with l-10g/litre of uranium. The use of conventional gel type anion exchange resins like Dowex 1x4 for purification of plutonium from such solutions has certain limitations due to rather slow kinetics of plutonium uptake by the resin. Literature survey indicated that the raacroporous resins have faster kinetics compared to gel type resins, resulting in better uranium washing and better 86 plutonium elution. In view of this, a strong base macroporous anion exchange resin Amberlyst A-26 was considered more suitable for purification of plutonium from solutions containing fairly high concentrations of plutonium. The literature survey revealed that this resin has not been explored so far for large scale plutonium purification. Ion exchange colum set-up. Since large volumes are involved, a large size glass column of dimensions 100mm dia & 350mm long, fitted with a G-l glass frit at the bottom and a 1000 ml reservoir bulb with a ground glass stopper (B-34/35) at the top was fabricated and set up. Two litres of Amberlyst A-26 resin was packed inside the column. In view of the long continuous operation envisaged most of the •operations like lifting of the feed/wash/ eluant etc. from the fsed bottle to the column head were automised using suction. The entire line from the closed feed/wash eluant container to the column head down the column connection to the receiving bottle were all airtight and one single air suction point connected to the final receiving bottle provided the requirtd lift of the solution and smooth running of the column. A schematic diagram of the entire set up is piven in Fig 7. A feu leaktight valves incorporated at the bottom of the glass column and the suction tubes provided good control of flow rates. Since the feed solution had some undissolvcd material, a cylindrical tube packed with quartz wool was inccrporated between the feed bottle and the column head which r •? I..-, i ncd all the solid particles, passing only the clear plutonium solution to the ion exchange column. Column Operations. Chloride form of the resin was converted to nitrate form by continuous washings with HNO3 until the effluent was entirely free from chloride. The entire plutonlum was processed in six -I.D. » 10 cms. h = 26 cms. -SAMPLING SYSTEM I.D. = 7cms. K = 22 cms. GLASS WOOL TO SUCTION^ GI-FRIT PEED SOLUTION Gi-FRIT Fig 7.SCHEMATIC DIAGRAM OF ION EXCHANGE COLUMN SET-UP. 87 runs, one after the other using the same resin, after reconditioning after each run. Plutonium in 7tt HNO3 feed solution was adjusted to Pu( IV ) using NaNC>2 and was confirmed by its extraction by 0.2M HTTA from 1M HNO3 . Plutonium solution was carried at flow rate of 1 litre per hour. Loadings were continued until the entire bed of ion exchange resin was nearly loaded to its capacity as inferred by the analysis of the column effluent for plutonium content. Washings with 7M HNO3 were carried out at the rate of 1 litre per hour. A total of 6 bed volumes (12 litres) washings were collected. Elution of loaded plutonium was carried out with 0.5M HNO3 at the rate of 0.25 litres per hour (8 hours per bed volume). Forecuts at the starting of elution were col lee ted separately, since most of the plutonium was eluted in the first two bed volumes. Further elution (from 3rd bed volume onwards) were carried out at the rate of 0.5 litres per hour. Total S bed volumes elution was used in each run. Samples were collected at frequent intervals during loading, washing and elution. Loading and washing samples were analysed by radiometric assay and the concentrated fractions of the eluted plutonium were analysed by redox t i triuetry^ One entire column operation right from starting to the end of elution took about 60 hours of non-stop operations followed by another 15 hours for analysis of a!1 the samples. This is in addition to the time taken for preparation of the column for each run (conditioning) as well as for the disposal of large volumes of liquid waste generated. 88 Performance characteristics of the ion exchange column. At the end of the first run with the Amberlyst A-26 anion exchange column, the following observations were made. Amberlyst A-26 (M.P) resin gave satisfactory results with 75-80g of plutonium per litre capacity under experimental cond i t ions. Channelling was observed, but this did not result In loss of appreciable amounts of plutonium to loading effluents. A high degree of decontamination from uranium was achieved in the product as confirmed by mass spectrometric analysis of the feed and product solution for uranium. Loading and washing effluent contained a total of 3 g of Pu alongwith all Uranium in 16 litres and had to be processed for further recovery of plutonium. Plutonium elution behaviour was excellent with about 355b of piutonium being eluted in just 2 bed volumes. However, the remaining 5X of plutonium needed 3 more bed volumes for complete elution of plutonium (Fig 8) Spectrographic analysis indicated that the eluted plutonium is free from metallic and boron impurities. The concentration of plutonium in the product solution was suitable for precipitation of plutoniuni as oxalate. Incorporation of weak base Macroporous anlon exchange resin, Amberllte XE-27O column In series. As mentioned above, the plutonium escaping to the effluents from the 1st Amberlyst A-26 resin column during plutonium loading had to be recovered. A weak base macroporous anion exchange resin, Amberlite XE-27O was chosen for this, which had shown favourable recovery characteristics, earlier for solutions containing low plutonium concentrations and higher Uranium concentrations. A 220 mm long, 70 mm dla glass column 100 h ELUTION RATE 0.25L/Hr 80 - 94. 87 98. 22 99 38 99 89 100 ( FOR FIRST 2 BED VOLUMES AND 0.5L/Hr 60 - SUBSEQUENTLY 40- 3 46. 33 UJ 20- 1.32 FORE .CUT 1 VOLUME ELUTED (IN TERMS OF BED VOLUMES) OV - Q 40 UJ 45.01 % 48.547. t— 30 Z> 7B.77g 84. 95 g in 20 - 3.35 7. 1.16 7o 0. 517- 0. 11 7. 10 5.87g 2.04g 0. 89g 0 21g 1.32 r\ nd rd fh th tK )StBV 2 B.V. 3 B.V 4 B.V 5 B.V 6 B.V VOLUME ELUTED ( IN TERMS OF BED VOLUMES) Fig 0. ELUTION OF f'u FROM AMBEKI-YST A-2b(MD COLUMN (BED-VOLUME : 2 LITRES). 89 fitted with * 800 mi of Amberlite XE-270 resin was incorporated in series to the Amberlyst A-26 column after the 1st run. The effluent from the 1st column passed through the second column at the same rate and the concentration of plutonium in the effluent of second column was low enough to permit their direct disposal. During the elution, these two columns were disconnected and plutonium from each column was eluted separately. Dilute plutonium solution', in the tailing part of these columns were used as the eluant in subsequent runs to get as much concentrated plutonium as possible which will be useful for the subsequent oxalate precipitations. From the results of all the six runs, the following observations could be made. Inspite of channelling observed in all the runs, by careful control, loss of significant amounts of Pu in the effluents could be prevented. Capacity of the Amberlyst A-26 resin for plutonium loadi1^ remained nearly the same inspite of repeated use. Elution characteristics were also retained by this resin after repeated use. Resultant concentration of plutonium in the first two beJ volumes of each run were high enough (> 30 g Pu/litre) for direct precipitation of Pu(IV) oxalate. Elution data of all the six runs are summarised In Table -36. Thus, the entire plutonium in HC1 medium receivesd from RMD was processed as detailed above. The set-up and operations which was more or less 1 ike a plant run, was continued In a single glove box. This successful operation has shown for the first time the feasibility of anion exchange recovery and purification of plutonium in large quantities using a macroporous anion exchange resin Amberlyst A-26. 90 References. 1. V.V.Ramakrighna and S.K.Patil, Paper IT-4, Radiochemistry and Radiation Chemistry Symposium,Bombay,Feb.1988 2. L. Drummand and R.A. Grant, Talanta, 13., 477 (1966 ). Table - 36 ELUTION DATA FOR AMBERLYST A-26 IMP) RESIN RUNS,BED VOLUME 2 LITRES RUN No. BED VOLUME 1 2 3 4 5 6 FORECUT Volumedit) 1.2 1.3 1.5 1. 4 1.2 1.5 XPu eluted 0.33 1.32 0.29 0.36 0.43 0.05 FIRST Volumedit) 2.0 2.0 2.0 2.0 2.0 2.0 XPu eluted 44 .29 45.01 48.96 27.65 42.e8 39.30 (44,.62) (46.33) (49.25) (28.01) (43.31) (39.35) SECOND Volumedit) 2..05 2.0 2,.0 2.0 2.0 2.0 XPu eluted 52.,82 48.54 47.. 34 3" 4 i),. 18 (97.,44) (94..87) (96,,59) (67,.25) (86.54) (88,.53) THIRD Volumedit) 0.,7 2..0 0.,5 2,,0 2.0 2..0 XPu eluted 1.93 3,.35 2.,02 30,.00 11.42 9. 22 (99. 37) (98.,82) (98. 61) (98..05) (97.96) 197.,751 FOURTH Volumedit) 2.0 2.0 2.0 2. 0 2.0 2.0 XPu eluted 0.55 1.16 1.11 1.56 1.06 1.48 (99. 92) (99. 38) (99. 72) (99. 61) (99.02) (99. 23) FIFTH Volume!1 it) 1.75 2.0 2.0 2.0 2.0 2.0 XPu eluted 0.08 0.51 0.24 0.27 0.50 0.49 (1001 (99. 89) (99. 96) (99. 88! (99.52) (99. 72) SIXTH Volumedit) 2.0 2.0 2. 0 2.0 2.0 XPu eluted 0.11 0.04 0. 12 0.48 0.28 (100) (100) (100) ( 100) (100) Figures in brackets refer to cumulative X of plutonium eluted. 91 2.3.3. STUDIES ON THE RECOVERY OF U-233 FROM PHOSPHATE CONTAINING AQUEOUS WASTE USING DBDECMP AS EXTRACTANT. V.B. Sagar, M.S. Oak, S.M. Pawar, C.K. Sivaramakrishnan and S.K. Patil. 2.3.3.1. 1 introduction Uranium-233 bearing nuclear fuels are routinely analysed for their uranium content by Davies & Gray method '*'. The analytical waste generated, contains phosphoric acid alongwith metallic impurities such as Fe, V, Mo etc. Uranium-233 being a valuable fissile material, it has to be recovered from such wastes for recycling. Conventional methods, such as TBP extraction or anion exchange can not be used for recovery of U-233 from such solution. During the last few years, various neutral bifunctional organophosphorus compounds have been synthesized'2) and some of their physical and chemical properties have been characterized. Schulz and Me 1 ssac^"6 ^ following an earlier suggestion of Siddall'-^ demonstrated that certain neutral bifunctional organophosphorus reagents are particularly suitable for removal and recovery of actinide ions from various highly acidic nuclear fuel cycle waste solutions. Recently the CMP group of compounds (dialkyl N,N-diethyl- carbamoyl- methyl phosphonates) are being explored for their suitablity as the extractants for actinides [8]. The extraction of selected transplutoniurn (III) and lanthanide (III) ions and of Th(IV) and U(V I ) from aqueous nitrate media by DHDECMP (Di-hexyJ N, N diethyl carbamoyImethyI phosphonate) has ben reported t9.10]. xne extraction of AmlNl), Pu(lV) and U(V1) by DBDECMP from aqueous nitric acid solutions haa also been 92 reported ^ 11 ] . Present work describes the use of DBDECMP for the extraction of Uranium(VI) with a view to develop a method for the recovery of U from an analytical waste containing HNO3, H2SO4 and H3PO4 . 2.3.3.2 Exper imental Materials: Uranium-233 was purified by anion-exchange from chloride medium,fo11 owed by TBP-extraction method. The radiochemica1 purity of U-233 was then ascertained by alpha spectrometry using a silicon surface barrier detector coupled with a 4K analyzer. DBDECMP was obtained as 99% pure compound from Columbia Organic Chemicals Co. Inc. Camden, SC.USA and was used as such. All other chemicals used were of A.R. Grade. Procedure: Stock solutions of known concentrations of HNO3, H2SO4 and H3PO4, were prepared by dilution of concentrated acids and standardizing them by tirating with standard NaOH solution. DBDECMP solution of desired concentration was prepared by diluting measured volume with xylene. Uranium-233 stock solution was prepared in 1M HNO3 . Extraction experiments with U-233 (VI): 5 ml each of aqueous solution of desired composition containing «* 5 ng/ml of U-233 (VI) and DBDECMP solution in xylene (unequi1ibrated) were pipetted in ground glass stoppered tubes. The tubes were equilibrated for 30 minutes using a mechanical shaker at 25°C(±0. 1°C ) . At the end of equilibration, the phases were allowed to settle and appropriate aliquots from both the phases were withdrawn to measure the U-233 concentration radlometrleally. The 93 concentrations of U-233 in both the phases were determined by alpha- liquid scintillation counting using a dioxane based scinti11ator. Extraction of U-233 from analytical waste. The analytical waste produced during Davies and Gray method has an approximate composition: HNO3-O.5M + H2SO4-IM + H3PO4-3M, U-233 0.2 - 0.4 rag/ml, Fe 2mg/ml, Mo 0.15mg/ml. It was adjusted in HNO3 concentration to 5M by adding appropriate amount of concentrated HNO3. 100 ml of this aqueous feed solution, with approximate composition H2SO4 - 0.5M, H3PO4 - 1M and HNO3 - 5M, was contacted 3 times with 50 ml of 30% DBDECMP-xylene (Preequi1ibrated with 8M HNO3) each time. 150 ml of the organic phase was collected together and washed twice with 50 ml of distilled water each time to remove the extracted nitric acid. Uranium-233, from this organic extract was back-extracted with 100 ml of 1M!NH4)2CO3 . Initial and final concentration of U-233 was determined by alpha-liquid scintillation counting. The concentration of the impurities was assayed by spectroscopic methods. Extraction of impurities. A synthetic mixture of U, Fe, V and Mo was prepared in an acid mixture of 0.5M H2S04 + 1M H3P04 + 5M HN03 . The concentrations of these metal ions were in the same range as expected in the actual analytical waste. The extractions, washings and the back extractions were carried out following the same prjcedure as is described above. The extraction studies were carried out in the presence and in the absence of uranium. The analysis of the impurities was carried out by spectroscopic methods. 94 Absorption spectral studies. Milligram amounts of natural U(V1) were extracted into 30% DBDECMP- xylene from 5M HNO3 and also from an acid mixture containing 0.5M H2SO4 + 1M H3PO4 + 5M HNO3 . The spectra of the organic extracts containing U(VI) were recorded on a Beckman Du-7 recording spectrophotometer using the respective organic solutions preequi1ibrated with the corresponding aqueous phase without U < V I ) as blanks, with 1 cm path length quartz cell. 2.3.3.3 Results and Discussion. Distribution ratio of I) (V1 ) from various aqueous media: The data obtained on the extraction of U(VI) by 10% DBDECMP- xylene from varying concentrations of nitric acid (0.5M to 5M) are given in Table - 37. It is seen that the distribution ratio of U(VI) increases with nitric acid concentration and U(V I ) is almost quantitatively extracted when the nitric acid concentration is 2 3M. The data obtained for the extraction of U(VI) from an aqueous phase containing 2M HNO3 + varying concentrations of H2SO4 ( . 25M to 2M) are included in Table -37. It is seen that the extraction decreases with increase in H2SO4 concentration. This may be attributed to stronger complexing of U(VI) with sulphate. The data obtained on the extraction from 0.5M H2SO4 varying concentrations of HN03(0.5 to 5M) are also included in Table - 37, which show that the extraction of U(V1) can be enhanced by increasing the HNO3 concentration of the aqueous medium. The data were also obtained for the extraction of U(VI) from 0.5M H2S04 + 3M HNO3 + Varying concentrations of H3PO4 (Table - 38), which show a further decrease in the extraction of U with increasing H3PO4 concentration. This again may be attributed to stronger complexing of U(V1) with phosphate. In order to optimise the conditions for the quantitative extraction of U ( v11 ) from a 95 mixture of H2SO4 , H3PO4 and HNO3 , D values of U ( V I ) were determined from an aqueous medium containing 0.5M H2SO4 + 1M H3PO4 + HNO3 concentration varying from 1M to 5M. The data obtained are also listed in Table - 38 and these indicate ~ 50% extraction of U(VI) from 4M and above with 10% DBDECMP - xylene.By increasing the concentration of DBDECMP to e, 30% and by resorting to multiple extration, if required, it was felt that U(VI) could be quantitatively extracted. Extraction of U-233 from actual analytical waste. The approximate composition of the actual analytical waste produced during Davies and Gray method is mentioned earlier. HNO3 concentration was adjusted to 5M by addition of concentrated HNO3 before extraction of U-233. Results of the recovery of U-233 from the adjusted analytical waste are presented in Table - 39. It is seen that "95% of U-233 is recovered by this method. Extraction of impurities. Fe, V and Mo are the major impurities introduced in the system during the analysis. The extraction of these impurities by using the above extraction procedure was investigated to find out the decontamination achievable for uranium. For this, the extraction of Fe, V and Mo from a synthetic mixture was studied in the presence and in the absence of U(V1) and also from the analytical waste generated. From the organic extracts, Fe,V,and Mo were back-extracted and their concentrations were measured by spectroscopic method. It was observed that the decontamination factor for Fe and Mo was better than 10 and that for V was 25. It can therefore, be inferred that by using the present method, U-233 can be recovered and purified from the impurities present. 96 Absorption spectral studies: Absorption spectra of the organic extracts of U(V1) extracted from 5M HN03 and from 5M HNO3 + 0.5M H2Sn4 + 1M H3PO4 were recorded and are compared in Fig 9. The absorption spectra are identical with respect to absorption maxima as wel1 as molar absorption coefficients (39 and 38 respectively for extraction from 5M HN03 and from 5M HNO3 + 0.5M H2S04 + 1M H3PO4 ) indicating that the U(VI) species extracted from the acid mixture is probably the same as the one extracted from only nitric acid. Spectra are also identical with that reported in the 1 iterature^". Hence it can be inferred that only U02 References 1. W. Davies and U. Gray, Ta 1 anta, H_, 1203(1964 ). 2. W.W. Schulz and J.D. Navratil Sep. Sci and Tech. 19., pp927-94i (1984-85). 3. W.W. Schulz, U.S.ERDA Report ARH-SA-203, Atlantic Richfield Hanford Co, Richland, Washington, 1974. 4. W.W. Schulz and L.D. Mclssac, in Transplutonium 1975 (U.Muller and R. Lindner, eds.) North Holland, Amsterdam, 1976, p.433. 5. U.W. Schulz and L.D. Mclssac,in Proc. 1SEC 77, Canadian inst. Min. Met., Toronto, Canada, 1978. 6. L.D. Mclssac, J.D. Baker, and J.W. Tkachyk, U.S. ERDA Report ICP-1080, Allied Chemical Co., Idaho falls, Idaho. 1975. 7. T.H. Siddal Jr., U.S. Patent 3, 243, 254 (19661. 8. U.U. Schulz and J.D. Navratil, "Recent developments in Sep. Sci.1, Vol.VII, Li N.N. Ed. CRC press, Boca Ratan, Fla., 1982 Chap.2. 9. t.P. Horwitz, D.G. Kaiina and A.C. Muscatello, Sep. Sci and Tech. 16_, pp403-416(1981l. 10. E.P.Horwitz,A.C.Muscatello, D.G. Kaiina and L. Kaplan, Sep. Sci. Tech. IJL.pp 417-426(1981). 11. A.V. Jadhav, V.K. Goyal, S.N. Pattanaik, P.S. Shankaran and S.K. Patil, J. Radioanal. Nuc). Chem., Articles, 82,229-245(19841. 1.0 1 : [U] =1.771 xlO M 2 : [U] • 1.592 *10"2M - 1.0 350 425 500 WAVELENGTH, nm Fig 9. ABSORPTION SPECTRA OP U Table - 37 Variation of Distribution Ratio of U(VI) with Aqueous Phase Composition Organic Phase: 10% DBDECMP-XyIene A B C Aqueous phase D Aqueous phase D Aqueous phase D 0.5M HN03 2.6 2M HNO3 + 0.25M H2S04 8.1 0.5M H2S04 + 0.5M HNO3 0.8 1.0M HNO3 5.0 2M HNO3 + 0.5M H2S04 6.7 0.5M H2S04 + 1.0M HNO3 2.2 2.0M HNO3 10.8 2M HNO3 + 0.75M H2SO4 5.4 0.5M H2S04 + 2.0M HNO3 5.9 3.0M HNO3 15.9 2M KMO3 + 1.0M H2S04 4.7 0.5M H2S04 + 3.0M HN03 9.4 4.0M HNO3 21.7 2M HNO3 + 1.5M H2S04 3.8 0.5M H2SD4 + 4.0M HNO3 13 5.OH HNO3 27.2 2M HNO3 + 2.0M H2S04 2.7 0.5M H2S04 * 5.0M HNO3 17 Table - 38 Variation of Distribution Ratio of U(VI) with Aqueous Phase Composition Organic Phase : 10X DBDECHP - Xylene. A B Aqueous phase D Aqueous phase D 0.5M H2S04 +3M HNO3 +0.25M H3P04 3.8 0.5M H2SO4 UH H3P04 +1H HNO3 0.2 0.5M H2S04 +3M HNO3 +0.50(1 H3P04 2.1 0.5M l!2S04 +1M H3P04 +2M HNO3 0.6 0.5M H2S04 +3M HNO3 +0.75M H3PO4 1.3 0.5M H2S04 +1H H3PO4 +3M HNO3 0.8 0.5M H2S04 +3M HNO3 +1.0 M H3PO4 0.9 0.5M H2S04 +1M H3P04 +4M HNO3 1 0.5M H2S04 +3M HNO3 +1.5 ti H3P04 0.4 0.5M H2SO4 +1M H3P04 +5M HNO3 1.1 0.5M H2S04 +3M HNO3 +2.0 M.H3PO4 0.3 98 Table - 39 Extraction of U-233 from analytical waste generated during Davies and Gray method:Aqueous phase - adjusted to 0.5M ^SO^.IM H3PO4 and 5M HNO3 Organic phase - 30% DBDECMP - Xylene (Preequi1ibrated with 8M HN03) Total U-233 taken = 13.2 rag Unextracted U-233 = 0.27 rag i" 2%) Loss in washings = 0.14 mg I" 1%) Back-extracted U-233 = 12.7 mg (~ 96X) (with (NH4>2C03 solution) 99 3. CHEMICAL QUALITY CONTROL OF NUCLEAR FUELS. 3.1 ANALYTICAL METHODS 3.1.1 DISSOLUTION OF Pu-Al-Zr ALLOY FOR THE POTENT IO11ETR IC DETERMINATION OF PLUTONIUM R.B. Manolkar, S.P. Hasilkar, Keshav Chander and S.G.Marathe Accurate knowledge of plutonium in Pu-Al-Zr alloy becomes necessary in its fabrication. Known methods C1.23 of .dissolution employ either 6M HC1 or HNO3 containing Hg(II) in trace amounts. Solutions obtained by these methods cannot be directly used for the potentiometric determination of plutonium using AgO oxidants^33. Therefore a method has been developed for the dissolution of Pu-Al-2r alloy that could be directly employed for the determination of plutonium. Various reagents like Cone. HNO3, Conc.H2S04, HCIO4, Conc.H2S04 -1M HNO3 etc. were tried for dissolution. HCIO4 and cone. H2SO4-IM HNO3 gave clear solutions but the latter was preferred for the dissolution of the samples as this does not involve the heating step and the use of hazardous HCIO4 is avoided. The presence of concentrated H2SO4 not only provides the heat of dilution but also prevents the formation of the oxide layer. The alloy samples (200-300 mg) were dissolved in concentrated H2SO4- 1M HNO3 as well as in conventionally employed 6M HCl also. The solutions obtained by dissolution in 6M HCl were fumed with concentrated H2SO4 in order to remove the chloride ion prior to the plutonium determination. The results are given in Table - 40 for comparison sake. It can be seen that 100 the results obtained by cone. H2SO4-IM HNO3 dissolution are in good agreement (within ± 0.4%) in comparison to those obtained by 6M HC1 dissolution method. This method of dissolution was employed for the determination of plutonivim in Pu-Al-Zr alloy fuel samples during fabrication campaigns. References: 1. O.J. Wick Ed. "Plutonium Hand-Book", 2,728(1980) 2. W.W. Schulz, BNWL-204 (1966). 3. J. L. Drummond and R.A. Grant, Talanta, JJ3 ,477(1966). Table - 40 Comparison of Pu-values in the alloy Pu-Zr-Al using the present and conventional methods of dissolution. SI. No. % Plutonium dissolution in % Deviation SM HC! Present Method 1 38.41 ± 0.02 38.56 ± 0.05 + 0. 39 2 39.39 ± 0.03 39.35 ± 0.01 -0. 10 3 38.71 ± 0.08 38.69 ± 0.01 -0.05 4 38.31 ± 0.01 38.43 ± 0.02 + 0.31 5 23.09 ± 0.03 23.09 ± 0.04 + 0.01 6 23.51 ± 0.01 23.54 ± 0.02 + 0. 13 101 3.1.2 ELECTROCHEMICAL METHODS 3.1.2.1 COULOMETRtC DETERMINATION OF URANIUM BY SUCCESSIVE ADDITIONS METHOD N. Gopinath, J.V. Kamat, H.S. Sharma, S.G. Marathe and H.C. Jain The successive addition method involves the analysis of new aliquot added successively to the same electrolyte medium containing previously analysed aliquots. Main advantages of this method are the reduction in analytical waste volume and also in overall analysis time. Efforts were, therefore, made for developing this method for uraniun determination in UO.?.' UN, UC type fuels. During the course of this study, 10-12 aliquots of a. solution were analysed for uranium and the accuracy 'and precision of better than ± 0.2% was achieved. it may be emphasised that only 20ml of analytical waste was generated including washings after analysis of 10 aliquots while by the conventional coulometric and potentiomet.i.- methods the volume of waste is normal ly 100 and 30Cir.! respectively. Results of uranium analysis by this method are presented in Table - 41. The influence of time delay on the analytical value of uranium in a fresh aliquot analysed in the same electrolyte containing uranium!(VI, accumulated from previously analysed ai'quots. was also investigated. The values were found to be positively biased and the bias increased with time delay (03% for 30 minutes,0.6% for 60 minutes and 1.6% for 90 minutes) but decreased with the increase in sulphuric acid concentration. On the basis oi 102 studies carried out, the positive bias was attributed to the oxidation of U( IV) by mercurous ions produced in the coulometric cell by dissolution of mercury in electrolyte medium. The positive bias could be eliminated when the electrolyte containing small amount of uranium Table - 41 Results of U analysis by successive addition Method in H2SO4 (U-Content per aliquot : 3 to 4 ag> No. of Total U Analysis Total waste (H2SD4] determi- accumulated Tiae voluae generated R.S.D M nations (rag) (ain) lal) X 0.5 6 20.18 72 6 0.1 1 6 22.05 75 6 0.1 2 13 35.89 150 9 0.1 3 8 24.03 100 7 0.1 103 Table - 42 Uranium analys is by successive additions method after the elapsed time interval fol1 owed by re-electrolysis. [H2S04] U, Expected u, Determined mg/g mg/g 30 rain 60 min 90 min 0.5 16.70 16.,70 16,,75 - 1.0 16.70 16.72 16.69 16.72 2.0 16.70 16.71 16.72 3.0 35.72 35.68 35.69 35 .71 104 3.1.2.2 CONTROLLED POTENTIAL COULOMETRIC DETERMINATION OF URANIUM IN PRESENCE OF IRON OR PLUTONIUM USING PLATINUM WORKING ELECTRODE U.M.Kasar,A.R.Joshi,C.K.Sivaramakrishnan and S.K.Patil Coulometric determination of uranium is conventionally carried out using mercury electrode. Use of mercury as working electrode requires critical control over purity of mercury, stirring conditions and purity of inert atmosphere during the determination of uranium. Use of solid working electrodes minimises some of these difficulties. Moreover by using solid electrode in gauze form, larger surface area of the working electrode can be easily achieved and unlike mercury electrode it can be maintained constant during the electrolysis without much difficulty. Philips and Grossley'-l^ have reported a controlled potential coulometric method for the determination of uranium using solid working electrode. In their method uranium is chemically reduced to U(IV) by eIectrolytical1y generated hydrogen and U(IV) thus produced is determined by electrolytic oxidation at platinum working electrode using Fe(III) as an intermediate. Fardon and McGowan^] have used Ti(lll) for chemical reduction of U(VI ) and U(IV) thus produced is oxidised at gold working electrode. Interferences from iron or plutonium in the above methods were overcome by the determination of both U+Fe or U+Pu by simultaneous oxidation in the first electrolysis step followed by the determination of only iron or plutonium in subsequent reduction step. The amount of uranium in such samples is computed from the difference in the charge collected in the two electrolysis steps. The major disadvantage in these methods is that they give rise to large errors in uranium values particularly when iron or plutonium is present in large quantities. An alternate method was,therefore, developed for the determination of uranium in 105 presence of large quantities of iron or plutonium in which consideration was given to selective oxidation of Fe(II) or Pu(lII) prior to determination of U(IV) to reduce time and errors in the analysis. Analytical procedure. The method consists of reduction of uranium and iron or plutonium in the aliquot to U(fV) and Fe(II) or Pu(III) in 8-9 fi H2SO4 by Tit 111). The destruction of excess Tit IN) and the selective oxidation of Fe(Il) or Put II I) was carried out by adding a few drops of 7-8 M HNO3 followed by addition of a few drops of 1.5 M sulfamic acid within 10-15 seconds after nitric acid addition. Uranium (IV) was then oxidised e1ectro1yticaI 1y at 0.75 V vs SCE after diluting the solution to 2-3 M H2S04 and adding 4-5 mg of Fe( I I I I. The charge collected in this electrolysis step was used to calculate the amount of uranium in the aliquot. Results. The method was employed for the determination of uranium in pure uranyl sulphate solution and the results obtained are summarised in Table - 43. During the application of this method for the determination of uranium in presence of iron or plutonium, it was observed that after destruction of excess Ti(III) and selective oxidation of Fe(ll) or Pu(lII) by 7-8 M HNO3, sulfamic acid must be added within 10-15 seconds. The time delay in the addition of sulfamic acid can result in catalytic oxidation of U(1V) by iron or plutonium giving lower values for uranium. The results obtained on the determination of uranium in the synthetic mixtures of U+Fe and U+Pu at varying ratios of U/Fe and U/Pu a:e summarised in Tables 44 and 45 respectively. It was observed from the results that the method gives reproducibi1ity of < ± 0.25% and the uranium values are in excel lent agreement with corresponding values 106 obtained using Davies and Gray Reference. 1. G. Phillips and D. Grossley, Proc. of Int. Symp. on Nuclear materials safeguards, IAEA, Vienna, 1978, IAEA -SM-231/56. 2. J.B. Fardon and I.R. McGowan, Talanta,1£, 132K1972). 3. W. Davies and V. Gray, Talanta,n,120311964). Table - 43 Determination of uranium in pure urany1 sulphate solutions. 5.No. Amount of uranium Concentration of uranium 1. 4.007 32.88 2. 4.783 32.98 3. 7.982 32.90 4. 4.339 33.03 5. 4.156 32.90 6. 4.903 32.94 7. 5.052 32.90 8. 4.884 32.93 9. 4.008 33.07 10. 4.990 32.85 11. 3.781 33.09 12. 4.197 32.99 13. 5.049 33.08 14. 3.245 32.99 15. 3.894 33.02 16. 3.352 32.98 17. 3.500 32.97 18. 3.464 33.05 19. 3.208 32.97 20. . 3.552 33.03 Mean = 32.98 mg/g R.S.D. = ± 0.2% Concentration determined by Davies and Gray method = 32.95 mg/g 107 Table - 44 Determination of uranium in presence of varying amounts of Iron. S.No. U/Fe ratio Concentration of R.S.D. uranium mg/g '% 1. 10 : 1 32.94 (11) ± 0.18 2. 6:1 32.96 (10) ± 0.23 3. 4:1 32.97 (11) ± 0.20 4. 2:1 32.93 (12) ± 0.21 5. 1:1 32.93 (10) ± 0.18 Concentration determined by Davles and Gray method 32.95 mg/g Figures in parentheses indicate the number of determinations. Table - 45 Determination of uranium in presence of varying amounts of plutonium S.No. U/Pu ratio Concentration of R.S.D. * Expected cone. uranium (mg/g) % (mg/g) 1. Pure uranium 31. 82 (4) ± 0.06 31 .87 2. 25 : 1 176.21 IE) + 0.08 176.25 3. 15 : 1 156.22 (5) + 0. 13 156.50 4. 5 : 1 59.443 (6) ± 0. 11 59.484 5. 1 : 1 109.36 (6) ± 0.24 109.41 6. 1 : 5 109. 16 (8) ± 0.22 109.41 Concentration determined by Davies and Gray method Figures in parentheses indicate the number of determinations 108 3.1.2.3 A NEW APPROACH FOR THE DETERMINATION OF URANIUM BY CONTROLLED POTENTIAL COULOMETRY BY MAKING USE OF COULOMBS VS TIME RELATIONSHIP P.K. Kalsi, L.R. Sawant and S. Vaidyanathan. In the earlier work, measurement of coulombs at particular current intervals was made use of to calculate the total coulombs due to the reduction of U < V I ) to U(IV).The measurement of current manually has some uncertainty whereas time and coulombs can be automatically recorded and stored in a multimeter. Hence a relationship of coulombs Vs time has been derived simplifying Lingane's equation. The equation derived is given by QR = (Qoo - Qt ) = Remaining coulombs at time t^ QR^ = (Qoo - Qto' = Remaining coulombs at time t2 Qw = total coulombs ; Q^. = Coulomb at time tj and Q^_ = Coulomb at time t2 This is a transcendental equation. It is solved by computational iterative method. This method is applicable to all reversible and irreversible ions provided they obey the Lingane's equation. In addition to the advantages like reduction in electrolysis time and background correction, the present technique needs less operator's attention, since both coulombs and time are automatically stored in the multimeter at intervals of one minute each.Subsequent 1y, the values of Q^ 1 at time t± and Qt at time t2 are recalled for calculating the 2 total coulombs Q. The mean value for the determination of 20 109 aliquots from the uranium stock solution was 11.71 mg/g with an RSD of 0.14% (expected value being 11.72 mg/g). 3.1.2.4 DETERMINATION OF URANIUM BY TI(III) REDUCTION AND BIAMPEROMETRIC TITRATION P. R. Nair, K. V. Lohitakshan, Mary X"av ier, S. G. Marathe, and H.C. Jain A novel method for the determination of uranium in presence of iron by Til III) reduction was reported earl ier'-^, During the period of this report, further studies on the fo!lowing aspects were carried out: Determination of uranium in presence of Plutonium Studies on uranium determination in presence of plutonium have shown that the method works satisfactorily. Pu(tll) formed during the reduction by Til III), gets selectively oxidised by HNO2 generated during Tit 111) destruction, in the same way as Fe(11), and does not cause any interference. The amount of uranium per aliquot was varied from 1 mg to 7 mg and plutonium from 1 mg to 3 mg. The proportion of plutonium in the samples varied from 20-70%. Satisfactory results are obtained when the amount of plutonium per aliquot is kept below 4mg. There is a tendency for negative bias when the plutonium amount exceeds 4mg. This may be due to the increasing prominence of the reaction between U(1V) and Pu(lV) giving rise to Pu(lll) before the destruction of the excess HNO2. This results in oxidation of Pu(III) by HNO2 instead of dichromate. As long as the amount of plutonium is less than 4mg there is no detectable bias. The precision of 37 determinations was jetter than 0.2 percent as can be seen 110 from Table - 46. The method thus can be adopted for mixed oxide, mixed carbide or mixed nitride samples. Table - 46 Determination of U in presence of Pu by Till 11) reduction S. No. of Pu-range U-range U cone RSD Devi. No. detmns. (X) (rag> 1 6 20-30 1-3 2.4-7. 2 13.826 0.13 +0.01 2 6 31-40 1-3 2.4-6. 4 13.807 0.. 12 -0.13 3 8 41-50 1-2 1-3 13.833 0.13 +0.06 4 8 51-60 1-7 0.8-4. 7 13.816 0, 13 -0.07 5 8 61-70 2-3.1 1.2-1. 5 13.814 0.17 -0.08 6 1 71-80 2.9 0.85 13.818 - -0.05 (77) Expected U Cone.=13.825 mg/g;0veralI mean value for 37 detmns.=13.819 mg/g overall RSD = 0.15% Deviation = -0.04% Ill Determination of uranium at microgram levels. The accuracy and precision of the method for microgram amounts of uranium was evaluated using uranium standard solution. At BQOng, 3OOj4g, lOOjjg and 50>ig levels, the respective precisions were 0.25%, 0.43X, 0.8% and 0.9%. A few experiments were also carried out in the presence of plutonium. The percentage of plutonium in these aliquots varied from 73%-94% and the uranium amounts varied in the range of 100pg-500>jg. The overall precision was 0.4% from 11 determinations as shown in Table -47. As the method works satisfactorily for ng amounts of uranium in presence of plutonium, it is worth testing its feasibility for determination of uranium in dissolver so Iut ions. Effect of foreign ions. The effects of chloride, sulphate, fluoride, oxalic acid, mellitic acid, Pd, Ru and Al were studied. Large amounts of chloride and sulphate do not have any effect. Also the presence of other ions 300>ig of F, 20mg of Al, 200ug of Pd, 50>jg of Ru did not affect the results. Mellitic acid did not have any effect even when present at lmg level while oxalic acid gave a positive bias. 112 Table - 47 Determination of U in microgram levels in presence of Pu by Ti(IIl) reduction method S. No. PIutonium U rani urn rag <(X) rag mg/g 1 1.90 73 0.709 1.281 2 1.63 73 0.595 1.287 3 1.48 73 0.538 1.281 4 1.55 74 0.533 1.284 5 1.45 73 0.523 1.276 6 1.59 75 0.524 1.283 7 1.51 84 0.276 1.292 8 1.58 85 0.276 1.291 9 1.52 85 0.262 1.279 10 1.57 93.5 0. 110 1.280 11 1.55 93.6 0. 106 1.287 Mean Value = 1.284 mg/g; RSD = 0.39% Expected value = 1.282 mg/g Reference. 1. FCD Annual Report,1987. 3. 1.3 TfTRIMETRIC METHODS 3.1.3.1 EFFECT OF AgO AND EDTA ON Fe( I I )/K2Cr207 T1TRAT10N FOR THE DETERMINATION OF PLUTONIUM S.P. Hasilkar, Keshav Chander and S.G. Marathe Requisite concentrations of thorium and plutonium are among the important specifications for thorium-plutonium oxide which is a potential fuel for the future generation reactors and an accurate knowledge of these becomes essential. The dissolution of ThO2~PuO2 can be carried out in HNO3-HF mixture of appropriate compositon. The determination of thorium in presence of plutonium can be done by complexometric titration 113 with standard ethy1enediamine tetraacetic acid (EDTA) solution and that of plutonium by AgO-oxidation method. These determinations have to be carried out in separate aliquots. Experiments in our laboratory, in connection with sequential determination of plutonium by AgO-oxidation method following the determination of thorium, gave irreproducib1e and higher values of plutonium. With a view to understand the role of various reagents, which can possibly affect the determination of plutonium, systematic investigations were carried out and the details are given here. Effect of AgQ The results of the studies on the effect of varying amounts of AgO on Fe(1 I)/K2Cr207 titre value (amount of K2Cr2O7 solution equivalent to Ig Fe Further experiments were carried out using various amounts of sulphamic acid for the same amount of AgO (200mg). It can be seen from Table 49 that there is a steady increase of negative bias in the titre value when more quantity of sulphamic acid is added. In fact/ reverse would be expected as Ag(Il) should 114 be destroyed more effectively with increasing amount of sulphamic acid. These observations suggest the formation of Ag! I I) -sulphamic acid complex. Incidental ly it was observed that by allowing some time interval between sulphamic acid addition and start of the titration, the bias decreased and practically vanished after 90 minutes (Table 50). This shows that complex formed is not stable in the medium of titration. Effect of EDTA Effect of varying amounts of EDTA (0.05M) was studied on Fe( I I )/K2Cr20"7 titre value. A positive bias was observed in the titre value and was found to increase with increase in EDTA amount added. It exceeded even 4% for 15mg of EDTA. Effect of EDTA on Fe! I I )/KzCrz0y titre value in presence of AgO (200mg) was also studied. Known amount of EDTA (equivalent to that required for the determination of 5-10mg Th) was taken in 1M H2SO4 AgO was added and excess of it was destroyed by sulphamic acid. Known amount of Fe(Il) was added and it was titrated against standard K^C^O^. The titre value was lower by 1-1.5% than expected which is similar to the one obtained when only 200mg AgO was involved. Fuming with HCIO4 eliminated the bias and the results for plutonium are reproducible, but this may not be a practical approach as large volume of solution is needed for the treatment with HCIO4. The results of these studies show that many complexities are involved when EDTA, AgO, sulfamic acid etc. are present in the titration system. The presence of oppositely acting parameters in varying amounts coupled with the unstable nature of complex species leads to inconsistent values in the FeI 1 I )-K2CT207 redox reaction making a precise sequential determination of 115 Table - 48 Effect of AgO on the Fet titre value (Medium 1M H2S04 0.4M HN03) S.No. AgO added Dichromate Titre value Dev iation 1 0 1.510 1.510 - 2 25 1.510 1.509 - 0.07 3 75 1.510 1.502 - 0.53 4 150 1.510 1.497 - 0.86 5 200 1.510 1.490 - 1.32 Table - 49 Effect of varying amounts of sulphuric acid on the Fe( I 1 ) /K-?Cr?07 titre valu-. (Medium : 1M H2S04 + 0.4M HNO3 + 200rag T S.No. Su1phamic Dichromate titre value Dev iat i on Acidt1.5M) expected obtained (%) (ml ) 1 3.0 1.784 1.764 -1. 12 2 6.0 1.784 1.7b3 -1.74 3 10.0 1.784 1.747 -2.03 Table - 50 Effect of time of analysis on the Fe(1 I)/I^Cr^Oy titre value IHedium : 1M H2SO4 • 0.4M HNO3 + 200mg AgO) S.No. T i me (min. I Dichromate Titre value Dev iat i on expected obtai ned it) 1 5 1.598 1.569 -1.81 •c 25 1.598 1.586 -0.63 3 90 1.598 1.596 -0. 13 4 180 1.590 1.598 - 116 3.1.3.2 DETERMINATION OF NITROGEN IN URANIUM NITRIDE MICROSPHERES. H.R.Mhatre, M.A.Mahajan, R . K . Ras to j> i , N.K.Chaudhuri. C.K.Sivaramakrishnan and S.K. ['at il in continuation with the study on the analysis of nitrogen in uranium nitride samples it was observed thi't r-:p;ne monon i t r i de microsphere samples obtained from fupl tlcvc i opnuMit. spot ion were susceptible to oxidation. The sulphuric acid-ccppi-r se'enate method recommended for dissolution cf r. it ride:* bv Milner et a 1 ^ * •* gave low recovor/ cf t, i * r of c;\l z ' . Additicn of ferric sulphate, mercuric sulphate, hydrogen pR r r. >: i d e etc. in 6M H2SO4 enhanced the rate of dissolution but the recovery of nitrogen was less. Phosphoric acid was f:.u:id ::. u i table fur these samples. About 50 nig of the snmpie w .-•(.ihKjr i ng wir^:-,. After disu'olut iun thy :; > 1 i!i ;.:..• w ;i ^ ( 1 ..; 1 s , • . : • •'. !. 1 • .1 ;- •. : i ;' i e ri K.eldahl apparatus, 15 ::i I of cc nccn I rct * t.'d N^l'l' WTT-, add^ii and the distilled ammenia w; dissolution but led to 4-5% low recovery of ammonia after disti1lation. References. 1. G.W.C. Milner, G. Jones, D. Crossley and G. Phillips., AERE-R 4713<1964). 2. M.A. Mahajan, H.R. Mhatre, R.K. Rastogi, N.K. Chaudhuri and S.K. Patilj Ann. Report of Fuel Chemistry Division, 1987. 3.1.4 MASS SPECTROMETRIC METHODS 3.1.4.1 SPARK SOURCE MASS SPECTROMETRY OF MOLECULAR IONS B.P. Datta, V. A. Raman, V.L. Sant, C.S. Subbanna and H.C. Jain. The ion beam produced from an r.f. spark ion source consists mainly of monoatomic ions of all the constituent elements (major, minor and trace) of the sample and, therefore, spark source mass spectrometry (SSMS) is commonly employed for elemental analysis particularly for trace analysis. The multi- atomic ions in the beam are generally low in abundance. However, the studies of molecular ions are important for following reasons : (i) for making possible an unbiased trace elemental analysis; (ii) to explore the types of molecules formed from a spark ion source and (iii) for understanding the processes which could be responsible for their formation. The studies of molecular ions, moreover, offer an opportunity to compare the behaviour of material in an r.f. spark ion source with that in the equilibrium vapours at high temperatures, thereby complementing each other. 118 A. Experimental observations, discussion and conclusions. The studies carried out are concerned mainly with the small molecular ions such as the carbides (MC ' ), oxides (MO ' ) n n and some other matrix-sensitive ions for a number of elements (M) of varying electronic configurations. Tables 51 and 52 give results of the intensity distributions for some of the carbides and oxides respectively. By and large the MC ion abundance pattern due to different metals M falls into two groups, namely, (1) monotonic decrease in the yields of MC ions as n increases, and (2) a zigzag abundance distribution of MC ions as a function of n with the maxima appearing at even values of n. For some elements such as antimony, the successive odd-even pairs, e.g. SbC and SbC , SbC_ and SbC , appear to be roughly equally stable. The oxide ions, MO , due to any given element (M) show only the decreasing trend in yields with n and the higher oxides ( n > 3) are generally rare. Other molecular ions,such as metal hydride(MH), hydroxide(MOH!, ha I ide(MX ) ,cyanide IMCN), polymeric species such as M ,M C ,M 0 , M 0 H , M X etc. were also observed. The n mn mn mnp mn studies show that, unless the nature of the molecular mass spectrum for a given matrix is predetermined, the elemental analysis may turn out to be erroneous. The details of this work programme are given elsewhere . Factors governing the bundance yield of molecules in the recorded ion beam and the probable processes of their formation have been discusssed. The transfer of material from sol id electrodes to the gaseous plasma in a spark ion source is basically a non-thermal process, but it resembles what is commonly known as high temperature mass spectrometry (HTMS) in two ways : (i) a spark ion source generates many species which are not formed in the low temperature range and thus the former serves as a device for crossing over the energy barriers for reac tions; and 119 (ii) the abundance distribution of molecules such as UC ions in SSMS parallels the abundance trend among the same series of molecules in HTMS, thereby indicating the importance of thermodynamic parameters of molecular ions in SSMS. I. The abundance correction factors for molecules: The yield distribution of M ions due to a given multi- n isotopic element (M) and given n. Let us consider a simple case such as the dimer molecules M_ of any bi-isotopic element M. If the two isotopes of the element are A and B, then two types of dimers are possible, namely, two homonuclear molecules, A_ and B_ , and a heteronuc1 ear AB molecule. In such cases, however, it is not explicit in the literature how the abundance corrections for the different M molecules of given M and n were carried out. But to know the total concentration of M molecules for given M and nJ it is essential to apply the abundance correction. For an element M having J number of isotopes, the abundance distribution of M for given n, among its different probable component molecules is expected to be governed by the express i on J Jx . 1) i i where is the fractional isotopic abundance for the ith i sotope. The mass of a particular M molecule will be given J by E jm. (j In order to verify the statistical model (Eq.l)^ the relative 120 yields of different homonuclear and heteronuc1 ear M ions of + n given n and particularly those of MM ions, of a given multi- isotopic element M have been estimated for eleven elements using different matrices. Typical results are given in Table - 53. The eleven elements according to our observation [ 2 ] which have been presented in detail , could be split into two groups ,name 1y, (a) C, Cl, K, Sr,Ag (in its own matrix), Sb, Te and Ba; and (b) Ni, Zn, Mo and Ag (in si1ver-graphite matrix). The experimental results for group (a) elements are in harmony with the statistical prediction, while experimentally unaccountable variations from the statistical model are registered for the group (b) elements. The observations could be regarded as an isotopic effect on molecule formation; however there is an urgent need for a proper explanation as the observations are not mass-dependent. Further investigations en this line are in progress. C.Compound Molecules. The abundances of compound molecules such as A B C , will bo n P q governed by the following statistical formula : J n K P q x i ; x ( i ) = ; Ex . i i where x. y. and z. are the fractional ir;o topic abundances of the ith isotope of elements, A, B ami C, having s lots I of J, K and L numbers of isotopes respectively. In fact, in SSMS where there is no stringent rule for identifying the molecular peaks, the consideration of Eq. 2 in this regard is very important. Our experiences for identifying the compound molecular peaks have been discussed elsewhere 121 References. 1. RF-Spark source mass spectrometric studies of molecular ions. B.P.Datta, V.I. Sant, V.A.Raman, C.S.Subbanna and H.C. Jain. Int. J. Mass Spectrora. Ion Proc, 91., 241 (1989). 2. Yield distribution of M Ions due to a given multi-isotopic element (Ml and given n in the beam produced from an RF-Spark ion source. B.P. Datta and H.C. Jain. Int. J.Mass Spectrom. Ion Proc., 91,241(1989). Table - 51 Concentrations and charge distributions of MC molecules due to different elements Mj yield ratio of MC to MC .[ ratio of the ntegrated concentration of MC of a given element M in the ion beam to the carbon to metal ratio in the sample, and concentration of C molecules- Matrix Molecule a MC / EMC / Molecule PPM PPM n n n MC MC , (C/M) (C ) n n-1 n PuO,-C PuC 193 95.6 8585 PuC2 2605 99.8 13.5 c, 10755 PuC3 23 0.009 c. 476 PuC, 166 7.22 728 10 0.06 c, PuC3 c» 103 18 1.8 PuCt c, 172 1.6 0.09 41 PuC7 c. 4.6 2.9 133.7 37 PuC8 c. ThC-r: ThC 406 91 ca 1227 ThC, 2772 98.4 6..83 c, 891 ThC, 20 0..0072 c. 16.4 ThC, 74 3,.7 c, 11.4 ThC, -1 c. 0.5 ThC. -1 148.7 c, 0.64 Ag-C AgC 15.3 C» 1894 AgC3 0.85 0.056 c, 1552 AgC, 0.34 0.4 1.83 c« 46 2nO-C ZnC 13 cs 1078 ZnC, 6.5 0.5 2.9 c, 1187 NiO-C NiC 128 c, 2339 NiC, 38 0.30 c, 3171 NiC, 2 0.05 c, 81 NiC, 0.3 0.15 24 c, :oo a: Concentrations (parts per million) of MCn(3rd column) and Cn (8th column) with reference to M and C, respectively, in the recorded beam. 122 Table - 52 Concentrations and charge distributions of M0n to Matrix Molecule PPMa (non ) U30B-Ag> UO 1.98x10' 99. 89 (3:1 bywt.) UO, 1.6 xlO4 99. 97 0.08 U03 19 0.0012 U30.-C UO 1514 95 UO, 176 0.12 (UPu)oxide-C UO 2331 97 UO, 125 0.054 PuO 951 98. 5 PuO, 23 0.024 ThO2-C ThO 936 98 ThO, 8 0.009 a: Concentrations (parts per million) of MOn in the recorded beam, b: Matrix sparked using a power amplifier anode voltage of 4 kV. 123 Table - 53 Experimental and predicted relative abundances of Mn molecules owing to a given multi-isotopic eleraent(M). Matrix Element Molecule Nominal Relative yield z/y Dev.a ratio i%) mass Experimental Predicted Iz) from Eq.i Graphite 36 1.0 1.0 1.0 37 0.0305 0. 0334 0.913 8. 7 38 4xlO"4 3. 71xlO"4 1.08 7.6 C, 60 1.0 1.0 1.0 61 0.049 0.0556 0.881 11. 9 62 0.0014 0.00124 1. 138 13. 8 72 1.0 1.0 1.0 73 0.063 0.0667 0.946 5. 4 74 0.002 0. 00186 1. 116 11. 6 C, 108 1.0 1.0 1.0 109 0. 11 0. 1001 1.09 9.1 0.0043 0. 00445 0.963 3.7 1*8 1.0 1.0 121 0. 13 0. 112 1. 149 15 122 0.0062 0.00557 1. 106 10. 6 C,, 168 1.0 1.0 1.0 169 0. 173 0. 156 1. 108 10. e 170 0.0114 0. 01126 1.016 i.6 c,. 192 1.0 1.0 1.0 193 0.204 0. 178 1. 144 14. 4 194 0.013 0.0148 0.874 12. 42 c,, 228 1.0 1.0 1.0 229 0. 181 0.211 0.858 14.5 KNOs-C 78 1.0 1.0 1.0 80 0. 137 0. 144 0.949 5.1 82 0.0039 0.00521 0.743 25.7 K,C2 141 1.0 1.0 1.0 0.250 0.2166 1.191 19. 1 SrC03-C Sr Sr, 174 0 1.0 1.0 175 612 0.689 1. 126 12.6 176 174 4.066 1.027 2.6 Sb-C Sb Sb2 242 0 1.0 1.0 244 391 1.490 0.933 6.7 246 0.561 0.555 1.010 1.0 a: Dev. is the percentage deviation of the experimental yield(z) from the predicted yield(y). 124 3.1.4.2 ANALYSIS OF SLUDGE SAMPLES FROM STORAGE BAY OF THE CiRUS REACTOR K.L. Ramakumar, V.A. Raman, V.L. Sant, V.D.Kavimandan and H.C.Jain. Six sludge samples received from the storage pool of the CIRUS reactor have been analysed for the trace constituents using SSMS employing photoplate detection system. The powdered samples were directly mixed with high purity graphite in 1:1 ratio by weight and homogenised. Table - 54 gives the typical results obtained for one of the samples. It is seen that the major constituents appear to be iron, Na, Mo, Ba and Pb with iron being present at a level of about 6000 ppmw. P,Cr,Mn,Co, Cu, Rb were also present but only at minor amounts. Table - 54 Concentration of elements in sludge samples from CIRUS Element Concentration pg/g of U Na 4696 P 66 Cr 82 Mn 88 Fe 6157 Cu 62 Rb 30 Mo 216 Ba 92 Pb 213 Pu 689 125 3.1.4.3 DETERMINATION OF ZIRCONIUM IN U-Zr-Al AND Pu-Zr-Al SAMPLES BY THERMAL ION[SAT ION MASS SPECTROMETRY(TI MS) K.L. Ramakumar, V.A. Raman, V.L. Sant, M.K. Saxena and H.C. Jain. Precise and accurate determination of Zr in alloy fuel materials like U-2r-Al and Pu-Zr-Al is essential from the point of view of its homogeneous distribution. Conventional methods like spectrophotometry give results for Zr with accompanied larger errors (5% or more). Other techniques like emission spectroscopy and X-ray fluorescence also give results with large uncertainties. An attempt has therefore been made to standardise an isotope dilution method for the determination of 2r employing thermal ionisation mass spectrometry (TIMS). Various parameters like sample dissolution, spike calibration, proper chemical exchange between the sample and spike isotopes, chemical separation of Zr and mass spectrometric analysis have been investigated for achieving the optimum conditions- Samples were dissolved in 6M HC1 + 0.1M HF. An enriched 91Zr spike (atom percent ^Zr = 95%) solution has been prepared by dissolving enriched 91Zr02 in HC1 + HF. The spike was calibrated against a standard solution of Zr obtained by dissolving exactly known amount of ZrO2 in HC1+HF. Concentration of the ^1 £r spike could be calculated with a precision better than 0.2%. A "Zr radioactive tracer has been employed to standardise the conditions for the anion exchange separation of Zr from the sample in HC1 medium. The sample solution was loaded on the column in 10 M HC1. After washing down the Al with 10M HC1, Zr was eluted in 2.5 ml of 3M HCI. The elution characteristics 126 were followed by counting the ^2r activity on Nai(TI) detec tor. One of the critical parameters in the isotope dilution mass spectrometry of Zr has been found to be the incomplete chemical exchange between the zirconium in the sample and the spike. Preliminary experiments carried out using Pu-Zr-Al sample solutions resulted in large errors in the Zr concentration. With a view to optimising the conditions for the complete chemical exchange between the sample and spike, further experiments were carried out with U-Al-2r samples. It was found that proper chemical exchange could be ensured in the presence of nascent hydrogen generated by the addition of a small A) foil into the solution. Zirconium was then separated employing conventional anion exchange separation procedures in HC1 medium. The precision obtained in the mass spectrometric analysis of 2r was about 1 percent. Investigations are going on to improve the precision and also to extend the method to Pu-2r-Al samples. 3.1.5 RADIOMETRIC ASSAY 3.1.5.1 DETERMINATION OF FLUORIDE BY RADIOMETRIC ASSAY OF 181Hf BACK EXTRACTED FROM HTTA TN BENZENE. M.A, Mahajan, R.K.Rastogi, N.K.Chaudhuri and S.K.Patil In continuation with the study on back extraction of from its thenoyl trifluoro acetone (HTTA) complex in benzene by fluoride in aqueous solution, a method has been developed for the determination of fluoride at low concentration level. A 0.01 M solution of HTTA in benzene was used for the extraction 127 of a suitable aliquot of 181Hf(IV) activity from aqueous 2M perchloric acid and a stock solution of *^Hf in benzene solution of HTTA was prepared which could be used for several experiments. 5 ml of this solution was equilibrated with 5 ml of the aqueous solution containing varying concentrations of fluoride. 2 ml of the aqueous phases were then gamma counted for measurement of Hf(IV) back extracted. Variation of time of equilibration showed that the equilibrium was attained within 20 min. A shaking period of 30 mins. was maintained in all subsequent experiments. Variation of the acidity of the aqueous medium showed that 2 to 2.5 II in HCIO4 was most suitable. Hence acidity was maintained in this range in subsequent experimets. When solutions containing varying concentration of fluoride were equilibrated with equal volume of the organic stock solution, the Hf(IV) back extracted showed a linear relation with the. concentration of fluoride in the aqueous medium at least upto 10 microgram/ml. The range could be extended by taking a higher initial concentration of Hf(IV) in the organic phase. RSD in 11 replicate determinations at a concentration level of 1.5 microgram/ml of fluoride was 2.2 percent. One organic stock solution of l^Hf was used over a period of one month without any adverse effect apart from the reduction of activity level due to decay of *®*Hf. A study on the effect of various foreign ions showed that 100 fold excess of Cl-.NOjj, PO"^, 50 fold excess of SO"2, and 100 fold excess of bivalent cations like Ca2+,Hg2+,Cd2+,Ni2+ etc. did not interfere. Among the trivalent cations 50 fold excess of Cr3+,Bi3+ , and 10 fold excess of Fe^+ could be tolerated but even 2 fold excess of Al3+ interfered. Though 5 fold excess of Zr4+ could be tolerated, even 2 fold excess of Th4+ interfered. The distillate obtained by pyrohydrolysis of 128 nuclear fuel sample does not contain these interfering ions and the sodium acetate buffer used for collecting the distillate had no adverse effect. When a few unknown solutions were determined using this method,the results agreed within 5 percent with those obtained by fluoride ion selective electrode, 3.1.6 X-RAY FLUORESCENCE 3.1.6.1 X-RAY FLUORESCENCE ANALYSIS OF ZIRCONIUM K.L. Chawla, N.D. Dahale and N.C. Jayadevan. Plutonium-zirconium and uranium-zirconium alloys are important metallic nuclear fuels. The chemical methods employed for the analysis of zirconium include a gravimetric method using selenius acid or mandelic acid and igniting to ZrO2- Spectrophotometrica1 1y it is analysed us;ng Arsenazo( I I I) reagent for colour development. Fluoride ions which are used to keep zirconium in solution interfere. These are either expelled by evaporation or complexed with a 1 urniniurn( 1 I 1 ). These procedures are time consuming. A rapid X-ray fluorescence method for the analysis of zirconium in solution was standardised. Yttrium was used as an internal standard. Linear calibration plots between background corrcted '2rK '''YK ratio and zirconium concentration were obtained. Mo or W target X-ray tube was used. The precision of the method was shout 1.3%. Pure zirconium stock solution was made by dissolving zirconium oxychloride in dilute HC1. Yttrium stock solution was made v.y dissolving Y2O3 in HNO3 . Weighed amounts of these two solutions were mixed in 5 ml volumetric flasks made up with dilute HC1 to obtain a series of solutions for 129 calibration. 1 mi of the solution was transferred to a cell and counted at each of the three 20° angles consisting of background (25° for Mo tube source and 22° for W tube source),2rKw line (22.55°) and YKa line (23.80°).The background was corrected using the variable background method c * -1. The results on the calibration for zirconium solutions using Mo target tube are given in Table - 55. The intensity ratio Ip (= ' Z r ^ ' Y ' could be related to the concentraton ratio Y/Zr, where Zr and Y are the zirconium and yttrium concentrations respectively, through the linear relation: I pj = m(Zr/Y). The slope m remains almost constant for the entire concentration range of 0.3 to lmg/ml of zirconium and has a precision of 1.3%. Linear calibration piots passing through the origin in the concentration range of G.3 to lmg.'ml of zirconium using Mo or W target tubes are shov. in Fig. 10. Use of U tube increases the slope m from 0.35 for ,'1o tube to 0.80 as expected with increase in the atomic nu.nber of the target material. Further, cha rac ter i •.. ' ; r line of Mo target lies on the long wavelength side of the I'' absorption edge of zirconium. When solutions of Pu-Zr-Al alloys containing about 2% rirconium solutions were analysed no peak for zirconium was observed. Plutonium is found to absorb 2rK& intensity beciuse comparatively smaller amount of zirconium is present associated with larger amounts of pIuton i um matrix having higher mass absorption co-efficient for riu as well as U radiation. The method described herp can be used for such solutions only after chemical separation of plutonium, since in these experiments, 1 ml solutions were used to avoid handling of large amounts of plutonium. We plan to use bigger 4.0 A Mo TUBE u O ° W TUBE u 3.0 CO CO I 2.0 3.0 4.0 5.0 Zr CONCENTRATION, mg Fig 10.CALIBRATQN PLOTS FOR ZIRCONIUM. 130 cells containing larger amounts of solution to obtain calibration plots with different amounts of plutoniura, so as to analyse zirconium without chemical separation of plutonium. Reference. 1. D. Ertel, J. Radioanal. Chem., 2,205(1969). Table - 55 Relationship between Intensity Ratio and Concentration ratio Y/2r Amount of 1 a tens i ty Zr Y ratio after ( lRx(Y/Zr) mg Bkg. correction 'R 1.4456 2.8584 0.1759 0.3478 1.5275 2.5202 0.2136 0.3524 1.8704 2.4580 0.2750 0.3614 3.0129 3.6298 0.2932 0.3532 4.4782 3.0174 0.5303 0.3573 Mean 0.3544 • 1.3% 131 3.2 PRIMARY CHEMICAL ASSAY STANDARDS 3.2.1 PRIMARY CHEMICAL, ASSAY STANDARD FOR URANIUM K.D. Singh Mudher, R.R. Khandekar, K. Krishnan, N.C. Jayadevan and D.D. Sood The assay of uranium in nuclear fuel is vf.uy important for quality assurance for ensuring strict chemie/'l specifications required for optimum performance. Out of the few evaluated chemical assay standards for uranium, U^flg is the? nne commonly used and is available as NIST reference material. However, its s to ich i oms t ry depends on the starting compound. sample? size, temperature and duration of ignition. Rubidium uranium sulphate i.e. Rb2U(S04)3 has hen evaluated as a primary chemical assay standard fur uranium. Uranium dioxide obtained from Uranium Netai F I n n 1. , BARC aa? dissolved in nitric acid, precipitated as aiEwon i uni •! i u rang t n , filtered, dried and converted to UO3 at 3!H.i - -,00 ° C. It was dissolved in 1 M H2.SO4 and U ( V I ) reduced to U < I V > in ,->••, electrolytic cell usinf* platinum PIPCU od"? as r.hcwn in Fig. 11. Rb2C03 (Rs 1.2 times) was added to IJ(1V) solution which wa::. evaporated under 1R lamp to giv» green coloured crystals f Rb2U(S04)3 which were washed with absolute alcohol several times. Seven different lots of 3-!5r; of Rbv'MSO/,)^ were prepared. The samples prepared v •--•;,• • •!; - r a." • c r i ?; ed l.iy different chemical and physical nu.'t Inul:;, Doth uranium a nd sulphate w e i •' 1 • h •-» ..1 i <.';, 1 i / <• n.. 1 y ;. ( • * . Uranium was analysed by the Davies and Gray method. Sulphate was analysed g r a v i mo t r i ca I 1 y sr, HciSD,, a i I < • 1 :; i.'pa 1 ;> t i 111; uranium by precipitating it with ammonia. The results of chemical analysis of seven different lots are g i ven in Tab1? - 5 C. It is seen ANODE CATHODE(-) Pt WIRE Pt WIRE MESH URANYL 1M H2SO4 SULPHATE SOLUTION FRlTJ FIG.-11 ELECTROLYTIC CELL USED FOR REDUCTION OF URANIUM(YI) TO URANIUM(E?). from the results that the deviations of the experimental values from the expected values for both uranium and sulphate are within the precision of the methods employed. The purity of the material was determined by taking into account the concentrations of trace metal lies obtained by AES method. Thus total impurity content was found to be less than 0.03%, thereby giving the guarantee that the material can be prepared in pure form. The 1R spectrum of Rb2U(S04>3 (Fig. 12) shows that there is no water of hyjration in the compound since the bands due to 0-H stretch around 3500cm"1 and H-O-H band around 1600 em'^-are absent. The appearance of (nue-3) V3 absorption frequencies (S-0 stretching modes) at 1000- 1250cm~ *• which are higher than those present in U(SO4)2-4H2O suggests that 3 contains chelating sulphate groups. The cell parameters of Rb2U! 30.4)3 as derived from single crystal X-ray Weissenberg photographs and refined by least squares methods are given in Table -b7. The compound belongs to the monoclinic system. Thermal analysis as shown in Fig.13 confirms the absence of water of hydration. On heating in air there is no weight loss upto 550°C, beyond which it decomposes at 6 50°C to form Rb2U02(SO4 ) 2• On further heating, the compound decomposes above 800°C with 3 continuous loss to form K'"12'-)2(-'7 • It is clear therefore that RI.^LMSO/.,^ is a stable, anhydrous compound which can be easily prepared in a pure form. It is also found to be easily soluble in common acids. Thus the material meets most of the requirements of a primary chemical assay standard. UJ o •z. < CO z: < Q: 1- 4000 3000 2000 1600 1200 800 400 WAVE NUMBER (cm"1) FIG.-12. IR SPECTRUM OF Rfc>2 U { S04> 3 . WT. OF SAMPLE = 517.5mg \ TEMPERATURE ( C) Fig 13. TG AND DTA CURVES OF Rb2U(S04)3. 133 Table - 56 Assay of uranium and sul phate ir1 Rb;2U(SO4J3 Lot Deviation from SO4" Deviation from+ No. expected vat ue * expected value 1. 34 . 14 - 0,.03 41. 11 - 0.,51 2. 34 .10 - 0., 15 41.26 - 0.,15 3. 34 .07 - 0.,23 41.22 - 0.,24 4. 34 . 13 - 0.,06 41.27 - 0.,12 5. 34 .20 + 0.,15 6. 34 .07 - 0.23 7. 34 . 13 - 0.06 Mean 34 .12 - 0.09 41.22 - 0.24 R.S.D 0.13 0. 18 * expected value 34.15% + expected value 41.32% Table - 57 Cell Parameters of Rb2U(S04 >3 0 a (A) 9.454 (5) 0 b (A) 16.487 (8) 0 c (A) 22. 959 (9) » 6 ( ) 1141,8 (1) z 12 Density-* measured, g.cm_^ 4.17 Density-* calculated, g. cm *" 4.27 134 3.2.2. PRIMARY CHEMICAL ASSAY STANDARD FOR PLUTONIUM K.D. Singh Mudher, R.R. Khandekar, K.Krishnan, N.C. Jayadevan and D.D. Sood Out of several materials proposed and evaluated as primary chemical assay standards for plutonium, Pu(SO4)2-4H2O has been found to meet most of the requirements of primary standard and has been prepared as an NIST standard. However, being a nuclear material, it is not easily available. Double sulphates of plutonium with alkali metals are known for quite some time. Some of them are stoichiometric and can be easily prepared. Anhydrous potassium plutonium sulphate K4Pu(S04>4 has been evaluted as primary chemical assay standard for plutonium. Plutonium was purified by loading on an anion exchange column in 7M HNO3 . The plutonium in nitric acid was evaporated to dryness and crystallised as Pu(SO4)2-AH2O by adding 4M H2SO4. Crystals of Pu(SO4)2-4H2O were washed with alcohol and dissolved in 1M H2SO4 and slight excess of K2SO4 solution than required for stoichiometry was added. Green coloured crystals of K4PU(SO4)4.2H2O were obtained on heating under an IR lamp. The crystals were washed with absolute alcohol and dried in air. These crystals were heated to 300°C for 3-4 hours to obtain red coloured anhydrous !<4Pu(S04)4. Samples were prepared in five lots at 2-5g level and stored in glass-stoppered bottles. The anhydrous potassium plutonium sulphate was characterised by chemical analysis, X-ray diffraction, infra- red and thermogravimetric methods. Analysis of plutonium was carried out by titrimetric methods whereas sulphate was analysed gravimetrica11y as BaS04 . The results of chemical analysis of five lots of anhydrous l<4Pu(S04)4 prepared separately are given in Table -58. The 135 2- standard deviation calculated for Pu and SO4 analyses are within the precision of the analytical methods. The total impurity content as determined by spark source mass spectrometric and emission spectrographic analysis was below 300 ppm. Anhydrous salt stored over a period of six months in a desiccator has shown no weight change indicating that the samples are stable to alpha radiolytic effects and atmospheric cond i t i ons. The inf rB.fed spectra of samples showed that the number of absorption bands observed were more than that required for even the lowest symmetry of the molecules, suggesting the presence of two or more molecular groups in the unit cell. Thermogravimetric patterns indicated that the compound is •stable upto 700°C, beyond which it decomposes to give K2SO4 and Thus anhydrous K4Pu(S04>4 has been found to meet most of the requirements of a primary chemical assay standard for plutonium. It has a defined stoichiometry and is easy to prepare. It is easy to purify and dissolves easily in acid solution. It is stable to atmospheric and alpha radiolytic effects. It is stable as anhydrous salt upto 700°C. The results of preparation at 2-5 gm level of Pu in five different lots have shown quite satisfactory results. Studies are planned with larger amounts of plutonium. 136 Table - 58 2- Assay of Pu and SO4 in K^Pu!504)4 Lots Pu Deviation from ^ SO4" Deviation frora+ expected value expected value 1 30.64 + 0. 12 49. 16 - 0.26 2 30.68 - 0.01 49.50 + 0.42 3 30.71 + 0.09 49. 13 - 0.32 4 30.71 + 0.11 49. 13 - 0.32 5 30.74 + 0.20 49. 18 - 0.08 Mean 30.69 + 0.10 49.22 - 0.10 R.S.D. 0.14 0.32 * expected value 30.66X + expected value 49.27% 3.2.3. CHARACTERISATION OF INDIGENOUSLY PREPARED CHEMICAL ASSAY STANDARD FOR PLUTONIUM K.L. Ramakumar, V.A. Raman, V.L. Sant, V.D. Kavimandan and H.C. Jain Double sulphate of potassium and piutoniurn,K4PU(SO4)4. has been identified to be the prospective candidates for the possible use as chemical assay standard. A test lot of K4Pu(S04)4 prepared in the Division has been characterised for the trace constituents by spark source mass spectrometry (SSMS) employing photoplate detection system. With a view to minimising the isobaric interferences at mass numbers of interest, the photoplate exposure was carried out at higher resolution (""40001. As both potassium and sulphur form different molecular ions between themselves and also with carbon and oxygen present in the sample, the identification of the mass numbers for the impurity element was carefully arrived at by checking the isotopic distribution as welI as the 137 presence or absence of any non-interfering molecular ion near these mass numbers. At times more than one isotope of the element was monitored to cross check the results obtained. Table - 59 gives the concentration of the trace constituents determined in K4Pu This work is being carried out as part of the Divisional efforts to prepare and characterise a chemical assay standard for piutoni um. Table - 59. Concentration of trace constituents in K4Pu(S04>4 S.No. Element ConcentratioConcentrat ionri ppmww with respect to K^jPulSO/, 1. Sodium 53. 7 2. Aluminium 14. 6 3. Phosphorus 1. 0 4. Titanium 8. 8 5. Manganese 1. 5 6. Cobalt 1. 0 7. Nickel 9. 6 8. Copper 4. 1 9. Zinc 16. 0 10. Molybdenum 10. 2 11. Tin 2. 3 12. Antimony 0. 9 13. Barium 7. 1 14. Tungsten 1. 5 15. Lead 2. 5 16. Bismuth 1. 7 17. Arsenic 0. 6 18. Silver 3. 8 19. Cesium 6. 2 20. Platinum 39. 3 21. Thorium 10. 5 22. Uranium 46. 1 Note: For elements 1 to 18, the concentration has been calculated by assuming the calibration factors (RSF) determined in U3O0 matrix. For elements 19 to 22, a unit calibration factor iRSF) has been assumed. 138 3.2.4. STUDIES ON THE STO ICHIOMETRY AND STABILITY OF RUBIDIUM URANIUM TRISULPHATE A.U. Bhanu, L.R. Sawant.P.K. Kalsi and S. Vaidyanathan Double sulphates of uranium and plutonium were prepared as possible substitutes to primary chemical assay standards. One of the compound under study is rubidium uranium sulphate Rb2U(S04>3. To confirm its stoichiometry , it is proposed to estimate uranium, rubidium and sulphate content individually. The work was initiated to standardize the suitable methods for sulphate and rubidium. Sulphate estimation. The major task was to remove uranium and then do an accurate determination of sulphate. Gravimetry procedure of precipitating sulphate as BaSO/^ was used. The results obtained were ± 0.3% accurate. The parallel runs of sulphate determination in sulphuric acid (standardised against Na2C03) and K2SO4 were carried out. Both are showing the similar trend in accuracy indicating the limitation of the accuracy level that could be attained. Rubidium estimation. To standardise the gravimetric method for rubidium estimation pure RbC1 was analysed by perchlorate precipitation methods. The results obtained were on the lower side and upto 99.4% of the values expected. Stability of Rb2U(S04)3. A study has been initiated to observe the stability of rubidium uranium sulphate both in presence and absence of light, by determinating the uranium content periodically. For this purpose, two samples of Rb2U(S04)3 were taken. One cample was 139 kept in a dark container and the other sample in a transparent glass weighing bottle. The uranium content is being determined periodically by potentiometry. The uranium contents at the time of storage were 34.24% and 34.22% respectively in the presence and absence of light. The corresponding values after two months were 34.14% and 34.14% respective1y.This work is being con t i nued. 3.3 ANALYTICAL SERVICES 3.3.1 ANALYSIS OF NUCLEAR FUEL SAMPLES FROM RADIO-METALLURGY DIVISION 760 samples of different types, like oxides, nitrides, carbides etc. were received from Radio-metallurgy Division for determination of various specifications, such as U,Pu, 0/M ratio, isotopic composition etc. Table - 60 gives a summary of the number and types of samples analysed for different spec i f i cat ions. 3.3.2 ANALYSIS OF SAMPLES FROM OTHER DIVISIONS AND RESEARCH GROUPS. Analytical services have been rendered to various units of D.A.E. and other Divisions of BARC towards their on-going research and development activities. Table - 61 gives more details on these analyses. Analytical services also included distribution of samples received from various sources to different Sections of FCD for carrying out the determinations of the required specifications. 140 Table - 60 Analyses of Samples received from R(1D. 233 Sample U Pu 0/M U XRD 0 N C Cl,F H20 Isot- Total Description opic Pu + HC1 25 31 - 1 57 solution UN 21 - - 183 190 — 394 (l/Pu)02 2 2 - 4 (UPu)N 10 10 - 10 10 - - 40 U02 8 - 29 _ 4 6 6 65 Th02-Ti02 - - 1 1 Zr-Ni - - - 4 Al-Pu alloy - 20 - 20 AI-233U-Zr - - - 33 33 Th02-U02 - - 4 h Al-Pu-Zr alI oy 67 - 67 ThO2 - - 1 1 Ti-Al alloy -• - - 8 B 16 U3O8 6 6 3 15 U02 +PuO2 +C: 2 2 - 4 UC 4 6 - 18 NaOH solution 3 - 3 ThO2 -Nb205 - - 4 4 U02 + ZrO2 1 - - 1 U02 +CaO - - 1 1 U02 +MgO - - 1 1 U02 + Nb205 - - 1 1 + U02 Ti02 - - 1 1 U02 +U205 - - 1 1 U02 +Y203 - - 1 1 Etched NaOH 1 2 - 3 Total 80 149 48 33 4 205 216 6 760 141 Table - 61 S. Sample Number of Units Analysed for No. description samples 1.. Uranium 511 Ch.T.D. 235y/238y atom rati0s by TIMS. 2.i Uranium 10 MDRS 235U/238U atom ratios by TIMS. 3., Uranium 17 Chemical 235U/Z38U atom ratios by TIMS. Group 4,, Uranium,piutonium 19 FRD Isotopic Composition and and dissolver concentration of uranium and solution samples. Plutonium by TIMS. 5.Boric anhydride 2 A.Ch.D. 10B/nB atom ratio by TIMS 6.Lithium pentaborate 2 NAPP 10B/HB atom ratio by TIMS 7.Plutonium and 3 RMD Isotopic Composition and (U.P'i)C concentration of uranuim by TIMS 8.SS samples 3 RMD B by SSMS 9.Cs2Cr2C>7 1 FCD Trace impurities by SSMS 10.Molybdenum samples 8 Univ. of Trace impurities by SSMS Madras 11.Sludge samples 6 C1RUS Trace impurities 12.Rare earth salts 36 RCD N i trogen 3 RCD X-Ra> Diffraction 25 RCD Thermal 16 Chemical X-Ray Diffraction Group 4 ChD X-Ray Diffraction 2 UCD X-Ray Diffraction 13.YAG:NG Crystal 2 TPPED X-Ray Diffraction 14.Water 2 DEED F 7 DFED N 15.Steel and Iron 29 Phys ical C Metallurgy 16.Titanium and 54 Metal 1urgy H 2 irconium 4 Metal Iurgy X-Ray Diffraction Grand Total 766 142 4. NUCLEAR MATERIALS ACCOUNTING 4.1 NUMAC DATABASE N. Maiti, P.N. Raju and S. Vaidyanathan. During 1988, the NUMAC database has been augmented to include the data on HEU from APSARA. Data from APSARA has been loaded in the database from May 1987 to uptodate. During this period, 563 accounting reports were received from 12 facilities. These were checked for internal consistency and loaded in the NUMAC database. On the last day of each month an inventory status report was prepared depicting the number of accounting reports received, consolidated status of inventory of nuclear materials, their location and form as well as nuclear materials in transit, if any. The increased use of PCs by almost all the fuel cycle Facilities in the DAE is opening new avenues for the nuclear materials accounting. It was felt useful to receive the data from the Facilities through floppy discs instead of the reporting forms. Receiving data through floppy discs has the following advantages namely, (i) the confidentiality of the data can be maintained ; (ii) time saving; and (iii) no chance of transcription errors, etc. This can also be extended to the safeguards data which are being communicated to IAEA . 143 To equip ourselves for the implementation of this , a micro computer system (QUANTUM AT) of 1 MB main memory with 20 MB hard disk, 5 1/4" floppy disk drive and advanced link software to communicate between the HP computer system and any other computer system has been selected. To get the system, necessary follow up work has been taken up. 4.2. TIME SERIES ANALYSIS M.B. Yadav and Hari Singh. A time series study of UCIL monthly production data for uranium (1975-1988) has been carried out. Monthly production data for uranium from March 1985 to December 1988 have been used for this study. Multiplicative model of Time Series Y = TCS1 has been used. Different components of time series viz., Trend, Cyclic, Seasonal and regular have been estimated. These estimates have been used in forecasting the monthly uranium production upto the year 1995. Irregular component has been used in estimating the error in forecasting. The 95 percent confidence intervals for the predicted values have been estimated. As part of this jtudy, four computer programmes namely, TIMES, SEASON, CYCLE and PRD were developed on HP-1000 system. 144 4.3. SOFTWARE DEVELOPMENTS 5. Jyothi and Hari Singh. A Iibrary program GAUSS for determining the integral of a real valued function on a given interval was developed based on a recursive alogorithm employing the 20-point and 10-point GAUSS quadrature formulas. A number of computer programmes such as QUADM, PEARS, W-TEST, KSTAT and PROBEQ developed earlier on the BESM-6 and ND-500 computer systems of BARC, have been modified so that they could be run on the HP-computer system of NUMAC. These programmes are useful in carrying out MUF analysis studies. 4.4. INSTRUMENTATION D.B.Paranjape and S. Vikram Kumar The procurement of various components of a 4096 channel analyser is underway for assembling it in collaboration with Health Physics Division. This MCA is 8085 based NIM compatible system coupled through 100 MHz ADC. A gamma ray spectrum can be acquired and analysed after proper shaping and amplification of the detected signal. The ADC and the display units have yet to be received. All the necessary units to set up a silicon surface barrier based oc-spectrometry have been tested. The pipe line that connects both the detector's vacuum chamber and the thermocouple gauge to the vacuum pump is under fabr ication. 145 A dehumidifier has been procured and this would be instal1ed in the counting room for humidity-control. The electronic circuitry of a photovoltaic pressure switch/gauge that will be connected to the glove boxes to control/monitor the inside pressure has been modified according to the user needs. Both the lower and upper set points provide good control on the pressure. It has been modified to latch the relay once the pressure inside the glove box is out of the set pressure range. 146 5. LIST OF PUBLICATIONS DURING 1908 1. Thermodynamic properties of Cr(1_x)Tex for x - 0.546 and 0.526, R. Prasad, V.S. Iyer, Z. Singh, V. Venugopal, S. Mohapatra and D.D. Sood , J.Chem. Thermodynamics, 20.,319(1988). 2. Thermodynamics of FeTeg.g, R. Prasad, 5. Mohapatra, V.S. Iyer, V. Venugopal and D.D. Sood, J.Chem. Thermodynamics, 20.45311988). 3. Standard Gibbs molar free energy of formation of Na2ZrO3, V.S. Iyer, V. Venugopal, R. Prasad, Z. Singh, S. Mohapatra and D.D. Sood , J.Chem. Thermodynamics, 20.751(1985). 4. Vaporisation thermodynamics of CS2M0O4 , R.P. Tangri, V. Venugopal, D.K. Bose and M. Sundaresan, Paper presented at the International Symposium on Thermodynamics of Nuclear Materials held at Chicago, USA, September!1988). 5. Thermal properties of Cs2Cr207 11. Simultaneous determination of the 235y/23By isotope ratio and concentration at nanogram levels of uranium employing a mixed spike in thermal ionisation mass spectrometry, S.K. Aggarwal, R.K. Duggal, P.M. Shah and H.C. Jain, Int. J. Mass Spectrom. Ion processes, 85,13711968). 12. Determination of trace constituents in tellurium and zircaloy-2 employing spark source mass spectrometry, K.L. Ramakumar, V.A. Raman, V.L. Sant, V.D. Kavimandan and H.C. Jain, J. Radioanal. Nucl. Chera. Letters,125(2). 467(1988). 13. Determinaton of uranium and plutonium in plutonium based fuels by sequential amperometric titration, P.R. Nair, K.V. Lohitakshan, Mary Xavier, S.G. Marathe and H.C. Jain, J. Radioanal. Nucl. Chem. Articles.12211).19t1988). 14. Complexometric determination of thorium in 2 S.F. Hasilkar, N. Gopinath, Keshav Chander, S.G. Marathe and H.C. Jain J. Radioanal. Nucl. Chem. Letters.122(1).69(1988). 15. Spark source mass spectrometry in nuclear technology for trace analysis, H.C. Jain and K.L. Ramakumar, Paper presented at the 11th International Mass Spectrometry Conference, France, August-September(19881. 16. Iterative computational method for rapid analysis of Fe and Pu by CPC and some interesting observations on the coulogram of Pu(III), R.C. Sharma, P.K. Kalsi, L.R. Sawant, S. Vaidyanathan and R.H. Iyer, J. Radioanal. Nucl. Chem. Letters. 126, H1988). 17. Assay of uranium in scrap and waste produced at natural uranium metal production and fuel fabrication plants, A.U. Bhanu, P.C. Kalsi, S. Sahoo and R.H. Iyer, J. Radioanal. Nucl. Chem.Articles, 121.29-4511988). 18. Some obervations on the superconductivity in Tl-Ca-Ba-Cu-0 System. Ram Prasad, N.C, Soni, R.V. Kamat, V.N. Vaidya, C.V. Tomy and S.K. Malik Paper presented at the Solid State Physics Symposium held at Bhopal Dec 20-23,1988 19. Mass Spectrometry in Nuclear Technology : Two decades of our experience, H.C. Jain, Invited talk IT-4, Fourth National Symposium on Mass spectrometry, lISc, Bangalore, January 4-6(19881 148 20. Diffused peaks in spark source mass spectrometry, K.L. Ramakumar, V.D. Kavimandan, C.S. Subbanna, V.A. Raman, V.L. Sant and H.C. Jain, Paper No. 1-12, Ibid. 21. Comparison of relative sensitiviy factors in pure U3O8 and U3O3-C matrices, K.L. Ramakumar, V.A. Raman, V.L. Sant, V.D. Kaviraandan and H.C. Jain, Paper No. NT-6, Ibid. 22. Analysis of plutonium bearing fuel materials by spark source mass spectrometry employing electrical detection system, K.L. Ramakumar, C.S. Subbanna, V.D. Kavimandan, V.A. Raman, V.L. Sant and H.C. Jain, Paper No.1-13, Ibid. 23. Relative sensitivity coefficient in spark source mass spectrometry and its relation with element sensitive physico-chemical properties, B.P. Datta, V.L. Sant, V.A. Raman, V.D. Kavimandan and H.C. Jain, Paper No. NT-7, Ibid. 24. Determintion of isotopic composition of nanogram amounts of uranium using U-233 as a spike, S.A. Chitambar, A.R. Parab, P.S. Khodade and H.C. Jain, Paper No.NT-8, Ibid. 25. Isotope fractionation factors in thermal ionisation mass spectroraetric analysis of uranium and plutonium, S.A. Chitambar, P.S. Khodade, A.R. Parab and H.C. Jain, Paper No.NT-9, Ibid. 26. Sol gel process for fuel fabrcation, V.N. Vaidya Invited talk at Radiochemistry and Radiation chemistry Symposium, BARC;Bombay, Feb. 1988 27. Macroporous resins in actinide separations, V.V. Ramakrishna, Invited talk, Ibid. ..3. Dissoluton of UC for Coulometric determination of .1), N. Gopinath, J.V. Kamat, H.S. Sharma, S.G. Harathe and H.C. Jain Paper No. CT-01, Ibid. 29. Recovery of Pu from phosphoric acid waste using mono octyl phenyl phosphoric acid, V. Shivarudrappa, P.D. Mithapura, S.G. Marathe and H.C. Jain Paper No. CT-07, Ibid. 30. Studies on the recovery of Pu from solutions containing thorium and EDTA R.B. Manolkar, Keshav Chander, S.G. Marathe and H.C. Jain, Paper No. CT-08, Ibid. 149 31. Studies on the sorption of plutonium on alumina from carbonate Media, A. Kadam, C.V. Karekar, M.M. Charyulu, C.K. Sivaranakrishnan Paper No. CT-13, Ibid. 32. Measurment of stability constants of the F" complexes of tetravalent actinides using F~ ion selective electrode -A feasibility study of Th(IV), R.M.Sawant and N.K. Chaudhuri, Paper No.CT-14, Ibid. 33. The extraction of U(V!) froraHCl acid by D2EHPA and nixtures of D2EHPA and HTTA. R.D.Bhanushali, S.Negi and V.V.Ramakrishna, Paper No.CT-17, Ibid. 34. Extraction of Put IV) by D2EHPA from sulphuric acid medium, D.G.Phal,S.Kannan,V.V.Ramakrishna, Paper No. CTrf-18, [bid. 35. Influence of nitrate ion on the extraction of PuUVJby D2EHPA fro* sulphuric acid medium. S. Kannan, D.G. Phal and V.V. Ramakrishna, Paper No. CT-19, Ibid. 36. The extraction behaviour of Pu (IV) from nitric acid and - nitric-perchloric acid mixtures into D2EHPA, D.G.Phal , S. Kannan, V.V. Ramakrishna, Paper No. CT-20, Ibid. 37. Synergic extraction of Pu((V) from nitric acid medium by D2EHPA with HTTA and TOPO, 5. Kannan, D.G.Phal, V.V. Ramakrishna, Paper No. CT-21, Ibid. 38. The solvent extraction behaviour of Pu(VI) from perchloric acid by D2EHPA, K.V. Chetty, P.M. Mapara, Rajendra Swarup, V.V. Ramakrishna, Paper No. CT-22, Ibid. 39. Some studies on the extraction of Pu from phosphate containing HNO3 solution using DBDECMP as extractant, V.B. Sagar, S.M. Pawar, M.S. Oak, C.K. Sivaramakrishnan Paper No. CT-24, Ibid. 40. X-ray studies in Rb-U-Ca-0 system characterisation of a new nixed oxide phase, K.D. Singh Kudher, R.R. Khandekar, A.K. Chadha, N.C. Jayadevan, Paper No. CT-26, Ibid. 41. Reactions of rubidium and caesium with U02-Th02 oxides, K.L. Chawla, N.L. Mishra, N.C. Jayadevan, Paper No. CT-27, Ibid. 42. Engineering scale facility for the production of UO3 gel spheres, R.V. Kamat, J.V. Dehadraya, N. Kumar, T.V. Vittal Rao, V.N. Vaidya, D.D. Sood, Paper No. CT-28, Ibid. 150 43. Synthesis of uranium raonocarbicte microspheres by sol-gel process, S.K. Mukerjee, J.V. Dehadraya, Y.R. Bamankar, V.N. Vaidya, D-D. Sood, Paper No. CT-29, Ibid. 44. Kinetics of nitrate leaching from UO3 gel particles prepared by internal gelation process , S.K. Mukerjee, V.N. Vaidya, D.D. Sood, Paper No. CT-30, Ibid. 45. Spectro-photometric studies on the behaviour of plutonium in basic media, M. Ray, l.C. Pius, M.M. Charyulu, C.K. Sivararaakrishnan, Paper No. CT-36, Ibid. 46. Studies for the electrodeposition of milligram amounts of uranium on electropolished stainless steel disks S.K. Aggarwal, P.M. Shah, R.K. Duggal, H.C. Jain, Paper No. CT-39, Ibid. a- 47. Studies on the decontamination factors using oiicroporous anion exchange resins, U.M. Kasar, l.C. Pius, V.B. Sagar, A.R. Joshi, C.K. Sivaramakrishnan, Paper No. CT-44, Ibid. 48. Absolute yields of fission products of ^^Mo and ^2je jn spontaneous fission of "^Cf, V.K. Bhargava, M.S. Oak, A. Ramaswami, Satya Prakash, Paper No. NR-07, Ibid. 49. Determination of half life of 240Pu relative to half life of 233U, S.A. Chitambar, P.S. Khodade, A.R. Parab, H.C. Jain, Paper No. NR-09,Ibid. 50. Direct potentiometric determination of uranium in organic extracts using Tit III) as reductant, A.K. Pandey, P.C. Kalsi, R.C. Sharma, S. Vaidyanathan, R.H. Iyer, Paper No. RA-01, Ibid. 51. Determination of U by Till II) reduction and biamperometric redox titration, P.R. Nair, K.V. Lohithakshan, Mary Xavier, S.G. Marathe, H.C. Jain, Paper No. RA-02, Ibid. 52. An improved method for the rapid determination of uranium by controlled potential coulometry using lingane's equation, R.C Sharma, P.K. Kalsi, L.R. Sawant, R.H. Iyer, Paper No. RA-03, Ibid. 53. Titrimetric determination of uranium in U-Pu- alloy, S.P. Hasilkar, Keshav Chander, S.G. Marathe, H.C. Jain, Paper No. kA-04, Ibid. 151 54. Assay of uraniumAMgF2 slag generated in the magnesiothermic reduction of UF4 at the natural uranium concentrates, P.C. Kalsi, A.K.Pandey, R.H. Iyer, Paper iio. RA-07, Ibid. 55. Controlled potential coulometric deternination of Pu at aicrogram levels, N.B. Khedekar, H.S. Sharma, S.G. tiarathe, H.C. Jain, Paper No. RA-09,Ibid. 56. Studies of controlled potential coulometric determination of plutoniua in mixed oxide samples, U.M. Kasar, V.B. Sagar, A.R. Joshi, V.K. Bhargava, Paper No. RA-24, Ibid. 57. Determination of sodium in cryolite, B.N. Patil, V. Shivarudrappa, S.G. Marathe, H.C. Jain, Paper No. RA-30,Ibid. 58. Alpha spectrometry for the determination cf 234u/238M rati0 in uranium samples, S.K. Aggarwal.P.M. Shah, R.K. Duggal, H.C. Jain, Paper No. RA-30, Ibid. 59. A comparative study of 239Pu, 238Pu and 233U spikes for determining Pu concentration, S.K. Aggarwal, R.K. Duggal, P.M. Shah, H.C. Jain, Paper No. RA-31, Ibid. 60. A new spectrophotometic method for the determination of nicroamounts of uranium, V.K. Bhargava, D.M. Naronha and J. Sharma, Paper No. RA-05, Ibid. 61. Determination of alpha specific activity and concentration of Pu in dissolver solution of low burnup fuels by alpha spectrometry, G.Chourasiya, (/.A. Raman, P.A. Ramasubrananian, H.C. Jain Paper No. RA-34, Ibid. 62. Chemistry of Nuclear Fuels. D.D. Sood, Invited talk at the 25th Annual Convention of Chemistry, Calcutta,1988. 63. Ion exchange in nuclear technology: some aspects. S.K. Patil. Invited talks, Ibid. 64. Synergic extraction of Pu(VI) by mixtures of HD2EHP and TOPO from perchloric acid, K.V. Chetty, P.M. Mapara, R. Swarup, V.V. Ramakrishna, Ibid. 65. Synthesis of (U,Ce)C microspheres containing 20% Ce» by sol-gel process, S.K. Mukerjee, J.V. Dehadraya, Y.R. Bamankar, V.N. Vaidya, D.D. Sood, Ibid. 152 66. Determination of uranium in presence of iron by controlled potential coulometry using platinum working electrode, U.M. Kasar, A.R. Joshi and C.K. Sivaramakrishnan, Ibid. 67. XRF analysis methods for actinide elements in nuclear fuels, N.C. Jayadevan, National workshop on X-ray emission spectrometry,lGCAR.Kalpakkarat1988). 68. X-ray diffraction study of the phases in K-Ba-U-0 system, A.K. Chadha, K.D. Singh Mudher and N.C. Jayadevan 8th Annual Conference of Indian Council of Chemists, Tirupati(1988). 69. Influence of some parameters in the coulometric determination of uranium by successive addition method, N. Gopinath, J.V. Kamat, H.S. Sharma, S.G. Marathe, H.C. Jain Presented at the National seminar on electrode, CECR1,Karaikudi, Tamilnadu, July(1988l. 70. Studies on synergistic extraction of Pu(IV) by HTTA-DBDECMP mixture from perchloric acid medium, A.V. Jadhav, K. Raghuraman, K.A. Mathew, P.S. Nair, H.C. Jain, 8th Annual Conference of Indian Council of Chemists, Tirupati(1988). 71. Effect of EDTA on ferrous-dichromate blank in the potentiometric determination of plutonium, S.P. Hasilkar, Keshav Chander and S.G. Marathe, Ibid. 72. Solvent extraction studiesof U(V1) and Pu(Vl) by D2EHPA and mixture of D2EHPA and TOPO from sulphuric acid medium, K.V. Chetty, P.M. Mapara, R. Swarup, V.V. Ramakrishna, Ibid. 73. Preparation of copper sol by alkoxide route, R.V. Kamat, K.T. Pillai, N. Reghu, V.N. Vaidya, D.D. Sood, 8th Annual Conference of Indian Council of Chemists, Tirupati(1988). ,. Synthesis of uranium mononitride microspheres by sol-gel process, S.K. Mukerjee, J.V. Dehadraya, Y.R. Bamankar, V.N. Vaidya, D.D. Sood, Ibid. 75. Chemical characterisation of plutonium fuels, D.D. Sood, Indo-German workshop on Techniques for Materials Characterisation, Hyderabad, 1988. 76. Experience in BARC on the preparation of gel microspheres of uranium, thorium and plutonium, D.D. Sood, V.N. Vaidya, An Invited talk at the International Symposium on 'Nuclear fuel Fabrication-1988', Bombay, December 1988. 153 77. Chemical quality control of nuclear fuels, S.K. Patil, Invited talk, Ibid. 78. Anubhati ke Indhan mein Rasayanaki ka yog, D.D. Sood, Seminar on ' Rasayanik Vigyan ke Lfbharte Kshitiz ', Bombay, Sept. 1968 79. Role of chemistry in nuclear technology, D.D. Sood, National Workshop on Nuclear and Radiation Chemistry, Poona, September(1988). 80. Sol-gel process for nuclear fuel fabrication - Experience at BARC, D.D.Sood Invited talk at the International Conference on ' Nuclear Fuel Fabrication ', BARC, Bombay 1988. 81. The role of chemistry in the fuel for nuclear power reactors, D.D. Sood, (Invited talk in Hindi) Hindi Symposium on Emerging Frontiers in • Chemistry, Bombay(1988). Published by : M. R. Balakrishnan Head, Library & Information Services Division Bhabha Atomic Research Centre Bombay 400 085s u • I IIK): -H ! 1. r . i>.n jraniuin nitride sample (CN-13) from Piioi L.'.v -• e 1 1 p: !• ;it :."•>'<:( i 1 i- >:•-•••. ••'', analysis of N - con ter, t 5.17 percent with a standard de v i a t i o.-, of ± 0.07 i.e. 1.4% in 17 determinations. The dissolution tin-.' v.as much less when the sample was powdered. But the sample ground it; air atmosphere gave low recovery suggestimj loss of nitrogen by oxidation in air. It was also observed that addition of 1 drop of 10% HF in the dissolution mixture enhanced th> rate of 117