Inorganic Ion Exchangers and Adsorbents for Chemical Processing

IAEA-TECDOC-337

INORGANI EXCHANGERN CIO ADSORBENTD SAN S FOR CHEMICAL PROCESSING IN THE NUCLEAR FUEL CYCLE

PROCEEDINGS OF A TECHNICAL COMMITTEE MEETING ON INORGANI EXCHANGERN CIO ADSORBENTD SAN S FOR CHEMICAL PROCESSIN NUCLEAE TH GN I R FUEL CYCLE ORGANIZEE TH Y DB INTERNATIONAL ATOMIC ENERGY AGENCY AND HELD IN VIENNA, 12-15 JUNE 1984

TECHNICAA L DOCUMENT ISSUEE TH Y DB INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1985 INORGANI EXCHANGERN CIO ADSORBENTD SAN S FOR CHEMICAL PROCESSIN NUCLEAE TH N GI R FUEL CYCLE IAEA, VIENNA, 1985 IAEA-TECDOC-337

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Radioactive off-gases and low- to high-activity liquid and solid wastes arise froe operatioth m f poweo n r reactors, spent fuel reprocessing plants and other nuclear fuel cycle facilities. The safe treatment and disposal of radioactive wastes is, therefore, an important engineering task. So is the extraction and recovery of useful fission product d actinidean s s from various sources. Inorganic ion-exchangers and adsorbents have been receiving attention for these purposes because of their strong chemical affinity, high retention capacit r certaifo y n radionuclide d higan s h resistanc o irradiationt e .

In response to the growing interest in these topics, the IAEA convened the Technical Committee Meeting on "Inorganic Ion-Exchangers and Adsorbent r Chemicafo s l Processin e Nucleath n i gr Fuel Cycles it t "a Headquarters from June 12 to 15, 1984 with the attendance of 29 participants fro Membe5 1 m r States. This Technical Document contains the 20 papers presented during the Meeting. A variety of inorganic ion-exchangers and adsorbents have been reported: zeolites, ammonium molybdophosphate, magnetite, sodium titanate, zinc-charcoal mixtures, polyantimonic acid, titanium phosphate, zirconium phosphate, hydrated alumina, silica, bentonites, mordenite, hydrous titanium oxidd an e others. The potential usefulness of these inorganic compounds has been prove n varioui d s area f nucleao s r fuel cycle technology, especialln i y the separatio d fixatioan n f fissioo n n product e d actinideth an s n i d an s treatment of effluents from nuclear power plants and of dissolver off-gases. The use of the inorganic ion-exchangers and adsorbents can entail many advantages over conventional processe e areath f o sn i s radioactive waste treatment and in the recovery of fission products and of actinides.

e AgencTh y wishe e scientists o th thant s l al k , engineerd an s institutions who contributed to this Meeting with their papers and their participation. Special thanks are due to the General Chairman, . H.J.AchMr e (Federal Republi Sessioe f Germanyth o c o t n d Chairmen)an , Messrs. L. Szirtes (Hungary), E.W. Hooper (United Kingdom), C. Madic (France), Weng Haomi e (Japan)n Ab (China . M .d )an e officee IAETh th Af o rresponsibl r thifo es Meetins wa g . DivisioUgajie M th f no f Nuclea no r Fuel Cycle. Assistanc preparinn i e g the final document was provided by S. Ajuria.

EDITORIAL NOTE

In preparing this material for the press, staff of the International Atomic Energy Agency have mounted and paginated the original manuscripts as submitted by the authors and given some attention to the presentation. Tïie views expressed papers,the statementsin the general the made and style adoptedare the responsibility of the named authors. The views do not necessarily reflect those of the govern- ments of the Member States or organizations under whose auspices the manuscripts were produced. The use in this book of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities institutions and delimitation the of or theirof boundaries. The mention of specific companies or of their products or brand names does not imply any endorsement or recommendation on the part of the IAEA. Authors are themselves responsible for obtaining the necessary permission to reproduce copyright material from other sources. CONTENTS

Introduction ...... 7 Sodium titanate - A highly selective inorganic ion exchanger for strontium ...... 9 /. Lehto.J.K. Miettinen The effect of gamma radiation on various inorganic ion exchangers ...... 19 SzirtesL. separatior TechnologS d an s disposarolC n d i f yean o l strateg higf yo h level waste 1 ...... 3 . L.H. Baetsle Evaluatio f zeolitno e mixture r decontaminatinsfo g high-activity-level watee th t ra Three Mile Island uni nuclea2 t r power station ...... 3 4 . E.D. Collins, D.O. Campbell, L.J. King, J.B. Knauer, R.M. Wallace Applicatio f silver-freno e zeolite removo st e iodine from dissolver off-gase spenn i s t fuel reprocessing plants ...... 3 5 . T. Sakurai, Y. Komaki, A. Takahashi, M. Izumo Radioactive ruthenium removal from liquid wastes of "Mo production process using zincharcoad can l mixture ...... 3 6 . R. Motoki, M. Izumo, K. Onoma, S. Motoishi, A. Iguchi, T. Sato, T. Ito Ion exchange characteristics of hydrous oxides and their use in the separation of uraniu thoriud man m ...... 5 7 . N.Z. Misak, H.N. Salama, EM. Mikhail, IM. El-Naggar, N. Petra, H.F. Ghoneimy, S.S. Shafik Inorganic sorbents in spent resin incineration — The ATOS process ...... 85 C. Airola, A. Hultgren Precipitated magnetite in alpha effluent treatment ...... 97 K. Hording The application of inorganic ion exchangers to the treatment of alpha-bearing waste streams ...... 113 HooperW. E. Separation and purification of fission products from process streams of irradiated nuclear fuel ...... 133 S.A.H.J.AU, Ache Efficacy of argillaceous minerals used as a migration barrier in a geological waste repository ...... 3 14 . E.R. Merz . Ten years of experience in extraction Chromatographie processes for the recovery, separatio purificatiod nan f actinidno e elements ...... 7 15 . C Madic, J. Bourges, G. Koehly Synthetic inorganic ion exchangers and their potential use in thé nuclear fuel cycle ...... 173 A. Clearfield Fundamental properties of acid salts of tetravalent metals and some considerations on the perspectives of their application in nuclear technology ...... 195 G. Alberti Recover f cesium-13yo 7 from radioactive waste solution scomple w witne ha x inorganic ion exchanger ...... 213 Zhaoxiang,S. Zhigang,T. Zun,H. L Zhenghao, Haomin,W. Taihua,L. BoliL. Treatment of liquid wastes containing actinides and fission products using sodium titanate as an ion exchanger ...... 223 Y. Ying, Meiqiong,L. XiannuaF. Ion-exchange propertie f zeoliteso theid san r applicatio processino nt high-levef go l liquid waste ...... 237 T. Kanno, MimuraH. Studie inorganin so exchangern cio s — Fundamental propertie applicatiod san n ni nuclear energy programmes ...... 249 R.M. lyer, A.R. Gupta, VenkataramaniB. Ion-exchange selectivitie antimonin so c acid metad san l antimonates ...... 3 26 . M.Abe Panel discussion ...... 277 Lis participantf o t s ...... 9 27 . INTRODUCTION

In spite of a slowing down of construction of new power plants in comparison o previout s forecasts, nuclear energy continue a prove e b o nt s sourc f electricito e n essentiaa d yan l componen e energth f yo t balancf eo many countries. By the end of 1983 there were 317 nuclear power reactors in operation, which produce e world'th df o abous % electricity12 t e Th . total capacity of these reactors was 191 gigawatts (GWe). The IAEA now estimates thae totath t l installed nuclear capacit worle th dn y i will amount to 275 GWe by 1985, to 380 GWe by 1990 and between 380 and 850 GWe by the turn of the century.

This developmen f nucleato r power will e depensafd th an en do efficient operatio f poweno r reactor d supportinsan g fuel cycle facilities. Nuclear power production generates actinides, fission product d activatioan s n products tha te safelb hav o t ey manageo t d protect the public. Radioactive off-gases and low - to high-activity liqui d solian d d wastes inevitably arise froe operationth m f poweo s r reactors and plants for uranium ore processing, fuel fabrication, spent fuel reprocessing and other nuclear fuel cycle activities.

Waste treatment is, therefore, one of the great engineering tasks thae nucleath t r industr o facet s .yha Other important o taskt e sar extrac d recovean t r useful fission product d fissilsan d fertilean e materials from irradiated nuclear fuel d othesan r sources.

In respons e growinth o t eg interes n thesti ea resul topics a td san of the increasing use of nuclear power, the IAEA decided to hold this Technical Committee Meeting within the framework of the Agency's activities on nuclear materials and fuel cycle technology. The applicatio f inorganio n c ion-exchanger d adsorbentan s boto t s h waste treatment and the recovery of fission products and actinides were of primary concern in this meeting. Because of the importance of the scientific knowledge neede o develodt p successful technical processes, the meeting covereo majotw e r th dfield f fundamentao s l studied san industrial applications. s beeIha t n shown that inorganic ion-exchanger d adsorbentan s s have high selectivity and high exchange capacity for certain elements which e importanar e nucleath n i tr industry. Moreover, thee virtuallar y y non-combustibl d generallan e y immun o radiatiot e n damag d biologicaan e l degradation, which are significant advantages over the organic ion-exchangers currentl n usei y . Ther e reasonablar e e possibilitieo t s use inorganic exchanger r processefo s s involving extraction, purification and retention of radioactive materials. Because of their versatility, inorganic ion-exchangers and adsorbents will play an important role in nuclear fuel cycle technolog e neath r n i futurey .

The main objectives of the meeting were to provide an opportunity for experts involved in chemical processing in the nuclear fuel cycle to exchange informatio n experimentao n l result d experiencean s o t d an s discus e specifith s c problems facing them e .firs th Thits meetinwa s g organize e Agencth o deay t yb d l with these topics. SODIUM TITANAT HIGHLA E - Y SELECTIVE INORGANI EXCHANGEN CIO STRONTIUR RFO M

J. LEHTO, J.K. MIETTINEN, Department of Radiochemistry, University of Helsinki, Helsinki, Finland

Abstract A strontium-selective inorganic ion exchanger, sodium titanate, has been synthetized from an industrial intermediate of a titanium dioxide pigment process. The distribution coefficient of sodium titanat strontiur efo highs H mi p a , t nearla 0 y1 above 7. The sorption capacity is also high: 1.4 mmol of strontium per gram of exchanger. The effect of high concentrations of sodium chloride, boric acid, sodium citrate and EDTA on the sorptio bees nha n studied sorptioe .Th n rat strontiuf eo n mi sodium titanat bees eha n rathefoune b o dt r - highex e .Th change highls ri y resistan gammo tt a radiation .gammA a irradiation dose of 10 Gy has practically no effect on its structure and sorption properties. Decontamination factors higher thahav0 n10 e been obtaine strontiur dfo m froactuao mtw - lnu clear power plant waste solutions. 1. Introduction pase Inth t twenty year greasa t numbe synthetif ro c inorganic ion exchangers have been developed for processing nucle- wastr a e solutions. Inorgani exchangern cio s have many important advantages in this field.' First, most of them withstand high temperatures and radiation doses. Thus, unlike organic ion exchange resins, they can be loaded with radionuclides to higa utilizee b treatmene h n th degresolu t ca n ho di d -f ean to tions. Second, som themf eo , like titanate zeolitesd san n ,ca be ceramized to a solid final waste product. The best known ceramic waste productitanate th s ti e mineral based SYNROC (1). Third, many inorganic ion exchangers are highly selective to certain key radionuclides, like some zeolites are to cesium. Sodium titanat knoweffective s b i e separatioe o nt th n ei n of strontium. However, it has often been synthetized from expensive chemicals like titanium tetrachloride and titanium alcoxides (2,3). In this paper is described a sodium titanate product whic produces hwa d fro intermediatn ma indusn a f eo - trial titanium dioxide pigment process. The raw material is chea readild pan y available synthesise .Th wels , a sorps la - tion arid radiation resistance propertie sodiuf so m titanate, are described. 2. Experimental Sodium titanat synthetizes ewa d from "titanium hydrate" whicintermediatn a s hi e fro titaniue mth m dioxide pigment process, the so called sulphate process used at a plant of Kemira Co., Finland. The raw material was slurried in hot buta- sodiunod lan m watehydroxid% w- r0 3 solutionaddes a ewa s da .

The mixture was stirred at 9C3 for a few hours. After the product had settled, it was filteredU , washed with water and finally dried at 105 °C (4). The sorption capacity ofpodium titanate for strontium and the stoichiometry of ttje Sr - Na exchange were studied using column experiments. Sr solution (0.01-M, pH 5) was pumped at- a flow rate of 0.6 ml/min through 3,0 g sodium titanate columns. The grain size of the sodium titanate was 0.315-0.850 mm, as in all other experiments. The Sr + and Na+ concentra- fractionl m 8 1 e tionsth n collectesi d froeluate mth e were measure meanatomiy n db a f so c absorption spectrometer (AAS). Usin eluate gth e volume + concentratio,r wherS e % eth 0 5 s nwa (50 % breakthrough) of the concentration in the feed solution, the sorption capacit calculates ywa d wit followine hth g formula: V(50 %} x C Capacity (mmol/g) = —————————, where V(50 %) = eluate volume at the 50 % breakthrough C = Sr concentration in the feed solution (0.01-M) mo = exchanger mass (3.0 g)

Th + eexchang Na stoichiometr - + calculates r eS wa e th f dyo by comparingthe total amoun+r S absorbe f totae to th lo dt

amount of Na + eluted from the column.

The distribution coefficient (1C) of strontium in sodium titanate determines functiowa a , s ,a pH f dno wit batcha h method as follows: 0.2 g samples of exchanger were shaken with

20 ml of 0.0005-M S rsolution containing Sr tracer. The solutione pH'th f so s were adjuste differen+ o dt t values using nitric acid. Afte houro rtw s shaking mixturee ,th s wereggentri- solutione pH'th e th f fugeso d sdan were measuredr S e .Th activitie solutione th f so s were measured usincrystal Na ga l and a single channel analyser. The distribution coefficients were calculated wit followine hth g formula: A -A V «D (ml/g) = "V x m»

whereftt- 8S= AS r activit e solutioth n yi n before shaking A° = Sr activity in the solution after shaking V = solution volume (20 ml) m = exchanger mass (0.2 g) The effect of boric acid on the sorption of strontium as a function of pH was studied at two different boric acid concentrations. The experiments were carried out as described above for K_ measurements, but the 0.0005-M Sr solutions were either 0.5- 1.0-r Mo M with regar borio dt c acid. The effect of different sodium ion concentrations on the sorptio strontiuf no testes mwa stirriny db sampleg 0 g1. s of sodiu 0.0005-f o ml m titanat r 0 MS solution 10 n ) ei 5 H s(p which where 0.0005, 0.005, 0.05, 0.5, 2.0 or 4.0 molaric with regar sodiuo dt m chloride (5) distributioe .Th n coefficients were calculated as described earlier.

10 The effect of organic complex forming agents, EDTA and sodium citrate sorptioe th n ,o strontiuf no studies mwa s da a function of pH at two different EDTA and sodium citrate concentrations. The experiments were carried out by stirring 1.0 g sample 0.0005-f o exchangel f m so r 0 MS solution 10 n ri s differenf o values H tr p S solution e .Th s contained these complex forming agent molaritien si 0.0f so 0.005r 5o e .Th Sr + concentrations of the solutions before and after two hours stirring were measured using AASsorptioe Th . n percent- ages were calculated with the formula: (A -A/A ) x 100 %, strontiue th whers i eA m concentratio solutione n th lef n i t s after stirring and A is the initial strontium concentration in the solutions (4)? The sorption rate of strontium in sodium titanate was studie stirriny db sampleg f 0 o g1. l exchangef m so 0 10 n ri Sr + solution. Aliquots of 1 ml volume were drawn from the solution at different time intervals and the Sr activities aliquote th f o s were measured describes ,a d earlier sorpe .Th - tion half times T ,_ were calculated as the times during which the concentratio strontiur no solutioe th n mi n decreasea y sb factor?of twoeffect e exchangee Th .th f so r grain.sizf o d ean + concentratiothr eS n were tested (5). The effects of gamma irradiation on the structure and the sorption properties of sodium titanate wgre studied. Sodium titanate samples were irradiate sourcedo C wit tota e ha .Th l absorbe dose th e d ratd an kGy/h5 e3 gammy G .a0 1 dos s ewa Afte irradiatione rth structure ,th studies ewa d using X-ray diffraction spectrometriR ,I thermoanalyticad can l methods. The specific surface area was determined. The effect on the sorptio strontiuf no s testemwa determininy db distrie gth - bution coefficient as a function of pH as described earlier (6). The efficienc sodiuf yo m titanatseparatioe th r efo f no strontium froactuao mtw l power plant (Loviisa, Finland) waste solutions was tested. The waste solutions contained high amounts of these inactive salts: boric acid, -sodium and potassium ions totae .Th l salt g/12 e content d Th . an 1 sg/ wer5 e21 pH values of the solutions were 12.5 and 10.0, respectively. The solution with pH 12.5, having the higher salt 'content, is a waste concentrate obtained by evaporation from the solution with pH 10.0. This solution originates from various sources like leakages in the primary circuits, and from decontamina- tions. Sr tracer was added to these waste solutions. Samples of 450 ml volume from the solutions were eluted through sodium titanate columns of masses of 1.0 g and 2.0 g. The élution rate was 1 ml/min. The Sr activities in the eluates were measured usin Ge(Liga ) crysta multichannea d lan l analyzer. The decontamination factors (D.F.) were calculated as the strontiuratioe th f so m concentratio feee th d n nsolutioi o nt the concentratio eluatee th n ni .

11 3. Results and discussion

Fig. 1 shows a breakthrough curve of strontium from a sodium titanate column. The breakthrough curve depicts the concen- eluate functioa th e s n ea th i f + no Na d an tration + r S f so eluate volume.

E EFFLUENTTH N I * ,No Sr-2-* D AN - l/1

20 .

IB .

12 .

4 ]

^o—_„o——- 200 400 BOO 800 VOLUME OF THE EFFLUENT, ML Fig. 1. A breakthrough curve of strontium from a 3.0 g sodium titanate column. Sr concentration in the feed solution: O.Q106-M, pH: 5. Elution rate: 0.6 ml/min. Solid line (+) = Sr , dashed line (o) = Na+.

The sorption capacity of sodium titanate for strontium, calculated using the 50 % breakthrough value of strontium, was fairly high 4 1 +mmol/g_,1. 0. .

2+ clearle b n Inca yFigt i see .1 n that+wher nS was ab- sodiue sorbeth n mi d titanate columnsimultaneousls wa a ,N y eluted out. when Sr began to break through, the Na concentration in the eluate decreased respectively. The molar ratio of Na+ eluted from the column to Sr absorbed in the column 0 1 +average_2. 0. s wa d froexperimentso mtw . This ratid oan the curveshape'e th Fign f si o ,indicat1 s^oiciometriea c r chemicaS ion e a ionN on exchangn .o y io lsb tw f eo

Fig. 2 shows the distribution coefficient Kp of strontium in sodium titanate as a function of pH. At the acidic region increaseC ,1 s steepl reached yan constansa t high value abovH p o. a fe7 t nearla 0 y1

12 5 logKD -*-

. 4

3 .

./ s pH l 23456789 10 11 12

Fig. 2. The distribution coefficient K~ of Sr in sodium titanate, as a function of pH. The initial Sr concentration in the solution 0.0005-M.

Boric acid is used as a neutron absorber in nuclear power reactors and thus it might be found in high concentrations medium-leved n lowi an - l power plant waste solutions- ef e Th . fec f borio t ce sorptio th aci n do f strontiuno mn sodiui - mti tanate is shown in Fig. 3. Up to pH 6 there is no effect. Above thi valueH sp e boric th , , acid concentratio f 1.0-no M reduce e distributioth s n coefficien factoa y b f 10.to r '

123456789 10 11 12

.Fig. 3. The distribution coefficient 1C of Sr in sodium titanate, as a function of oH, at two different boric acid concentrations. The initial Sr + concentration in the solution: 0.0005-M. * = 0-M H BO o - 0.5-M H B0q (dashed line), + = 1.0-M

13 Nuclear waste solutions often contain high amountf so sodium. A high concentration of 4.0-M of sodium decreases the distribution coefficien strontiuf to sodiun mi m titanate only by a factor of 10, as can be seen in Fig. 4. Dosch obtained similaa r result with sodium titanate synthetizee th n di Sandia National Laboratories (7).

5 D IDK

104i

1D3J

CNa ]

. . D-D-1 10 DMM. OD1-M 1- . Dl- 0 M. 0 M

Fig. 4. The distribution coefficien sodiun i tmr S ^f o ?

titanate functioa s ,a Na concentration f no initiae .Th r lS i ione Th . concentratio+ n in the solution: 0.0005-M, pH: 5 (5). In order to fix waste nuclides in inorganic ion exchangers, the nuclides first elute e havb o et d from spent organic reac- resinsr to . This elutio carriee usiny b b n t gnca dou organi c complex forming agents. Methods base thin do s principle have been developed in Sweden (8). Figs. 5 and 6 show the effects of two complex forming agents on the sorption of Sr in sodium titanate. EDTA decreases sorptio greaa o nt t extentd ,an excludee b n possibla thuca s dt a si e eluting agent. Sodium pars citratit t r affectefo s sorption only slightl pH<6t ya .

SORPTION% , 100. -H————!————l————f

W —— 80.

60 l I 40.

20.

7 B 5 4 3 2 B 2 g1 1 11 0 Fig. 5. The sorption percentage of Sr"2+' in sodium titanate, as a function of pH, at two different EDTA concentrations. The initial Sr + concentration in the solution: 0.0005-M. + = 0-M EDTA, o = 0.005-M EDTA (dashed line), * = 0.05-M EDTA (4)

14 SORPTION% ,

100 jj «. O • O~*"'O O"^£ **———*———•*»„ , 80

123456789 10 11 12 sorptioe Th Fig . 6 .n sodiupercentagn i mr S titanate f eo , as a function of pH, at two different sodium citrate concentrations. The initial Sr + concentration in the solution: 0.0005-M 0.005-= .o M sodium citrate 0.05-= ,* M sodium citrate (dashed 1ine)(4).

Sodium titanat highls ei y resistan gammo t a radiation. Practicall chango yn s structurs it s foun it en wa i n di d ean specific surface area, when sampleexchangee th f so r wer- eex posed to a gamma irradiation dose of 10 Gy. No effect on the sorption of strontium was found either, as can be seen in Fig. 7. Kenn adecreas% founstrontiue 0 5 th da n ei m capacitf yo the sodium titanate synthetize Sandie th n di a National Labora- tories after having irradiate exchangee dth r wit gammha a irradiation dose of 2 x 10 Gy (9).

1-9*0

. 4

3 i

H 2 3 4 5 6 7 8 9 10 11 12

Fig. 7. The distribution coefficient K^ of Sr 2+ in sodium titanate functioa whe , s exchangee ,a pH nth f no expos rwa : tgammoa a irradiatio y (dasheG n0 1 dosd f line(o)eo d )an when not exposed (solid line (*)). The initial Sr concen------tration in the solution: 0.0005-M (6).

15 The sorptio sodiun i + nmr S rat titanatf eo satisfactos ei - rily hig exchangee h th considerin f o column e ri us e ngth opera -

tions. Sorption half time of 1.6 +_ 0.2 min was obtained for sorption fro 0.001-ma +r MS solutio sodiun ni m titanate with graia n siz 0.315-0.85f eo hale Th f . tim0mm e decreases when smaller grain sizes are used and increases when the Sr concentration gets higher. The experiment separatioe th n so strontiuf no o mtw froe mth nuclear power plant waste solutions, using sodium titanate, gave satisfactory results decontaminatioe Th . n factord san the volume reduction factors obtained are compiled in Table I. decontaminatioe Th Tabl . eI n factor volume sth (D.F,d ean " reduction factors (v.R.F.) obtained frocolume mth n experiments using sodium titanate for the separation of strontium from two waste solutions taken from the Loviisa nuclear power plant, Finland. 450 ml of the solutions were eluted through the columns. The flow rate was 1 ml/min.

Salt contenf to e Masth f so D. F. V. R. F the waste solution exchanger

1 g/ 5 21 1 P 40 + 10 280 215 g/1 2 g 170 + 40 140 2 g/1 1 g 130 + 20 280 2 g/1 2 g 270 + 50 140

4. Conclusions

Sodium titanate is highly selective toward strontium. The capacity and the distribution coefficient are high, remaining sufficient eve higt na h concentration borif so c acisodiud dan m chloride. Thus this exchange suitabls ri separatior efo f no strontium from nuclear waste solutions containing high amounts inactivf o e salts. industriae th f o The leus intermediate "titanium hydrate" materiaw ra a sodiun s li a m titanate synthesis reduces preparation costs significantly compared with other synthesis methods. "Titanium hydrate" is available in large amounts and is cheap. Furthermor synthesie eth sodiuf so m titanate from "titanium hydrate" is rather straightforward.

16 Separatio wastf no e nuclides using sodium titanatn a s eha additional advantage in that the loaded exchanger can be turned into a solid final waste form. One method is hot press- ing used exampler ,SYNROfo e th n C,i process. Another method developed in our laboratory includes mixing the titanate with clafirind yan mixture solia gth o t 100t dea for0C m (10). This method also give highlsa y resistant waste form.

5. Acknowledgement This stud supportes y wa Ministr e th y db Commercf yo d ean Industry, Finland.

6. References

(1). Ringwood, A.E., et al., Nucl. Chem. Waste Manag. 2(1981)2.87. (2). Forberg Olssond an . ,S , P.-J., Metho Preparinf do g Titan- ates Suitable as Ion-Exchange Material, U.S.Patent . 4.161.513, 1979. (3). Lynch, R.W., et al., The Sandia Solidification Process- BroaA d Range Aqueous Waste Solidification Methodn ,i Managemen Radioactivf to e Wastes froNucleae mth r Fuel Cycle, IAEA, Vienna, 1976 361. ,p . (4). Heinonen, O.J., Lehto, J. and Miettinen, J.K., Radiochim. Acta 28(1981)93. (5). Lehto, J., Heinonen, O.J. and Miettinen, J.K., Radiochem. Radioanal. Letters 46(1981)381. (6). Lehto, J. and Szirtes, L., Radiochem. Radioanal. Letters 50(1982)375. (7). Dosch, R.G., Final ApplicatioRepore th n to Titanf no - ates, Niobates and Tantalates to Neutralized Defense Waste Decontaminatio Materialn- s Properties, Physical Form Regeneratiod san n Techniques, Sandia National Labora- tories, Report SAND-80-1212, 1980. (8). Forberg, S., et al., Fixation of .Medium-Level Wastes in Titanates and Zeolites: Progress Towards a System for Transfer of Nuclear Reactor Activities from Spent Organic to Inorgani Exchangersn cIo Scientifin ,i c- BasiNu r sfo clear Waste Management, Vol. 2., ed. C.J.M. Northrup, Plenum Publishing Co., 1980 867. ,p . (9). Kenna, B.T., Radiation Stability of Sodium Titanate Ex- changer Materials, Sandia National Laboratories, Report SAND-79-0199, 1979. (10).Lehto, J., Heinonen, O.J. and Miettinen, J.K., Ceramiza- tion of Inorganic Ion Exchangers Loaded with Nuclear Waste intClad oRe y Tiles Scientifin ,i c- BasiNu r sfo clear Waste Management VI, ed. D.G. Brookins, Elsevier Science Publishing Co., 1983 589. ,p .

17 THE EFFECT OF GAMMA RADIATION ON VARIOUS INORGANIC ION EXCHANGERS

. SZIRTEL S Institut Isotopesf eo , Hungarian Academ f Sciencesyo , Budapest, Hungary

Abstract

Thy eG effec 0 1 gammf o o tt a0 1 radiatio x range 5 th f eo n i n 3 7 s beeha n followine studieth r fo d g inorganic ion-exchangers: amorphous zirconium, titaniu d chromiuan m m phosphates, zirconium telluratd an e crystalline zirconiu ceriud an m m phosphates e solubilitTh . acidin i y c solutions, the ion-exchange capacity and the specific surface area of these compounds were examined before and after irradiation. Zirconium phosphate has a very stable crystalline structure which remains unchanged during the irradiation process. The crystal lattice of cerium arsenates is partly destroyed during irradiatio e relativelth t bu n y high stability O bande As groupo th ff o s s prove tha o chemican t l change accompanies this process e ceriuTh . m phosphate crysta completels i l y destroyed an d becomes amorphous after irradiation.

Usin n exchangerio g n radiochemicai s l practic mann i e y cases they are exposed to high doses from charged particles and/or gamma radiation e organicTh . n exchangerio e submittesar d to radiolysis which results in changes of different properties, i.e. swelling, capacity etc r exampl.Fo Russiaa e n anion exchanger type AV-1 originas it 7f losseo l^ capacit50 d y after irradiation with a dose of about 5x10 Gy (l). Other observation n thii s s field were collectee th d n latei n o r boo Yegoroff o k f (2). These facts made difficultie usinn i s g ion exchange procedure n radiochemicai s l practice t leasa , t in cases when high activity level exists.

It i s a pitiable fact that in contrast to the general view that all inorganic ion exchangers have a good resistance against both charged particles and gamma-rays only some works were found treating this subject (3-5).

19 Therefore, several years ago we made some investigations on amorphous chromium, titanium, zirconium phosphates and circonium tellurat n crystallino d an e e zirconium, cerium phosphate ceriud an s m arsenate (6-9).

EXPERIMENTAL

The amorphous zirconium, titanium, chromium phosphates and the zirconium tellurate were synthetized by the methods describe s earlieu y b d r (10-12) e crystallin.Th e zirconium phosphat s prepare methoe wa e th f Clearfiely o db d d (13), •hfeile cerium phosphat d arsenatan e e were synthetized according to Alberti (14-15). The samples were irradiated with gamma-rays using the K-80 type gamma irradiation facility of the Institute of Isotopes (source: Co, dose rate 0 1 :Gy/hr) e sample.Th s were irradiate variour fo d s timy than sucwa i e t a h they were expose o 5x1t d , 0 46? y doseG 0 ,1 respectively d an 0 1 1, 0 . Doses were controlled by chlorobenzene dosimetry (17). All experiments were carried out after irradiation in parallel on the treate d originaan d l samples. The amount e componentth f so s were determine y analyticab d l chemical methods, i.e. zirconium, titaniu d chromiuan m m by usual gravimetric method as oxides; cerium was determined spectrophotochemically (l?) while phosphat d telluratan e e colorimetrically (l8) and arsenic iodometrically in the usual way. The X-ray diffractograms were obtained using a DRON-2 type high-angle powder diffractometer with CuK (Ni-filter) radiatio 0,1541= X ( n 8 nm). a The i.r. investigations were carrieroot a mt temperaturou d e (200-400 usin) " Perkin-Elmea g0cm typ5 22 re spectrophoto- meter e spectre soliTh .th f do a samples were obtained from mults in Nujol, hexachlorbutadiene and Voltalef. The weight losses were determined by thermal analysis usin typM MO ge derivatograp n temperaturi h e interval 25-1000° heatina y Cb g rat f 5°C/mino e .

The specific surface areas of the amorphous materials were determined by the krypton adsorption method, using a volumetric gas adsorption apparatus at liquid nitrogen

20 temperature and were analysed by the BET method with a computer program. uptakn io s examinee wa e Th d using tracer techniqus a e follows: several samples of the ion exchangers (0.1 g. each) were equilibrated with 10 ml of 0.2 M (MCI+ MOH) solution, where M denotes Na*, K+, Rb+ and Cs respectively. After shakin t rooa g a temperature till attaining equilibrium (usuall o days)tw y amoune th ; alkalinf o t e metals in the samples of the supernatant liquid were determine measuriny b d e radioactivitieth g e initiath f o s l d equilibratean d solutions, while phosphate r assath fo ef o y , tellurate and arsenic content the methods mentioned before were used; the process was controlled and the pH measured by Radiometer type automatic titrimeter r labellinFo . g fla, T£, Kb and Cs nuclides of high specific activities were used.

Result d Discussioan s n

The ratios of the ligands to the central metal ions of the examined materials - both initial and irradiated - and the formulae calculated from the analytical data are collected in Tabl. 1 e

The solubility data in various acidic solvents are collected in Tabl. 2 e

valuee Th f specifiso c surfac amorphoue e th are f o a s materials are given in Table 3.

21 Table 1.

Ratio f ligands/metao s r varioufo l s matériels

n exchangeIo r Formulae Ratios original irradiated

a. ) amorphous

Zirconium phosphate (ZPa) Zr(lIPO- O Hg ....;i )x 2 . , 1.95 1.V5

Titanium phosphate (TP) Ti(lIP04)2 x 1.32 1I20 1.90 1.85

Chromium phosphate (CrP) Cr gO (lIPO ^) g x 3HgO 1.10 1.04

Zirconium tellurate (ZTe) ZrO(]ITeO^)2 x 21i00 1.22 1.20

b. ) crystalline

Zirconium phosphate (Z'P) Zr(lIP04 )x 1.33»2 20 2.00 1.98

Cerium phosphate (CeP) Ce(llPO x 0.331;i) 0 1 1.99

Cerium arsenate (CeAs) Ce(llAsO;j )x l.UOHg2 O 2.01 1.97

"absorbe. Gy 0 1 dx 1 dos s i e

Table 2.

Solubility data in various acidic solvents (mmole PO.orTeO, or As/g, ion exchanger)

II SO ÜC1 1IC10 n exchangIo e materials Remarks 0.05M 0. 5-^ 0.1*. 1.0" 0.1* l.O*»

ZP. original 0.04 0.05 0.02 0.03 0.02 0.02 a in all irradiated 0.05 0.05 0.02 0.02 0.03 0.03 casee th s TP original 0.05 0.06 0.04 0.07 0.05 0.07 irradiated 0.05 0.05 0.06 0.06 0.07 0.06 solubility of tbe CrP original 0.12 0.10 0.12 0.11 0.12 0.13 Materials irradiated 0.11 0.10 0.12 0.13 0.13 0.11 s strougli y ZTe original 0.08 0.07 0.08 0.10 0.09 0.07 increased irradiated 0.10 0.08 0.09 0.09 0.07 0.07 above ZP original 0.02 0.05 0.02 0.01 0.02 0.03 pll > 8 irradiated 0.03 0.05 0.03 0.02 0.02 0.03 CeP original 0.27 0.28 0.20 0.40 0.18 0.25 irradiated 1.57 3.24 1.60 2.42 1.65 1.78 CeAs original 0.08 0.61 0.12 0.25 0.10 0.22 irradiated 0.29 0.92 0.34 0.37 0.35 0.39

22 Tabl. 3 e

Specific surface area values

2 Ion Exchanger S(a /B) before after irradiation

Zirconium phosphate 10.8 10.0

Titanium phosphate 8.7 1.3 Chromium phosphate (CrP) 26.8 2.12

'Zirconium tellurate (ZT0 ) 1G.5 1.0

The thermoanalytical dat f amorphouo a s material e showar s u in Fig. 1. (dotted lines for irradiated samples).

meg/g ' 5 r No i •

10pH 2 H 2 p 10p0 1 H 10p 2 H

meg/g Cs* . Rb*

1 rir****

10pH 2 »pH 2 W pH 2 »pH

CP ZT meg/g

10pH 2 XJpH 2 10pH 2 10 pH

meg/g Rb Cs Rb Cs' \S 5-

"iOpH 10pH 2 10pH 10pH

TP uptakn Io alkalf o e . i1 F. metal ig amorphoun o s s materials

23 n uptakio e eTh datr alkalinfo a e metal e showar sn i n Fig. 2. From the data it can be seen that no characteristic changes occur in solubility, the ion uptake and thermal behaviour of the zirconium, titanium phosphates and zirconium tellurate.

1000

100 -

Fig. Thermogram f amorphouo s s materials (dotted line for irradiated samples)

In cas f chromiuo e m phosphate unde e irradiatioth r e th n •specific area value decreased abou e ordeon t f magnitudo r e and the exothermic process existing at 250 C on the thermogra e initiae founb th n o f o dt o m lt samplno s wa e that of the irradiated one.

This phenomenon can be explained by supposing that the

polimer chainjVCr -HPO. -Ür/-H2PO^] ^ was broken during the irradiation process n casI . f crystallino e e materials further X-ra d i.ran y . spectrometric investigations were carried out e result.Th f thesso e measurement e showar s n in Fign.3 and 4. e seeb n ca afte t Ai sr irradiation zirconium phosphate revealed exactl e samth ye diffractogra e originath s a m l sample n casi ;f ceriueo m arsenate some extra very weak

24 V—s

Fig. 3* X-ray diffractograms of crystalline materials 1-zirconium phosphate, 2-irradiated zirconium phosphate, 3-cerium arsenate, 4-irradiated cerium arsenate, 5-cerium phosphate, 6-irradiated cerium phosphate. peaks appeared and the diffractogram of cerium phosphate contained no peaks et all. The i.r. spectra of zirconium phosphate (Fig. 4) shoved practiall o differencyn e betwee e initiath n d irradiatean l d samples n casI .f ceriuo e m arsenate changee s th onl n i y characte As-Oo bondt e e th H du sf group o r s (2880, 2360

and 1245 cm") and H0 (3590, 3485 and 1595 cm") were found. 2

1 1

25 l l i 1 1 i.l 1. L.. . _J

Fig, Infrated spectr f crystallino a e materials 1-zirconium phosphate (both original and irradiated) 2-cerium phosphate, 3-irradiated cerium phosphate, 4-cerium arsenate, 5-irradiated cerium arsenate

r ceriuFo m phosphat e followinth e g changes were observed: absencbane t th 142a d f " o e0 (P-Ocm Hplann i e -1 m deformationc 0 98 banpresencw e t ne th a d a d )f an o e which is due to a P-O-P structure. The weight loss data got by thermal analysis are shown in Fig. 5. As it can be ssen both zirconium phosphate and cerium arsenate have practically the same water content befor d aftean e r irradiation, whil e ceriuth e m phosphate after irradiation had about twice higher weight loss than the initial sample.

26 meg/9 meg/g' meg/g No*

lOpH 2 10 pH 10 pH 1 meg/q meg/g

2 1H 0p 10pH Ce P

2 'OpH 10pH Ce As Fi g . 3 • Ion uptake of alkali metals on crystalline materials (dotted line shows the solubility)

Comparing the results of these three measurements the following conclusion e drawnb n ca :s

1. zirconium phosphat vera s y ha establ e crystalline structure which remains unchanged during the whole irradiation process. e resultTh . 2 s concerning cerium arsenate indicated tha n thii t s case crystath e l lattic s partlwa e y destroyed during irradiatio e relativelth t bu n y high stabilit bande th AsOjf f f o so y ~ groups prove thao tn chemical change accompanies this process.

e ceriuTh . 3 m phosphate crystal lattic s completeli e y destroyed after irradiation and according to the X-ray diffractogram the material became amorphous.

27 100 200 300 iOO 500 600 700 800 *C

Fig . 6 . Weight e crystallinlosseth f o s e materials •-•-•- zirconium phosphate, o-o-o- irradiated phosphate, A-i-*- cerium phosphate, A-A-A- irradiated cerium phosphate, m-«-*- cerium arsenate, a-d-D- irradiated cerium arsenate

Moreove i.re th r . spectra indicat n intensiva e e decrease

of HPO; and H2PO~ groups and presence of P-O-P structural elements.

The observations suggest a resultha s ta f irradiatioo t n the original structure is split and different phosphorus- -oxygen compound e formedar s , whic e usuallar h y rather volatile (l9)> This assumption seems to be strenghtened by the ion uptake (Fig. 6) and weight loss data; i.e. the higher amoun f solublto e materials highee th , r exten f vater-loso t e characteth d san f thio r s process (the weight loss starts at a relatively low temperature and it is practically finished at 350°C). Based on the above mentioned it can be stated that the examined amorphous materials except chromium phosphate do not change their basic behavious under the irradiation process. The investigated crystalline materials showed individual behaviours under their irradiation with gamma rays.

28 Taking into consideration these facts, finishing my lectur wanI e o emphasizt e tha e mus tw t generaliz tno e the good resistanc f inorganio e n exchangerio c s against gamma radiation, but before using them we must investigate their behaviour under the irradiation. We must make this n casi f newlo el firsal yf o tsynhetize d materials.

REFERENCES

] [I KISELEVA, E.D. ,. FizicheskoZ i Khimii (USSR§ )3_ (1962) 2465 [2] YEGQROFF, E.V., NOVIKOV, P.D., Deistvie ionizinyuschih islutseni a ionoobmennih i e matériau Atomizdat Moscov Russiann (i 1965 ) [3] ANPHLETT C.B., Proc. 2nd Int. Conf. on Peaceful Uses of Atomic Energy Geneva 2J3 (1958) 17. [4] YASUSHI, I., Bull. Chem. Soc. Japan 36 (1963) 1316 [5] GROMOV, V.V., SPICIN, V.l., Atomn. Energ. (USSR) 14 (1963) 491. [6] ZSINKA , SZIRTES,L. , STENGER,L. Radiochem, ,V. . Radioanal Letters 4_ (1970) 257 [7] ZSINKA SZIRTES, L. , MINK, ,L. KÂLHA, ,J. , NJ. Inorg. Nucl. Chem. 36 (1974) 1147 [8] SZIRTES, L., ZSINKA, L., J. Radioanal. Chem. £1 (1974) 267 [9] ZSINKA, L., SZIRTES, L., MINK, J., J. Radioanal. Chem. 30 (19769 )13 [10] SZIRTES ZSINKA, ,L. ZABORENKO, ,L. , K.B., IOFA, B.Z., Acta Chim. Hung. 54 (1967) 215 [II] ZSINKA, L., SZIRTES, L., Acta Chim. Hung. 6_9 (1971) 249 [12] SZIRTES , ZSINKA,L. Radiochem, ,L. . Radioanal. Lett. 7 (1971) 61. [13] CLEARFIELD STYNES, ,A. , J.A. . Inorg,J . Nucl. Chem. 26 (1964) 257 [14] ALBERTI, G., COSTANTINO, U., DIGREGORIO, F., GALLI, P., TORRACCA, E., J. Inorg. Nucl. Chem. 30 (1968) 295

29 [15] ALBERTI, G., COSTANTINO, U., DIGREGORlO, F., TORRACCA . J Inorg , E. , . Nucl. Chem^ 31 . (1969) 3195 [l6j HORVÀTH, Z., BÂNYAI, É., FÖLDIÄK, G., Radiochdn. Act 3 (1970-1 a 0 15 ) [17] GREENHOUSE, U.C., FEINBUSH, E.M., GORDON, L., Analyt. Chem 9 2 (1957. ) 1531 [IS] BERNHARDT, D.N-, V/REATII, A.R., Analyt. Chem. 27 (19550 44 ) N WASERVA [19], J.R., Phoshoru s CompoundIt d an sl sVo I pp. 217 Interscience, New York (1958)

30 TECHNOLOG SEPARATIOr S D AN ROLs D C YAN F NEO IN DISPOSAL STRATEGY OF HIGH LEVEL WASTE

L.H. BAETSLE Centre d'étude de l'énergie nucléaire, Mol, Belgium

Abstract The separation of Cs and Sr from the HLW by inorganic ion exchanger s beeha sn extensively studiee sixtie th a treat n i s da s - ment method to separate "long lived" fission products from short lived. This approach needs very high decontamination factora d an s subsequent partitioning of a-emitters. The separation of Cs and Sr in acid medium was thoroughly investigated at S.C.K./C.E.N. Mol e resultanth d s obtaine e brieflar d y discusse e perspectivth n i d e of an over-all waste management strategy. Based on the existing technical information different flow sheets are discussed which identify the results to be aimed at for each of the separation steps involved r .separatioS d Thougan s C hn froW showee b HL m o t d f o little additional advantage s unlesquantitativi t i s e i.e. involving very high decontamination factors, a semi-quantitative separation approach showed to be very attractive in easing the terminal disposal problem. Any separation of both radionuclides from HLW reduces in a propor- e heath tionat y e residuaoutpuwa th l f o t l waste fractiod an n decreases the Cs contamination problems during vitrification. One of the limiting factors in disposal of vitrified HLW in conjunc- tio a-wasted n an e deca witW th ML hs y ,i heat emittee canisth y b -d ters. Roughl 1 KW/TEQy U afte6 KW/TEQ 0. 0 year 1 rd an sU afte0 5 r years, a 90 o/o removal of both heat producing radionuclides (137Cs and 90Sr) can shorten the standard cooling time in engineered structure 0 year2 so r les t o , fro s0 5 mdependin e th n o g therma e lformation th loa f o d r fractione separateS th ,d an s sC d e a-freb hav e moro t us ee e t (i.eno y .ma n Ci/gd belo an 0 ) 10 w e initiath f o l o equivaleno/ tha 0 1 n t glass volume. Thes- re e quirement e fundamentaar s d r calveran fo ll y - specifiex n io c changers with high absorption capacity. These matrices constitute after drying and conditioning a first choice storage form during 0 years30 . Beyond that perio e radiatioth d n hazard becomes négli- geable.

1. INTRODUCTION Separation of Cs and Sr from High Level Waste (HLW) solution is a subject which was intensively studied some 20 years ago. The use of inorganic ion exchangers as separation method seemed obvious becaus e classicae th non f o e l partition processes e.g. liquid extractio r precipitatioo n n could achieve sufficiently high decontamination factors. The main aim of the studies was the development of inorganic ion exchangers which possessed sufficiently high selectivity coefficients for the two most important fission products 137Cs and 90Sr and a reasonably high uptake capacity.

31 In the frame work of waste treatment it was at that time hoped tha y separatinr b frotS witW HL md h an Decontaminatios C g n Factors (DFf 10o )e 1effluen*th t from this process coul e treateb d s a d Medium Level Waste (MLW) for which adequate conditioning and disposal methods were available. e presencTh f a-emittero t econcentrationa W HL n i s t s no whic e ar h compatible with MLW compositions was at the beginning not recog- a nisever s ya d important fact t graduallybu , , with increasing fuel burn-up t infusei , e scientifith d c community that evee th n most perfect Cs-Sr partition method coul t transforno d W intHL mo MLW. r SEPARATIOS d an . STATU 2 s C BELGIUN I F SO FRO W HL MM AROUND THE MIDDLE SIXTIES 2.1. Separatios C f o n Several inorganic ion exchangers displaying high selectivities for Cs in concentrated HN03 solution were developed. Zirconiu• m phosphate [3) Titaniu• m phosphate • Ferro-cyanide molybdate e firsTh t selective inorgani n exchangeio c r capabl f separo e - ating Cs from simulated HAW solutions (see fig. 1 for composi- tiod generaan n l PUREX flow sheets s zirconiuwa ) m phosphate. The product was characterized as an amorphous or semi-crystalline ion exchanger with mono functional phosphate groups. The acidity of HAW or IWF being about l M HN0 and the chemical compositio e solutioth f o n n rather diluted, zirconium phosphat3 e is technically usable for Cs removal. However the introduction of a Cs partition module between the HAW and IWF or between the e higth hd IWactivitan F y concentrato a rea s li r technological problem which cannot be realized unless the reprocessing process planneha s e partitioth d n module durin e desigth g n phase e plantth f o . Experiments with reaW solutionHA l s have shown that Cs can be removed from the solution by using a zirconium phosphate cartridge but that the adsorption is rather irreversible. Regeneratio e samd repetitivth an n e f o zirconiu e us e m phosphate cartridge seemed to be problematic. About simultaneously a.second generation ion exchange product "titanium phosphate" was developed and showed to be superior to zirconium phosphate. The functional groups acidity of titanium phosphat e reactios higheth - wa e d ex an rn n io kinetic e th f o s change s severawa r l times faster than tha f zirconiuo t m phosphates. Cs separation experiments were conducted in hot cell conditions with High Activity Waste Concentrate (HAWC) solutions produced by the Eurochemic facility in 1966-1967. These solutions typical of PUREX HAWC stream contain 7 M HN03 and a wide variet f chemicao y l impuritie s showa s fign i n 2 .(compos - ition 1). The composition 2 was artificially prepared with a tenfold lower Al-concentration because this t elemenno s wa t considered typical of a PUREX HAWC product but occurred in the Eurochemic waste solution. Volume reduction factors of about 50

32 :UN O.B s HN03 0.20 Pu.s/1 0.04 Flow 353 Sp.Gr. C.9& V.Pu »7.9 V.U ¥ OF. Pu.g/l f Curiej 0 24 s Flow 75 Ns[FP)«6 T*mp 50«C(i»a>) NtiFPJrlD SpGr. to16o 1 O M U . V ;DF. ta10* rCuriîs 3x10*

KjO MNOj 2.4» KÛ.* o.oa Ft*** O.OT2 Sp.Gr. 109 Flow 413

«OH ACTIVTTT U7 CONŒXTRATOR I.OK Sf.Qr. «3 FUv OS

FIG. 1.

with DF's exceeding 100 have been experimentally obtained in t reaconditionsho l e definitioTh .e volumth f o en reduction factor is the ratio between the volume of feed solution and that of the ion exchange cartridge on which selectively all Cs ions e presenfeeth dn i tsolutio n have been concentrated. Repeated use of one cartridge was shown to be feasible and f titaniuo therefor e us me th phosphate s mucwa e h more focussed e separatiooth n 137f o nC s from fission product solutiona s a s means of cheaply producing 137Cs sources for irradiation purposes.

33 COMPOSITION OF HAW SOLUTIONS

Element 9/1 g/t

Fe .1.25 1.25 Cr 0.54 0.54 Ni 0.94 0.94 Mn 0.02 0.02 Mo 1.35 1.35 Cu 0.035 0.035 Zr 5.45 0.0545 Al 22.6 0.226 U 3.35 .3.35 Pu _ _ RE 7.2 7.2 Cs 1.9 1.9 Sr 0.83 0.83 Na 6.8 0.88 SI 0.032 0.032 SO; 1.34 1.34 NH; 0.1U 0.144 pot 1.5 1.5 H* 7. M 7. M

FIG. 2

A third type of ion exchange material is an acid coprecitate of molybdenum and ferro-cyanides. This type of compound is much more crystalline than the M(IV) phosphates and displays the highest selectivity towards Cs in acid solution. This ion exchange material calle r simplicitfo d y "Ferro-cyanide molybdate" is specifically suite r immobilizatiofo d f 137o n Cn viei sf o w final storage. The product is chemically less stable than the M(IV) phosphates and must consequently be used in a continuous mode feed operation. The reaction kinetics is similar to that f titaniuo m phosphat e sorptioth t bu en e conb reactio o -t s i n sidered as irreversible. Tests with real HAWC solution have been carried out till plugging of the cartridge because discon- tinuous mode feed operation d beeha sn imposet cellho n .i d In s conclusiosorptioC e th n f e o stateb nparagrap y dma t i h that according to 1965 state-of-the-art two ion exchange materials behaved quite well in real hot cell conditions : titanium phosphat d ferran e o cyanide molybdat e firsth t; ebein g preferred because of chemical stability.

2.2. Separation of 90Sr from HLW [8) A much more difficult partition process to be developed was the separatio r froS mf o nHAW C solutions. Non f thoso e t thaa e t time known sorbents permitted a relatively selective separation n suci a complehr S f o x mixtur t suca f iono ehd an higs h acidity. The discovery in 1965 of polyantimonic acid as Sr sorbent in acid medium constitute a read l breakthroug n thii h s field. Laboratory tests with simulated HAWC solution showed the sorption e reactiocapacitth e higt b bu ho n t y kinetics very slow.

34 By workin t highe a a greasonnabl ) r°C temperatur0 8 yo t p (u e good adsorption kinetics was obtained. The main ion exchange competition came from the rare earths and from sodium ions if present in solution. By continuous mode feed operation at very w flolo w rates practica n exchangio l e saturatioe b n ca n achieve d voluman d e reductio 0 accordin2 no t factor 5 o 1 t g f o s the accepted breakthrough percentage have been obtained with HAWC solutions. As a conclusion of the previous paragraphs describing the state of-the-art on Cs and Sr partitioning from HLW in the I9601s in e stateb Belgiun dca t thai m t selective separatio n irrevero n - sibly loaded inorgani n exchangio c e material s feasiblwa s e with HAWC solutions representative of typical PUREX-HLW solutions. However a better result can be obtained by installing a Cs-Sr partition module between the HAW tank and the 1 WW evaporator. A schematic flow sheet with typical adsorption datr titafo a - nium phosphate (TiP) and polyantimonic acid (PA) is given in Fig. 3. The block diagramme shows that the 137Cs volume reduction facto s adequati r t thae 90bu Seth t r sorption capacity ought to be increased if the combined Cs + Sr cartridge is to become smaller thae initiath n l waste vitrification volume (120 l/TeqU),[9).

Cs - Sr SEPARATION FROM HAW

HAW lEurochemic) ••» 3600 1 MLLW HNOM 4 j 2- C* S»0 mo/l- 48 1 TIP A P •18I 0 SCi WC$ ÏÏ -} mg/l0 S20 r - L J1 H TIP PA r 5CS i 1 1 Corrosion products L Acfmidts TIP PA £ Rare earths t High activity Cartridge Cartridge concentrator 4000 I/toU q nE disconnection disconnection Drying compaction Drying compaction-

TIPmTitinium phosphate PA»Polyantimonic acid s C y Dr Dry Sr Vitrification storage storage 120 I HLW 44 1 90 1 la" i___ ENGINEERED STORAGE DISPOSAL

FIG. 3 .

3. ROLE OF Cs-Sr SEPARATION ON LIQUID HLW HANDLING AND STORAGE e heaTh t outpu• a functio ts a produce W f tim o nHL shows i ey b dn on Fig. 4. It clearly illustrates the fact that from 5 till 100 years storage the heat problem is dominated by the presence of 137 90d CSrsan . Indeed during this perio e over-alth d l heat output curve practically coincides with the Cs-Sr curve and starts deviating consistently t whica 0 froyear 16 , mh on s momen e heath tt output produce y actinideb d s (21tl 243d Aman A m essentially) equals that of the fission products.

35 5 year Watt 0 yea1 r 25 year SO year 100 year

(470y) (7600y)

10'

Fission products "Sr - "Y (28 j) 137C - s137 0 B(3 a 10

25 100 500 103 104 Cooling time (years)

FIG . 4 . THERMAL OUTPU STANDARA F O T W HL D GLASS BLOCK CORRESPONDINO GT . URANIUEq N TO M 1

In the early storage period, between 5 and 10 years, after discharge froe W reactosolutioth HL m e s th rstorei n n liquii d d for n specialli m y equippe W storagHL d e tanks (see FigTh . .5) safety aspects relate o thit d s typ f storago e e surpassel al y b s means those associated wite entirth h e PUREX processing schemes A cost benefit analysi e Cs-Sth rf o s separation might therefore help to endorse the thesis that even at short notice the simplified constructio W storagHL f o ne tanks might outweige th h construction of the Cs-Sr partition module.

• During the vitrification process liquid HLW solution is heated t temperatureua p f llbO-120o s C causin° 0 a partiag l evaporation of 137Cs (15 to 30 %) which is partly recycled into the me!ter partl e cooth yl n o plate surface e ventit th ou d f -o s lation system. Suppressio 137f o n Cs depositio n colo n d surfaces and reduction of liquids recycling is another advantage of initia s separatioC l n froW solutionHA m s promotea s n thii d s paper. Howeve s difficuli t i r o evaluatt e over-alth e l benefit compared to the HLW vitrification as a stand alone process since alsW treatmenML o t cost se take b hav o nt e into account.

36 SKETCH OF HIGH-LEVEL LtOUID «ASTE STORAGE TAMK AT TOKAI, JAPAN

Steam (in)

Air purge dn> concentratW L A H i e[m

- Stea, ) mim S HAU* concern rate HA L W concent raie* ''- ""' Coding waler

FIG. 5.

4. ROLE OF Cs-Sr PARTITIONING ON LONG TERM STORAGE OF VITRIFIED HLW Solidified HLW resulting from the vitrification process contains all the fission products and as a consequence (explained in the previous paragraph) produces lot f heao s t (see Fig ) whic4 . h have e dissipateb o t d inte geologicath o l formation. The actinides in the HLW produce a much lower heat which does not decrease substantially beyond 500 years. Several scenarios (see Fig. 6) have been considered for geological a-wastd an disposa W HL e f resultino l g yea0 fro3 ra m nuclear electricity production programme r eacFo h. scenario discharge from the reacto e time origi th s supposeth i e r = f e o grapb nt o t t a dh 0, reprocessing with simultaneous production of a-wastes and li- year5 quivitrificatiod = occurW an st HL d years0 t 1 a s = .t t a n

SCENARIO 1 Disposal of a-waste 10 years after t = 0 and disposal of HLW at years0 5 = .t

37 .urne in years 0 5 10 20 30 40 50 60 70 60 enario 1 Nucliar programme ' 10 Disposal of a. waste StoragW HL f o e A 50 Disposal of HLW

Nuclear programme enario -••••1 — — — t * —. ... _ storag e* wast e It Disposal of «waste 30 StoragW HL f o e 0 13 A Disposal of HLW enario Nucltar preyMMe . 1 • «.,,.- storage «waste • , , , u « "Disposa _«wastf .o l e 50 * StoragW HL f o e 0 15 A DisposaW HL f o l enario Nuclear programme W . *••"•, • "•"•"•" Storage «waste Disposa wast< f o le s|orjgeofHLW| 10 r . . U 30 DisposaW HL f o l STORAGE SCENARIO' S OF REPROCESSING WASTE FIG. 6 .

SCENARIO 2 Simultaneous disposal of HLW and a-wastes at t = 30 years after discharge.

SCENARIO 3 Simultaneou 0 years5 a-wastd s= an disposat . W t a eHL f o l

SCENARI4 O Disposal of a-waste at t = 10 years and disposal of HLW at t = 30 years. e limitinTh g l scenario'factoal n i r s withoui s y douban e t th t permissible heat load of the geological formation which limits the number of canisters per unit volume or surface to be disposed off a give t a n time.

The earliest disposal scenario for HLW is scheduled at t = 30 year stretched an s t til 0 yearsou 5 l s afte e firsth r t fuel element discharge from the reactor. The a-wastes can be disposed off in a

38 suitable geologic formation as soon as the repository is available. Economic considerations have shown that simultaneous dis- W posawaste HL f alph o s lpreferabld i san a o consecutivt e e disposal of a-wastes and HLW wastes. The economic benefits of simultaneous disposal can be deduced from the operation cost of a repository beyond the period necessary for a-wastes. In scenario 1 the time span considered amounts to 40 years whereas in scenario 4 (a formation with high thermal load) the interval between the end of the a-waste disposal period and the end of HLW disposal amounts to 0 years2 . Every yea a rrepositore kep b operation i to t s - ha yre n quires the services of about 100 skilled technicians. By integrat- ing these services ove 0 respectivel2 r 0 year4 y a stota l expend- iture benefit of 2000 resp. 4000 man year is obtained. By dis- posing a-wastes together with Cs-Sr from HLW another over-all gain in disposal expenditures is expected from the size of the repository. Presently the repository site is mainly determined by the spacing of heat sources rather than by the total volume of a-waste though 0 time1 thes e sar emor e importan n volumi t e than HLW y pursuin.B g the Cs-Sr from HLW option the small HLW volumes have merely to be added to the a-waste volume which increases then with 10 % and the design differences between HLW and a-waste galleries disappear in the safety projecte margith f o n . In orde o illustratt r e influenc- th ee heaup th e tf th o loae n o d take capacity of a repository for HLW canisters a typical example connected to the currently investigated Boom-clay layer situated t 180-28a 0m belo e nucleath w r sit f Molo e , Belgiu s discussei m d in some detail, [10]. The following constraints were e imposedesigth n no d criteri: a the temperature at the contact between HLW canister and the clay o avoit formatio C t ° dexcee no dryin0 y 12 d dloca ma nan g l col- lapse, the temperature increase at the interface between clay and sand n ordei y t o exceeavoiC t no rlayer° y 5 dd ma sconvectiv e current e sandth n yi s aquife d finallan r e grounth y d surface temperature increase may not exceed 0.5 °C.

.Taking into account these very conservative hypotheses and relying n experimentaoa n l heat transfer coefficien n cla i tf 1.6 o y m 9W/ °C it was calculated that these design conditions lead to a maxi- = 10.00 A (h 0 muA mh 2)r therma.pe W k l 2 2 loa f o d After 30 years each HLW canister (corresponding to 1.5 TeqU) still ha a heas twhicW k outpu l h f correspondo t 2 canister2 A o h t s r pe s or the equivalent yearly HLW production of 1 GWel cooled during 30 years. By waiting 50 years the heat output per canister decreases to 0.6 r eacko W h hectar 6 canisters3 s roor ha efo m . The removal of the heat producing fission products has the same effect as a cooling time. By removing 5 years after discharge from 137f o 90C r o % Ssr 0 correCs-Se 10 th th r e o r -f o reacto % 0 5 r spond coolina o t s f abouo g 0 yearst2 .

39 Cs - Sr SEPARATION FROM FISSION PRODUCTS REQUIRED DECONTAMINATION FACTORS HLW 50 YEARS COOLING 390 W.

DF EQUIVALENT COOLING TIME reduced to : years

100 7.5 95 90 8.5 80 9.9 50 21. 0 50 DECONTAMINATION FACTO «EMITTERF O R S n Ci/ BASI0 n exchangeg10 Io : S v - r Vol. red. fact. : 50 / —— «Activity rCi 674 U3 \ decayear 0 year0 25 y50 s s

FIG. 7.

Fig 7 show. s systematicall e influenc th ye remova th f o el percen- e tagequivalen f th Cs-So e n o r t cooling time t clearlI . y shows that 80 % removal of both heat producing isotopes is sufficient to produce waste canisters which can be disposed off into a geological formation immediately after their vitrification, i.e. at t = 10 years. The existing separation technology developed earlier is capable of providin e necessarth g e removayth basr f Cs-So lfo e r froW HA m streams. Eighty percent removal correspond n "ioi s n exchange" 5 whic = s ver i hF D y terma eas o o obtait t ys n industriai n l circumstances. The question arises what to do with the separated Cs and Sr frac- o aspecttionTw ? sse addresse b have e o volumth t e th d : dan e e separatepuritth f o y d fractions. As already shown in Fig. 3 the separation technique of Cs is sufficiently developed to put it into operation since the volume of the separate s cartridgC de vitrifie th f thao f o t% s onl i e3 d 3 y 1 compared'twast 0 (4 e1 glass) 0 12 o A .decontaminatio n factof o r % removal 5 (9 n easil 0 )ca 2 e guaranteeb y n piloi d t scale facil- itie d thian s s separation alone reduce e coolinth s e g th tim f o e residual vitrified waste to about 24 years. e performeb o t s n orde i ha dFurther separ S o brinD t r e & -th gR r ation technique at the same chemical and technological level as that of Cs. e seconTh d aspect which needs .some further commen e puritth s i ty r e fractionseparateS th e separatior th o f o f s I C .d n technique o offe t w perspective s ne r ha e fielth f geologicao n di s l disposal it must provid r fractionS d e an pur s sC e fre f long-liveo e - o d emitters.

40 s showA Fign e decontaminatioi n th . 7 n factor f a-emittero s s must be very high in order to get a cartridge whose residual a~activity is equa r loweo l r tha0 nCi/ 10 n exchanger io g . This a-activits i y now considered necessary to keep the separated fractions below the limit of a-waste which must otherwise be disposed off in geological formation. The second most important R & D task is the development of ion exchange techniques capable to cope with these severe restrictions on the a-purity of the separated Cs-Sr isoto- pic fraction. e Cs-SIth f r separation techniqu e poin e improveth b o tn t ca ep u d wher e tota r fractioth eS l+ volums C n f doeo et exceeo t no s 0 2 d 30 % of the regular glass volume and if these fractions are free of a-emitters their storage in bunkers during 300 years is suffi- cien o reduct t e their healt w levehlo hazarla wasto t d e type which may be dispersed in the biosphere. In these circumstance e nucleath s r electricity productios ha n disposed off safely all the a-emitters in deep geological formation and the short-lived fission products can be stored safely during a period which is short in comparison with the lifetime of a civilisation.

REFERENCES

[1] BAETSLE, L.H., PELSMAEKERS, J. J. Inorg. Nucl. Chem. (21) pp. 124-132 (1961-). [2) BAETSLE, L.H., HUYS, D., J. Inorg. Nucl. Chem. (21) pp. 131- 140 (1961). [3] BAETSLE, L.H., J. Inorg. Nucl. Chem. (25) pp. 271-282 (1963). [4] PIRET, J., HENRY, J., BALOU, G., BEAUDET, C., Bull. Soc. Chim. Fr. , p. 3590, (1965). [5) BAETSLE, L.H., MARKL MUSYL, L., HUYS, D., "Ion Exchange in the Process Industries", Soc. Chem. Ind. pp. 209-215 (1970). [6) BAETSLE, L.H., VAN DEYCK, D., HUYS, D., J. Inorg. Nucl. Chem. (27683-69. pp ) 5 (1965). [7) BAETSLE, L.H., HUYS, D., VAN DEYCK D., J. Inorg. Nucl. Chem. (28) pp. 2385-2394 (1966). [8) BAETSLE, L.H., HUYS, D., J. Inorg. Nucl. Chem. (30) pp. 636- 649 (1968). ) BAETSLE[9 , L.H., HUYS , SPEECKAERT,D. , Ph., Belgian Report from S.C.K./C.E.N., BL6 487 (1973). [10] COMMISSION OF THE EUROPEAN COMMUNITIES, "Nuclear Science and Technology : Admissible Thermal Loading in Geological Forma- tions. Consequences on Radio-active Waste Disposal Methods", Vol. 1 Synthesis Report. Report EUR 8179 (1982).

41 EVALUATIO ZEOLITF NO E MIXTURES FOR DECONTAMINATING HIGH-ACTIVITY-LEVEL WATER AT THE THREE MILE ISLAND UNIT 2 NUCLEAR POWER STATION

E.D. COLLINS, D.O. CAMPBELL, L.J. KING, J.B. KNAUER Oak Ridge National Laboratory,* Oak Ridge, Tennessee R.M.WALLACE Savannah River Laboratory,** Aiken, South Carolina United State f Americso a

Abstract Mixtures of Linde™ lonsiv™ IE-96 and lonsiv™ A-51 zeolites e Submergewerth en i evaluate e dus r fo d Demineralizer System (SDS) thas installewa te th t a d Three Mile Island Uni Nuclea2 t r Power Statio decono t n - taminate ~2780 m3 of high-activity-level water (HALW). e originaTh S flowsheeSD l s conservativelwa t y designed r removafo f cesiuo l d strontiuan m d woulan m d have S columnsSD 0 -6 . f o Mixe requiree us d e zeolitth d e tests, which were made on a 10~5 scale, indicated that the appropriate rati f IE-96/A-5o o s 3:2wa 1 .A mathemat - ical mode s use wa o prediclt d e performancth t e th f o e mixed zeolite columns in the SDS configuration and with the intended metho f operationo d . Actual loading results were similar to those predicted for strontium and better than those predicted for cesium. The number S columnoSD f s neede o procest d e HAL s reduceth sWwa o t d -10.

1. INTRODUCTION

Large volumes of high-activity-level water (HALW), containing >100 Ci/m , were createe accidenth y b dt Thre a t e Mile Island, 3

Uni (TMI-2)2 t . This HALW consiste bodieo tw e f th s— thao d n i t reactor coolant system (-34thad an t 3 0) m whic s spillehwa d into the reactor building floor (-2440 m3). Early selectioa f o n flowsheet for decontamination of the HALW was necessary so that process design and equipment fabrication could be initiated. After careful consideration, the TMI-2 Technical Advisory Group selected a flowsheet that was based primarily on adsorption of the bulk of the radioactive contaminants, cesium and strontium, onto an inorganic ion exchanger. This exchanger, Linde™ lonsiv™ IE-95, is a commercially available chabazite type of zeolite, having a histor f successfulo y , large-scale usage.

* Operated by the Martin Marietta Energy Systems, Inc., under contract DE-AC05- 840R2 1400 with the U.S. Department of Energy. ** Operated by E.I. Du Pont de Nemours & Co. under contract DE-AC09-76SR00001 with the U.S. Department of Energy.

43 The processing system was designed by Allied Chemical Nuclear Servicee Cheth m r Nucleafo s r Company e primth , e contractor fo r equipment fabricatio d installationan n e systeTh .s designe wa m d so that the equipment items containing high levels of activity e spenth wer f to e fue e houseon l n handlini d g pool n ordei s o t r use the pool water for shielding. Therefore, the process was calle e "Submergeth d d Demineralizer System, r SDSo " , althouge th h process was not intended to demineralize the water during its decontamination.

S procesSD e sTh flowshee s evaluatewa t a serie n i df test o s s k RiagOa mad Çt a eNationa l Laboratory during early 1980, usinL 3 g of TMI-2 HALW.IÏ e testTh Js showe de cesiu thath e bul f th to mk and strontiu s effectivelwa m y adsorbe e zeoliteth n o t dtha bu , t subsequent treatment with organic-based polishing t resinno d di s provide additional decontamination from cesiu d strontiuan m r o m removaminoe th rf o lcontaminants , 125 106d Sban Ru , whice ar h anionic in nature.

The original SDS flowsheet called for the contaminated water e clarifietb o d thean dn passed throug n exchanga serieh io f o s e columns s illustratea , n Figi d . 1 Fou r small columns, each containing -23f sorbento 0L ,e locate b wer o t e d withi e spenth n t fuel pool. The first three were to contain IE-96 zeolite, the fourth a strong-acid type cation exchange resin.

SPENT FUEL POOL

CLARIFIED * m d mL1- u.~C ZEOLITE COLUMNS CATION POLISHIN COLUMN COLUM!> ILHUU_ Ü •lÉt?*aj. aJ.H _ 1

FIG. I ORIGINAL SDS FLOWSHEET.

e IE-9Th 6 zeolitn excellena e s founb wa eo t dt sorbenr fo t cesium and adequate for strontium, if the contact time between the wate d zeolitan r s sufficientlwa e y long (>20 min). Thereforee th , manne f operatioo r e proposes designeth wa f S o n o accomSD dt d - modate the needed contact time.

e columnTh s were modula d weran re intendee th e use b s a do t d radioactive waste containers after being loaded e flowsheeTh . t e column callee moveth b r o dt fo sd after each 45,00 watef o 0L r had been processed. This is equivalent to 200 bed volumes, based on each column t thaA . t point e e firscolumth th , tn i npositio n e dischargedb o t s e wa othe th o move, tw r d forwar e positionon d , countercurrent to the flow of water, and a new column installed in the third position. In this manner, the cesium would be loaded in the column in the first position and all three columns would provide a sufficient contact time for strontium sorption.

44 2. EXPERIMENTAL PROCEDURES

2.1. Consideratio r increasefo n d loadinf o g cesiu d strontiuan m n zeolito m e

The original flowsheet was conservatively designed with respec o cesiut t d strontiuan m m remova d woulan l d have requiret a d leas 0 column 6 HALWe t th proceso .f t so Thusl al s U.Sa , . Depart- men f Energo t y Task Force l J evaluate e technicath d d financiaan l l 2 benefits of loading the columns with cesium to a higher level than that specified by the original*flowsheet. This higher level was limited by the loading characteristics of strontium.

During evaluatio originae th f S flowsheeto n SD l ,a 1000-be d volume tesd beeha t n made e resultTh . s indicated tha s mana ts a y ~600 bed volumes could be processed before strontium breakthrough froe thirth m d column would occur. Sinc e organith e c cation resin, which was to be used originally in the fourth column, was found to be ineffective for providing additional decontamination, consideration was given to using zeolite in the fourth column in orde o increast r e capacit e systemth e th f o y. Then e throughth , - put could be increased to 600 bed volumes, while maintaining a sufficient safety margin. This would reduc S e numbeSD th e f o r columns needed to ~25.

2.2. Consideration and evaluation of the use of mixed zeolites

A further increas n loadini e s envisionewa g d afte mora r e "strontium-specific" zeolite, Linde" lonsiv"1 A-51, was identi- f IE-9o e fied6us zeolit e . Th l-*r cesiuJ fo e m sorptio d A-5an n 1 zeolit r strontiufo e m sorptio n eithei n r mixed beds, layered beds, or alternate column s f alternatconsideredo wa s e us e e Th .column s would involve maintaining two types of columns.that might require different treatment during subsequent waste solidification opera- tions. If layered columns were used, and if flow distribution at the bottom of the column was poor, the bottom layer could be adversely affected. f mixeThereforeo e d us columne th , s appeared to be most appropriate.

A 1500-bed volume test was made with TMI-2 HALW and a mixed zeolit d containinbe e g equal part f IE-9o sd A-51 6an e break.Th - through curves obtained for cesium and strontium, using this mixture, were compared with those obtained when using only IE-96 zeolite (Fig. 2). The cesium breakthrough, when using only IE-96, s <0.01wa % durin e entirth g e 1000-bed volume throughpu f thao t t test.

These results show that, when usin e mixeth g d zeolitee th , capacity for strontium sorption is increased by a factor of ~10, even thoug kinetice e strontiuth h th f o s m sorption s indicate,a d by the lesser slope of the breakthrough curve, are somewhat slower e capacitTh . r cesiufo y s adequati m a throughpu r fo e f o t ~200o ut pd volumes 0be . That loading would represen factoa t r increas0 o1 f e ovee originath r l flowsheet design.

45 0 10060 0 40 2000 0 20 5000 10.000 COLUMN THROUGHPUT (BED VOLUMES) FIG. 2. COMPARISON OF PERFORMANCE OF SINGLE ZEOLITE AND MIXED ZEOLITE.

5000

STRONTIUM

4000

3000 CESIUM z

? 3000 CT t- O Sr, TRACER-LEVEL TEST • Sr, HALW TEST + S SrSO , 1000 o Cl , TRACER-LEVEL TEST HAL, Cs W TES• T + C», SOS I I __ I I

0 8 0 6 0 4 0 2 100 % IE-96 IN ZEOLITE MIXTURE

FIG. 3. CESIUM AND STRONTIUM DISTRIBUTION COEFFICIENTS

OBTAINED WITH VARIOUS ZEOLITE MIXTURES.

The proportionzeoliteo tw e th s f necessaro s o achievt y e balanced loadings of cesium and strontium were determined by meana serie f o f stracer-leveo s l tests synthetiA . c solution formulate a compositio o t d n similae TMI-th o 2t r HALW (200m pp 0 of boron, 1600 ppm of sodium, and minor amounts of other elements, 137 with Cs and o9-90sr added as radioactive tracers) was used with

46 3/1 RATIO IE96/A51

to4 io2 to3 to4 to2 to3 to" COLUMN THROUGHPUT (BED VOLUMES)

FIG. 4. RESULTS OF TRACER-LEVEL COLUMN TESTS.

the volume of TMI-2 HALW available was sufficient for only one test. That tess madwa t e zeolit1 wit1: a h e mixture s describe,a d e tracer-leveth f o above e On .l test s alswa s o made1 wit1: a h zeolite mixture and produced results which were not substantially different from those obtained wite actuath h l HALW. Other tracer- level tests were made with 2:1 and 3:1 ratios of IE-96/A-51. A correlation of distribution coefficients obtained for cesium and strontium with different zeolite mixtures is shown in Fig. 3, while cesiu d strontiuan m m breakthrough dat e showa com ar an o n- parativ e serie th basi f testr o s fo sn Fig i s . 4 .Thes e data indicate that the proper ratio for balanced loading is between 1:1 and 2:1. Thus mixtur,a eA-5f o containin % 1 40 f IE-9o d % 6an 60 g (a ratio of 3:2) was selected.

2.3. Modeling e testTh s provided breakthrough dat r onle columnfo aon y . Therefore, the data were fitted by means of the J-function, using e constanth t separation factor model develope y Thomas,l^b d o t J enable calculatio e masth s f o ntransfe r coefficient d extrapoan s - lation of the data to obtain the estimated performance of a second, third, and fourth column in series.

e generaTh l Thomas equatioe reactioth r fo nn kineticn io f o s exchange in a fixed bed is as follows:

3X 3Y •( X(l - Y) - RY(1 - X) (1)

where X and Y are the dimensionless concentrations of the solute e flui d th soli an dn i d n phasesio , respectivelye th s i R d an ,

47 separation factor e variablTh .s definei X e s C/CQa d , d wheran C e CQ are the concentrations of the solute ion of interest in the effluent and feed solutions, respectively. The variable Y is defined as q/q , where q is the actual concentration in the solid phase, and q is the concentration in the solid phase when it is in equilibrium with fluid at the inlet concentration, CQ. When the concentratio s smale soluti th n lf io eo nrelativ e conth -o t e centration of the replaceable ion in the feed (as it is in this case) ,R approache s e isotherunityth d s linearan ,i m .

The variable N represents the length of the exchange column n transfei r s definee i expressiounit th d y an sb d n

N = K; K /(f/v) , (2) abP a in which K^ is the distribution coefficient when X = 1, pg is the bulk densition-exchangere th f mass-transfee o y th s i & ,K r coefficient characteristi e rate systemth f floth o e f s o f i wo c f , solution through the column, and v denotes the overall volume of the sorbent bed, including the void spaces. The throughput parameter, T, is defined approximately by:

T = (V/v)/KjpB , (3)

e volumth wherf solutios o ei V e n processed throug e columnth h . Note that V/v is the number of "bed volumes" of solution.

Since pg is essentially constant, it is convenient to define a volume-basis distribution coefficient q= v /C0^ ,K , s wheri y q e the concentration of the solute ion per unit volume of the sorbent bed (sorbent plus void space) and C0 is the concentration in the feed solution n the ca e expresseb n . ) (3 Equation d s a dan ) (2 s

N = K.K /(f/v) , (2a) Q 3

and

T = (V/v)/Kd . (3a)

Equation (1) has been integrated [Eq.(16-128a) in réf. 5] for the special case of reversible second-order reaction kinetics (appropriate to ion exchange) with the solution being

______J(RN.NT) J(RN,NT) + [l-J(N,RNT)j exp [(R-l

where J is a mathematical function* related to the Bessel function, IQ.

48 For large values of RN (a condition approached in SDS opera- = tion and in the small-scale tests), C/CQ "-0.5 when T = 1, independent of the values of RN. This characteristic can be employe e datth a n i analysisd . Experimenta e uselb o datt dn ca a construct logarithmic-probability plot f C/Co s s V/vv 0 . These plots are nearly linear and can be used to estimate Kjj, which is approximatel e pointh t a whery v equaV/ e o t lC/C Q= 0.5 . Values of R and N can then be obtained from the experimental data through the iterativ Eqsf o e . us e(3a d (4))an .

A numerical solution model of the Thomas equation was developed to accept input data in a form that simulates the cyclical mode of operation proposed for the SDS. A numerical solution was required to analyze the multibed system in which the partially loaded column e movear s d forward (countercurren e wateth ro t t flow) periodically, because an analytic solution is not practical unless the initial loading on each bed is zero.

e fou Th S werrSD e column e representeth f o se mode th y b ln i d two 4000-point array, usinY) r gs fo 100(on d an 0e X eacr fo h point r eacfo sh column. Calculations were carrieo simulatt t ou d e the passage of the desired volume of feed through the four columns in series, with initial values of zero for X(n) and Y(n) for all points.

At the end of the first feed cycle, the values of X(n) and Y(n) were replaced by the previously calculated values of X(n + 1000) and Y(n + 1000) for values of n between 1 and 3000 and were set equal to zero for values of n between 3001 and 4000. This procedure simulated removing the first colunn, moving the last three columns forward one position, and putting a new column in the fourth position. The calculations were then repeated for another cycle, using this configuration as the initial condition. This modeled rotation procedure was repeated for the number of cycles necessary to process the total volume of HALW.

2.4. Predicted and actual SDS performance By interpolating between the experimentally derived distribution coefficient e calculateth d an s d mass transfer coefficients and separatio 1 zeolit2: d an en 1 mixturesfactor1: r fo s , values wer2 mixture3: e e deriveth .r fo Thesd e value e show ar sn Tabl i n e , alonI g with corrected values obtained from earl S operationsSD y . e observeTh d differences wert greatlno e y significant, even though the scaleup factor froe testh mt columS columSD ne n th siz o t e sizs ~10wa e . Wherea e testh st column data indicated nearly balance5 d loading of cesium and strontium, the actual data showed similar performance for strontium but better performance for cesium. As shown in Table II, using the early SDS data, cesium and strontium breakthroughs were calculated for six loading cycles in which 2760 m^ of HALW are processed (460 m^ in each cycle). Although the strontium breakthrough continued to increase throughou x cyclessi e e breakthrougth tth , h froe fourtth m h column did not exceed the concentration (0.1%) of a nonexchangeable specie f strontiuo s m observe e HALW e numbeth Th .n i f dzeolito r e columns use o proces t de HAL th t TMI-e bula W f th s sreduceo k 2wa d to -10.

49 Table I. Comparison of predicted and actual performance parameters for cesium and strontium

Test column data S columSD n data Parameter Cesium Strontium Cesium Strontium

Distribution coefficient (Kd) 2805 3760 3800 3000

3 3 Mass transfer coefficient (Kg) 8.0 x 10"*+ 2.9 x 10~^ 2.2 x 10~ 5.6 x 10~ Separation facto ) (R r 1.10 51. 1.6 50 1.

Table II. Calculated cumulative breakthrough of cesium and strontium for six cycles of column replacement

Cumulative breakthrougf o h % ( feed) Cesium Strontium Cycle Column Column Column Column Column Column Column Column number 1 2 3 4 1 2 3 4

1 0.62 a a a 13.8 0.18 a a

2 0.71 a a a 22.9 0.92 a a

3 0.73 a a a 29.5 2.14 a a

4 0.73 a a a 34.5 3.62 0.17 a

5 0.73 a a a 38.3 5.21 0.35 a

6 0.73 a a a 41.4 6.81 0.60 a

Calculated breakthrough less than observed concentrations (0.003% of cesium and 0.1% of strontium) of nonexchangeable species.

3. SUMMARY

The original SDS flowsheet at TMI-2 was conservatively designed for the removal of cesium and strontium and would have required the use of -60 SDS columns to decontaminate the HALW. As a process improvement f mixeo e d us zeolitee th , o . enablt s - in e creased, balanced loadings of cesium and strontium was evaluated. Tests made on a 10~5 scale indicated that the appropriate ratio of Linde™ lonsiv" IE-96 and lonsiv™ A-51 was 3:2. A mathematical model was used to predict the performance of the mixed zeolite columns in the SDS configurations and with the intended method of operation. Actual loading results were similar to those predicted for strontiu d bettean m r than those predicte r cesiumfo d e Th . number of SDS columns needed to process the HALW was then reduced to -10.

50 4. ACKNOWLEDGMENTS

This work was performed at the Oak Ridge National Laboratory and the Savannah River Laboratory under contracts DE-AC05-840R21400 and DE-AC09-76SR00001, respectively, with the U.S. Department of Energy.

REFERENCES

[1] CAMPBELL, D. 0., et al., "Evaluation of the Submerged Demineralizer System (SDS) Flowsheet for Decontamination of High-Activity Level Water at Three Mile Island Unit 2 Nuclear Power Station," ORNL/TM-7448 (1980).

] Evaluatio[2 f Increaseo n d Cesium Loadin n Submergeo g d Demineralizer Systems (SDS) Zeolite Beds, DOE/SDS Task Force, DOE/NE-0012 (1981). al. t e ] COLLINS [3 , D. "Wate . ,E r Decontamination Process Improvement Test d Considerations,an s " AIChE Symposium Series (19828 213_7 , . )9

HIESTER] [4 , VERMEULENK. . N , , KLEINT. , Chemica, ,G. l Engineers' Handbook, Fourth ed. (PERRY, J. H., et al., Eds) McGraw-Hill, New York (1963) 16-2.

[5] HIESTER, N. K., VERMEULEN, T., Chem. Eng. Prog., 48 (1952) 505.

51 APPLICATION OF SILVER-FREE ZEOLITES TO REMOVE IODINE FROM DISSOLVER OFF-GASES IN SPENT FUEL REPROCESSING PLANTS

T. SAKURAI, Y. KOMAKI, A. TAKAHASHI, M. IZUMO Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken, Japan

Abstract The present work dealt with the interaction of zeolite 13X with such organic iodides as CH3I and C2H5I in the presence of N02 with the intention of applying silver-free zeolites to the iodine removal process of the dissolver off- gas. On the assumption that the dissolver off-gas would be dried for the purpose of removal of tritium, the experiments were carried out under a dry condition.

Zeolit X remove13 e s dilute CH3I fro procesa m s streaga s m with high efficiencies at =25 °C. It was found that N02 decomposed CH3I on the zeolite. Infrared analysis and mass spectrometry confirmed thafollowine th t g reaction took place a silver-exchanged alsn zeolit o an n o X 13 e d zeolitt a e =25 °C.

CHN02 . 32I+ NO ———> + 2 •I CH2 31/ N0 + 3

A similar reaction occurs zeolitese alsr Cth 2tÎ5fo on o l :

C2H5I + 2 N02———)• C2H5N03 + 1/2 I2 + NO. Whe a simulaten d dissolver off-gas (containing =10m 0pp -I2, =10 ppm-CH3I and =1 %-N02) was supplied to a zeolite 13X bed, the I2 which was produced by the decomposition of CH3I was adsorbed at the same place of the bed as another I2 which s originallwa y include gase th .n i d

Atrapd zeolite t be th =10 sX , 0onl13 e°C y 2 iodinN0 d an e passes through the bed without being adsorbed: the zeolite can separate iodine chromatographically from N02. On the basis of these experimental results iodinn ,a e removal process swa discussed.

I. Introduction

Caustic scrubbin s probablgwa firse th y t iodine removal proces s beesha nused adoptean d nucleae th mosn i df o tr reprocessing plants in the world. Because of its relatively poot trappinne r g efficiency (especiall organir fo y c iodides), caustic scrubbers requir secondara e y iodine trapping process if high efficiencie e neededsar . Recently, several new iodine removal processes with higher efficiencies have been developed in many countries, as an alternativ caustie th o t ec scrubbing technique [1]. Among these alternative processes, a dry process which uses silver- exchanged zeolites or silver-impregnated adsorbents has been considered to be promising by many workers. Both elemental

53 iodine (I organid 2)an c iodides form stable complexer o s compounds witsilvee th h n thesi r e adsorbents [1~3]e Th . disadvantage is in that silver is valuable and expensive; a successful recovery procesadsorbene t th bee no f o nss tha develope t [I],ye d

On the other hand, silver-free zeolites, e.g. zeolite 13X, are placed in a low position in the field of the off-gas treatment e reasoTh . n e thaseemb to t ssilver-fre e zeolites lose their iodine removal efficiencies when co-adsorptiof o n water occurs. Wilhelm has reported that there was no useful removal efficienc methyr fo y l iodide (CH3I) when zeolitX 13 e s allowebewa d reaco t d h water adsorption equilibrium wite th h normal laborator r (relativai y e humidit highet ya d 30^6ran ) 0% relative humidity ) [3]% .5 (8

In order to remove tritium, however, there is a possibility e neath r n futuri e tha f s a silictfilteinstalleo i d l be arge d off-gae th n i s treatment dissolvee systeth f o m r off-gas (DOG). When this s realizedfiltei d rbe watee ,th r concentratioe th f no proces s enteringa s iodine th g e removal process decreaseo t s considerably low levels. This situation makes us reconsider the use of less expensive silver-free zeolites for the iodine removal proces DOGf o s .

On this assumption e hav,w e been studying silver-free zeolite n ordei s probo t r e their potential abilitie f adsorbo s - bees ha nint I alreadg presence iodin. th O yN n i ef confirm o e - ed that some kind f silver-freo s e zeolites adsorb elemental iodine with high efficiencies [4], Our next concern is in behavio f organio r c iodidee zeolitesth n o s , which would totae th lf constitut o iodin % 0 1 e- quantit1 e DOGn i y o .N experimental data are available for the interaction of organic iodide with silver-free zeolites in the presence of NO . The main specieorganie th f so c iodid s beeeha n postulatee b o t d methyl iodide (CH3I) [5]. e presenTh t wor s carrieobtaio t kwa e informatiot nth ou d n about the interactions of organic iodides with silver-free zeolites in the presence of NO ; zeolite 13X was used as the sample. A dynamic method was applied for most part of the present work with the aid of a tracer technique using radioiodine 131 IA stati. c metho s adoptewa d s occasioa d n demanded.

II. Experimental

1. Materials

The zeolites used were: (i) zeolite 13X in the form of 1/16" pellets d (ii,an )silver-exchangea e d th zeolitn i X eAg form of 1 mm4> particles for reference.

Commercial grade methyl iodide (CH3I), ethyl iodide (C2H5I), and nitrogen dioxide (N0) were used without further purifica-

tion. 2

Radioiodine, 131I, was purchased from Japan Radioisotope Association in the form of a carrier-free Na131I solution.

54 2. Preparation radioactivf o s e methyl iodide d (CHan 3I) elemental iodine (I*)

The methyl iodide was prepared through the liquid-phase reaction between dimethyl sulfate ((0113)2201,) and sodium iodide (Nal) labeled with Na131I [6] producte .Th s driea wa n o d zealite 3A bed at 100 °C.

131 Heating a mixture of K2Cr207 and Nal labeled with Na I to 400 °C yielded radioactive elemental iodine [6], which was also dried on zeolite 3A.

Their specific activities wer mCi/g-CH2 e= =30d 3an I0 yCi/g-I2. . Apparatu3 d procedur an e dynamis th r cfo e method

The dissolver off-gas entering the iodine removal process of the reprocessing plants, has been postulated to include =1 % of NO and 50^200 ppm of iodine. Organic iodides constitute 1 - 10 % of the total iodine quantity.

In the first stage of the experiments, the removal efficienc CHr zeolitf s studieo yfo 3 Iwa X y supplyin13 eb d r ai n a g n a strea o t m) % containin1 (= 2 ppm0 N0 10 g d ~ ) CHan 0 3(1 I adsorption column (15 mm i.d. x 200 mm length) packed with ^25 g of zeolite 13X. In the second stage, a simulated DOG was fed inte adsorptioth o n colum n ordeni predico t r behavioe th t f o r organic iodidedissolvee th n i s r off-gas. Figur illustrate1 e s the apparatu r feedinfo s a simulateg d DOG, together with experimental conditions. (In the case of the first stage experiments Ie 2-feedinth , g lins closed.ewa maie r Th )nai strea s driemwa d wit a zeolith bedA 3 e . Prio useo t re th , zeolite bed was heated at 400 °C for l h in an air flow.

cc/mi3 - 19 Ai— r» n Air rnSH 36.3 cc/min

I2,30*C 1 (V.P=0.5 lorr) TEDA filter

CH3I*-78*C (VP=Q7torr) I

Zeolit cohX e13 m KI-I2 impregnated charcoal N02,0*C

[I10= 2]0 ppm

CCH3I]=: 10 ppm [NOj] A 1 % Fig 1 Apparatu. r feedinfo s e simulategth d dissolver off-gas. (V.P. =Vapor pressure)

55 Nitrogen dioxide representee presenth n i t workO N d . This woul t affecno d t validit e resultth f o ys obtained, because N02 is a main component of NO and also another component NO is partly oxidized to N02 by oxygen and partly passed through the zeolite bed without being adsorbed.

The mass flowmeters 'Model SEC-LU1 from Standard Technology Co. measured the flow rates of the air flows. Since o experimentan l dat d beeha a n availabl vapoe th r fo pressure e of CH3I at -78 °C, we checked it with a pressure gauge 'MKS Baratron pressure measurement system' from MKS Instruments INC. and obtaine valua d e on e e ) clos(0.th Pa 7o 3 t e 9 tor r o r extrapolated from Antoine's formula [7] e triethylenediamin„Th e (TEDA )KI-Ie filteth 2d -impregnatean r d charcoal captured iodine released from the adsorption column.

s supplga Afte f prescribee o yth r d time, distributiof no 131s examineIwa d alon adsorptioe lengte th th g f o h n column and the filter. In measuring radioactivity, a y~scintillation counter (Nal s shielde)wa d m witthic m blockb 0 P o h5 ks f so as to count it only through a slit of 10 mm wide in the front Pb block.

4. Apparatus and procedure for the static method

The interactions between CH3I and zeolites were also studie meany a statib d f o s c method e experimentaTh . l apparatus consiste CHa 3f Io d container (Monel) 2 containeN0 ,a r (Pyrex glass) Mone,a l reacto loadee b o t rd with zeolite sample4 (2 s m depth)mm m 0 i.d 15 Mone,a x . l cylinde controo t r e th l amounts of the gaseous reactants, a pressure gauge (Monel) and a vacuum manifold.

reactoe th n d i an r t pu e s zeolitth wa f X o e13 g m Abou 0 50 t higa n hi vacuumh l e reacto r Th .fo s C rwa ° 0 drie40 t a d closed and cooled with liq-N2. On the other hand, CH3I and N02 were introduced intcylindee th o r unti prescribee th l d pressure os reachef wa eacs ga hd therein e gaseouTh . s mixtur s theewa n transferre reactoe s condenseth wa s entranco d t it dan r t a d e end. The cold trap was removed from the reactor and the system was kept at room temperature (=25 °C) for =1 h, in order to allo zeolite th w o adsort e gase th b . Subsequently e reacto,th r s detachewa d froapparatue th m r analysi e reactiofo s th f o s n products. The infrared analysis and mass spectrometry of the products were carried out.

III. Result discussiod an s n

1. Removal efficienc CHr zeolitf o yfo 3 I X e13

The interaction of zeolite 13X with CH3I was studied by the dynamic method in the presence of N02. An air flow including =100 ppm-CH intd adsorptio e %-N01 fe = th o s d 3I2wa *an n column (25 CC) at a linear gas velocity of 2.3 cm/sec. In^order to realize a CH3I* concentration of =100 ppm, the CH3I container (Fig. 1) was cooled to -50 °C and an air flow was passed through it at a rate of 2.6 cc/min. Distribution of 131I in the column was measured at every 5 h intervals during the gas supply. Figure 2 shows the results. The abscissa of the figure denotes the

56 distanc ee ordinat e colum froth e inleth th d me amounf an o ntth e t of CH3I adsorbed at each position along the length of the column.

Methyl iodid s adsorbeewa d e colum neainlee th th rf no t and accumulated there with elaps othee th f timeo e r n O hand. , no radioactivity was detected at the TEDA filter downstream of the column. It is therefore evident that zeolite 13X is capabl f retainino e g successfully CHroot a 3Im temperature; a maximum amount of the adsorbed CH3I exceeded 50 mg/g-zeolite 13X. In a different run, a CH3I* concentration of 10 ppm, almost equivalen thao t tDOGn i ts realize wa , y coolinb d e th g e samCHTh 3eI . resultcontaine°C 8 s-7 wero t r e obtaines a d to the feature of 131I distribution in the adsorption column. 60r

20 h 20cm

T0 cm Gos

r Ai Corne, s go r [CHjJ i 100pp, J m

C N02] ,=il%ppm Gos velocity , 23 cm /sec Temperoture , 25 *C

0 2 4 6 8 10 18 20 Distonce from the inlet of column,cm

Fig 2 .Distributio e zeolit f CHth o n 3X column Ii *13 e n after the gas supply of 5, 10, 15 and 20 h.

When supply of the gas was started, a yellow adsorption band of N02 was first noticed at the bottom of the zeolite bed. Behind this band, the zeolite bed turned pink due to the adsorptio iodinef no .r strea ai Whe e mnth containing CH3I* alone (100 ppm) was supplied to another adsorption column, 13 CH3I was also adsorbed on the zeolite, showing a peak of distributio same th e t positioa n Fign i s . na However2 n ,i thi st colore no case zeolite s allt a d,th wa d . ebe Therefore , appearance of the pink adsorption band described above means that the decomposition of CH3I* into I* took place in the presence of N02. The position of the pink adsorption band coincided witfiguree th peae he adsorptio n th thaTh i k . f o t n front of N02 traveled much faster than that of I* in the directio s supplyga e e formef flowo nth ,th f :o r afteh 0 2 r reache distanm positioa dc 4 1 tf no e fro inlee th th mf o t column, wherea dstanca t lattee m a c th s f onl s 5 o e rwa 3. y froe inletth m .

57 4000 3200 2400 (5001000 400 Wove number cm"1

Fig 3 .Infrare d spectr © CH f N0) o a3 I2(2 ,, (2 NCh^^Mi), (D the free gas remaining

afte' adsorptioe th r n process © CHd NOj,an . 5

Heating the adsorption column in an air flow resulted in d thean nC ° desorptio 0 20 - 0 f iodinno 10 th t ea ereleas 2 N0 f o e beyond 300 °C. A large amount of iodine (I2) crystallized in the Pyrex tubing between the adsorption column and the TEDA filter. Hence, it became clear that CH3I was decomposed by N02 on the zeolite:

Another adsorption column charged with s zeolitwa X 13 e . h Methy 5 r ltreatefo iodidC ° d0 e 10 witprocese t th a h s ga s was adsorbed completely, showing a peak of 131I distribution centered at the position of 3 cm from the inlet of the column; however, the yellow adsorption band of N02 no longer appeared anywhere in the column. Keeping the column at 100 °C in an r floai w cause adsorptioe th d o travet n* I banl f graduallo d y directioe th n i f floo n w with elaps f timeo e . ZeolitX e13 separates iodine chromatographically from N02 at around 100 °C.

. Reactio2 CHf Cd o n23 HIan 5 I wit2 N0 h

Since N02 was found to decompose CH3I, we tried to identify the reaction pruduct by means of infrared analysis and mass spectrometry. Figure 3 shows the results of the infrared analysis. spectre th CHN0e f d o a3ar Ian 2 , @ d an D ( Fig- 3 . respectively. D show( Fig spectrue - .free th s3 s th ega f mo which remained in the reactor after the adsorption process. In this case gaseou,a voluma sn i mixtur e2 N0 CHf d o e3 an I s introducewa 2 rati1: f o d intreactore th o e amountTh . f o s 2 werN0 CH ed 17.d respectively, 3119.Ian an 4mg g 4m e Th . spectrum reveals presence of N0> N20 and other unknown substances. Mass spectrometry indicated thamaie th tn component of the remaining gas was NO and the amount of N20 was very small t furtheI . r revealed tha o fren t e oxyge s presennwa n i t the gas. Afte analysese th r s evacuate wa e fres ,th ga e y b d pumping.

58 Fig. 3-0) suggests the presence of other products adsorb e zeolite th e reacto th n e o do s warmes ;rwa d slowl n ordei y o t r desor unknowe th b n products. thei, Beyon°C r0 5 ddesorptio n became observable pressure th ; e increased gradually with increasing temperatur d reachean e valu0 tora d 3 4 kPa ( rf o e ) at 120 °C. Fig. 3 - © shows the spectrum of the gas desorbed. Analyse e infrareth f d masso an ds spectra reveales dga thae th t desorbed was methyl nitrate (CH3N03) in an almost pure form [8].

When a gaseous mixture of CH3I and N02 was kept in the absence of the zeolite for one hour, its infrared spectrum was merely a superposition of those of the two substances. The reactio CHf o n 3I wit 2 needh N0 presence zeolitee th s th f o e ; s thereforii t e written dow s followsna .

C 5 =2 (1/2)I> CH 32N0ICH+ . + 2 3— NO 2N0 + 3

In different runs silver-exchangee ,th d zeolite (AgXs )wa treated with the CH3I - N02 mixture under the same experimental conditions as described already. The same reactions were noticed thereof.

Whe e [N0th n 2]/[CH3I] rati s increaseowa a o fou t dn i r different run, the infrared spectra obtained consisted of the absorption bands of CH3N03, NO, and excess N02.

Although CHbees 3Iha n postulatemaie th n e componenb o t d t organie oth f c iodide n DOGi s , ther s stili e possibilita l y that other organic iodides having molecular weights greater than that of CH3I may be present therein, e.g. ethyl iodide (C2HsI). So the same experiments were performed for C2HsI. The result was formation of ethyl nitrate (C2H5N03), I2, and NO:

1//2 12+ ^' C2^H 52N0I+ CaHsNO . 2 NO + s

Thus, the present work has confirmed that such organic iodides as CH3I and C2HsI react with N02 in the presence of zeolites, producing I2, the alkyl nitrates and NO.

3. Behavior of CH3I in the simulated DOG on zeolite 13X In order to obtain the information about the reaction rate, a simulated DOG was supplied to the adsorption column (25 °C) alineaa t velocits ga r f =2.o y 3 cm/sec e experimentaTh . l conditions were describe Figurn i d experiments. e 1 ePrioth o t r , 131 either I2 or CH3I was labeled with I in order to distinguish their behaviors experimentse th n I . simulatee s ,th wa G DO d supplied for 25 h at room temperature, and then the column was s agaiwa ; s finallnh ga supplie1 e r th yfo o C t d° 0 heate40 t a d the column for 6 h at room temperature. At the end of each step, distribution of 131i was examined.

streams ga e Figur"th n i show4 e 2 resultse "I th s e Th . curve in the figure shows the distribution of the I2 which was trapped directls ga s strea ye ga froth e mf th mo durin h 5 2 g 2 frosupply"I me CHTh .3 I" curve correspond distribue th o t s -

59 Corner gas , Air X ,= tOOppm ro [CHjI] , = lOppm

Gas velocity , 2 3 cm/sec Temperatur *5 C2 , e

O E

2 4 6 8 10 18 20 Distance from the inlet of column , cm

Fig. 4 Distribution of the adsorbed iodine (12) alon e e zeolitlengtth th g f X columno h13 e , (The gas supply, 25 n;

2 whicI e hth tio originallf no y came fro streae m th CH n 3mIi throug reactios it h n wit 2 durinhN0 firse . suppls th gh) ga t5 (2 y The location of the two peaks coincides each other: in both cases, iodine concentrate vicinite o n th inlee d th n i dan tf o y radioactivit s detectablwa y TEDe th A n filtero e .o tw Ther e ear findings available for assuming the reaction mechanism: (i) as described already, the CH3I - N02 reaction does not take place in gas-phase, and (ii) zeolite 13X adsorbs CH3I itself and there is little differ- 131 ence in I2 distribution between adsorptions of CH3I 2 mixtureCHa N0 f 3I- o alon d .an e Therefore e firsth , e treactio th ste f o p n woul e adsorptiob d f no CH3I on the zeolite and then N02 oxidizes in situ the CH3I adsorbed.

Heatin columr e floai th g w n i nresulte desorptioe th n i d n of N02 beyond 100 °C and also in small migration of the adsorbed zeolite iodinmor, th °C e n 0 i e% thae 40 0 bed 9 nt A . iodine o th fs desorbe ewa withid d be . froh e n1 th m

Subsequently the simulated DOG was again supplied to the 2 I e streas th ga d e an m th n i 2 I zeolite Th e. h colum6 r fo n from CH3I were also located at the same position of the column as in the first adsorption. The zeolite can be used repeatedly in removing iodine from DOG.

60 Drying Iodine ennchmen!

H removol)

Iodine fixotion Cul

Fig. 5 An iodine removal process using zeolite 13X.

IV. Conclusion

Zeolite 13X removes both I2 and CH3I with high efficiencies conditiony dr unde e 2 th rcoexists N0 o lon s ;s a g presence ,th e of such organic iodides as CH3I and C2HsI does not bring about additional problems e stabilitTh . f adsorbeo y d iodins i e however considerably d lowe an r zeolit X fo r Ag X thar 13 e fo n other silver-impregnated adsorbents. For the long-term storage, the iodine adsorbe n zeolito d X shoul13 e e transferreb d n a o t d appropriate solid material to form stable complexes or compounds,

We are considering an iodine removal process illustrated i n e dissolveFigurTh . 5 e r off-gas whic s driehwa d wita h =10t a d a zeolit0, C be o C t X d 13 efe f s silicfilteo i l d age rbe where iodin s trappeO i eN passe d an ds withoud througbe e th th being adsorbed e adsorbeTh . d iodin s thei e n desorbed froe th m s allowei d bereaco an t =40t da d C ° 0t with coppe o fort r m Cul, e long-terth whic o t hd mwoulfe storagee b d .

REFERENCES

[1] Burger, L. L. and Scheele, R. D., "The Status of Radioiodine Contro Nuclear fo l r Fuel Reprocessing Plants", PNL-4689 (1983).

HighlA " Kellerd , an yMaeck] Pence, H. . [2 J. . T ,J . . ,W D Efficient Inorganic Adsorber for Airborne Iodine Species (Silver zeolite development studies)11, CONF-680821, (1968) 185.

[3] Wilhelm, J. G., "Trapping of Fission Product Iodine with Silver Impregnated Molecular Sieves", KFK-1065, (1969).

Sakurai] [4 Izumo, ,T. Takahashi, ,M. Komakid an , . ,A , T. J. Nucl. Sei. Technol. 0 (1983,2 ) 784.

"IodinEggleton, d J. an . . E Atkins] F e [5 . . Compound,A H . ,D s Forme Releasn o d Carrier-Fref eo e Iodine-131", AERE-M-1211, (1963).

61 Noguchi] [6 Murata, ,H. Tsuchiya, ,M. Matsui, d ,Y. an . ,H Kokubu MethoA " , Quantitativf o d,M. e Measuremenf o t Radioiodine Species Using Maypack Sampler", JAERI-M-9408, (1981).

[7] Lange . (Ed.)A . ,N , "Handboo Chemistry"f o k , Handbook Publishers, Inc., Sandusk, OHIO, (1956) 1424.

] Strinhagen[8 Abrahamsson, ,E. Mclaffertyd an . ,S , W. . ,F "Atlas of Mass Spectral Data", Vol. 1, Interscience publishers, New York, (1969) 97.

62 RADIOACTIVE RUTHENIUM REMOVAL FROM LIQUID WASTES OF "Mo PRODUCTION PROCESS USING ZINC AND CHARCOAL MIXTURE

. MOTOKIR IZUMO. M , . ONOMAK , . MOTOISHIS , . IGUCHIA , , T.SATO*,T. ITO Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken, Japan

Abstract 5 23 9 9 reactioe th productioe y b f Th o no M U(n,f f no ) using UC>2 as target material was carried out in JAERI. Medium- and high-level liquid wastes arising from this production process contained U, Pu and fission products in which radioactive ruthenium in various valency and complex states was included. removae Foth r f radioactivo l e ruthenium, after testing various combinations of adsorbents, a novel method using a column packed with adsorbents composed of zinc and activated charcoal (charcoal developeds )wa e effectTh . f compositioo s n mixturesf o , flow rates f feeo dH ,p solutions, washingd ,an repeat use on ruthenium removal efficiency were investigated. A feature of this column method is that the efficiency of removal can be recovered by washing with dilute nitric acid and it is possible to treat a large quantity of liquid wastes. It was found that this column method was also effective for the remova 239p, U f o ul> 144ce, 155 d E125sbuan . This method was applied to the treatment of actual medium- and high-level liquid wastes quantitiee ,th f whico s h were 600L and 124L respectively. After the removal of U, Pu and fission products by conventional removal methods such as coprecipitation, filtration, electrolysi d adsorptioan s n with inorganic adsorbents, residual lO^R s removewa u usiny b de colum th g n packed wit mixtura h f zineo d charcoal an c . Decontamination factors for 106RU ranging from 1()2 to 104 were obtained. The

concentration of U, 239p 144ce, 155Eu, and 125sb after treat- j

ment were beloe foun b e limio th t wddetectionf o t .U

1. Introduction »q Through 1977 into 1979, weekly productio froo f 20CM no mi

fission product of 235] carriej d out in JAERI. In the pro- s

duction, irradiate s dissolvewa 2 U0 dnitrin i d wa c acid 99Man do s extractewa d frosolutioe th m n with di-2-ethylhexyl phosphoric 99Me th acido s productioA . n procedur s similae ewa th o t r Purex process, the liquid wastes resulting from the production were also very much like that from the nuclear fuel reprocessing. It has been well-known that radioactive ruthenium might be presen multipln i t e valenc d complean e x states simultaneousln i y nitric acimose th dt f solutioo questionabl e on s ni t [1-7i ed an ] nuclide treatmene th n i s radioactivf o t e wastes [8-15]. Develop- n efficiena men f o t t metho removaf o d radioactivf o l e ruthenium, therefore, has been much expected. e treatmene coursI th th nradioactive f th eo f o t e wastes arising fro99Me th mo production nove,a l procedure [16,17s ]wa developed for the removal of radioactive ruthenium on the basis

leavn O e* from Mitsui Minin Smelting& g Co., Ltd.

63 240g Irradiated DO.. 235 (2.6% U) _L Trappinf o g Decanning noble gas

Dissolution Distillation 10M HN03 IL

Trappinf o g Dilution iodine H.O 0.8L

r Extraction 30% D2EHPA-CC14 J~ Washing (1) 5M HNO 0.5L

IM HNO IL Washing (2) 3 0.5M HN03 0.5L _L Û.5M UNO, 0.48L Back extraction ~r 4 }12°2+30 20m1' Washing CCI. 0.9L

Washing waste of apparatus Medium-level liquid waste Aqueous phase

Decomposition of NaN03 30g in H.

Purificatioy b n IM NH.OH 0.03L A1O column

duality control

99Mo products (20Ci)

Fig.l Flo wo productiochar^M f o t n from irradiate2 UÛ d

of columns packed with a mixture of activated charcoal (charcoal) and metal powder, of which zinc was found to be most effective. The usefulness of this method was demonstrated by the successful results obtaine e treatmenth n i d f totao t l amoun f actuao t l wastess naturai t I .l tharadioactive th t e ruthenium removal procedure developed here shoul e applicablb d e treatmenth o t e t of wastes occurring frovarioue th m s e phasenucleath f o sr fuel cycle. The present paper describe resulte th s f developmeno s , of t treatmene th e actua anf th o d f lo t wastes with novee ,th l procedur radioactivf eo e ruthenium removal. 99 2. The nature and amount of liquid wastes from Mo production The procedure of 99Mo production and the origins of wastes procedure ith n e showar e Fig.l n i ne amoun Th d .character an t - istics of liquid wastes are summarized in Table I together witdecontaminatioe th h n factor necessar r eacfo yh nuclido t e e disposeb s low-levea d l liquid wastes.

64 Table I Radioactivity concentration of liquid wastes and decontamination factor (DF) necessary for disposal s low-levea l liquid wastes (ß-10-3-10: y - a:<5xlO-5yCi/mL)

Medium-level liquid waste (Assayed in March 1981) Radioactivity Nuclide concentration DF (UCi/mL)

106Ru x 10~ 6 2 60

137CS 1 x 10"1 100 144Ce 2x 10" 2 20 a x 10" 2 4 4

6-Y 4 x 10"1 400

pH -13 NaNO 2 0.027M U 3. Omg/L volume 600L NaNO. 0.3M

High-level liquid waste (Assayed in July 1981) I II Radioactivity Radioactivity Nuclide concentration DF concentration DF (IJCi/mL) (UCi/mL)

106RU 2.5 2.5xl03 5 5 XlO3

137Cs 11 11 14 14 144Ce 22 22 55 55 125Sb 0.2-2 0.2 " 0.34 0.3 " 155Eu 0.21 0.2 " 0.28 0.3 " 90Sr 11 11 14 14 147Pm -3 -3 -3 -3 a 0.03 0.6xl03 0.05 1 xlO3 6-ï 48 48 xlO3 89 89 xlO3 U : 23mg/mL volume I:73L, HNO: I:3.BM, , II:4.5M

The medium-level liquid waste consisted of the scrub solutio f sodiuo n m hydroxid removae eth use r radioactivf o lfo d e iodine and dilute nitric acid used for washing of the apparatus. This waste contained a considerable amount of sodium nitrate and was supposed to be equivalent to the washing effluent in the nuclear fuel reprocessing plant in terms of the radioactivity level [18]. Since this wast s somewhaewa t alkaline, some part of radioactive ruthenium of chemical forms that was prone to precipitate was supposed to have formed coprecipitate with sodium uranate. Whil e high-leveth e l liquid waste consisted of the nitric acid solution of irradiated UC>2 from aqueous phase after extraction with organic solvents (di-2-ethylhexyl phosphoric acid and carbon tetrachloride) and washing solution of the organic solvents and was stored in two tanks, tank-I an de high-leve Th -II. l liquid waste her s equivalenewa o t the medium-level liquid waste fronucleae th m r fuel reprocessing.

65 In orde o decreast r e radioactivitth e y concentration i n these liquid wastee acceptablth o st e e low-levelevelth f o s l liquid wastes definer Instituteou n i d , lO^Ru decontamination factors of >60, >2.5xlO^, and >5xlO^ were required respectively for the medium-1evel, high-level liquid wastes-I, and -II (see Table I).

3. Development of Ru removal techniques In preliminary experiments o effectiv,n e metho s founwa d d removae foth r f 106Ro l f chemicao U l forms whose separation were difficul y conventionab t l means suc s coprecipitationa h , adsorbent columns, and ion exchangers as previously reported on treatmen f liquio t d wastes from nuclear fuel reprocessing. In the experiments above, it was observed that several kinds of metal powder d charcoaan s l were considerably effective th r fo e removal of radioactive ruthenium when used separately. Based on these experiences, a method using a column packed with a mixture of metal powders and charcoal was devised and tested. In the experiments a mixture of zinc powder and charcoal was foun havo t d e high removal efficienc r 106Rfo y U. 3.1. Radioactivity measuremen d decontaminatioan t n factor Decontamination facto f eacro h nuclid s calculatewa e d from the ratio of concentration before and after removal treatment f samplo e solutions e concentratioTh . s obtainewa n d respectively by measuring y-ray intensity with a y-ray pulse-height analyzer using Nal(T£) and Ge(Li) detectors, a-ray intensity with a 2-rr gas flow counter, and ß-y ray intensities with a gas flow GM counter. 3.2. Waste solutions used for experiments Various chemical species of ruthenium were contained in nitric acid solutio n solutioi d an nf sodiuo n m nitratd ean species which could be removed without difficulty by coprecipitation and filtration were also included. In order to examine the efficienc removaf o y rutheniuf o l e for f anionth o mn i m s or in abnormal chemical state where it is difficult to be separated from the solution, ruthenium components which could be removed easily shoul e separateb d pretreatmeny b d n advancei t . Such components were coprecipitated with aluminium hydroxide flocculate generated from electrolysi d filtere e an cass th en i d of medium-level liquid waste, and they were coprecipitated with sodium uranat neutralizatioy b e nitrif o n c acid with sodium hydroxide in the case of high-level liquid waste. Decontami- nation factors of 106RU obtained by these coprecipitation

method smedium-levele werth -20, d r ~4 ~10e an fo , , high-level liquid wastes-I respectivelyI -I d 2 an , . Each solution obtained by the pretreatment was named the pretreated medium-level, high-level liquid waste-II -I d ,an respectively. thesf o H ep waste Th e sample solution s adjusteswa d with nitric aci r sodiuo d m hydroxide prio o experimentsrt e th r Fo . preparation of solution containing all chemical species of 106RU) uranium and 239p were extracted from the high-level liquid

waste-II with 30%u tri-n-butyl phosphate in kerosine. All chemical species of 106Ru appeared to remain dissolved in the residual aqueous solution sinc o l°en 6Ru radioactivits wa y detected in organic phase. This solution was named the ext- racte d sodiu an solution H mp nitrate Th . e concentration were controlled by adding sodium hydroxide.

66 Table II Decontamination factor of fission products with metal powder-charcoa f solutioo l H colump d nan n

- nu -x. NO. a 155EU -Ce 125Sb 106RU 137CS pH Méta. \x l

1 Al 2 1 1 1 8.3xl03 1 4.5

- - _ 2 Sn > 53 2.9x10 16 1.4x10 8.3x10 1 6.3 3 Cu 1 1 1 2 32 3 3.1 4 Fe 2 2.8xl03 l.OxlO5 1.3xl03 B.3xl03 1 6.0 5 Zn >340 4.9xl03 1.9xl05 2.2xl03 l.GxlO4 1 8.4

NaNO, : No. IM : 1.6 M, No. 5 : 2.5 M Flow rate : 0.24-0.4 cm/mm

5 1. : 5 : 2.7 . 4 No ,1- . No f pl o fee: i d

For other detailcs e text-se , .

3.3. Removal of 106RU using zinc-charcoal column In orde o fint r combinatioa d n adsorbena r fo n metaf o t l powde charcoad an r l suitabl r 106ßfo e u removal, mixture tinf o s , iron, zinc, or copper powder and charcoal were tested. Metal powders having a purity of 99.9% and a grain size of <100 mesh and charcoal (Tsurumicoal Co., Ltd. GVA) of 60 to 300 mesh were mixed in water and packed in columns of a diameter of 8 or 18mm. For the purpose of investigating the removal efficiencies for all chemical species of 106RU iQOmL of the extracted solution prepared abov s passewa e d through each column packed wita h mixture of 3g each of metal powder and charcoal and the decontamination factor f eacso h nuclide changH p th f d o e ean of solutions were evaluated as shown in Table II. Decontami- natio r 106ßfo n s 4 uwa obtainefacto 10 f o rusiny b d g these adsorbents excep mixtura t f coppeeo d charcoaan r d amonan l g these a mixture of zinc and charcoal indicated the highest removal efficienc nuclidesl al r peako fo yN .f 144ceso , ^^Eu,

125sb, 10(>Ru, j239p u in y-ray and a-ray spectra of the anc > passed solution weru e detected except ^"Cs. 3.4. Removal witu conditioR h zinc-charcoa r fo n l column In order to investigate the effects of zinc to charcoal ratio, flow rat f feeo e d n removasolutiono H p ld ,an efficienc y for lO^Ru, the following experiments were carried out. For these experiment pretreatee th s d medium-leve d high-levean l l liquid waste-I were use s sampla d e solutions, sinc e objeceth t of removal was chemical species of 106R whose removal were

difficult. U 3.4.1. Effect of zinc to charcoal ratio After adjustin 4.9o t pretreatee H ,p gth 30m f Lo d high- level liquid waste-I was passed through columns packed with a mixture of 0-lg of zinc and Ig of charcoal. The results obtaine e showar dFig.2n ni . Decontamination factof ro increased with zinc to charcoal ratio, indicated 55 at 40% rati d increasean o d aboumort a t 80% a ed t 2 an ,twic10 o t e than that decontamination factor approached saturation. Passin pretreatege th 30m f Lo d medium-level liquid waste through columns packed with Ig of 10-85 wt.% zinc-electrodeposited

67 a. o

z 10' o

§ 10 D 0 4 0 2 ) 60 80 100

% ZIN O CHARCOAT C L

: 0.01NaN0M3 Flow rat: 0.88cm/mie n BedL volum3m : e

Fig.2 Effect of zinc/charcoal ratio on decontamination factor of 106Ru

10J

K O EH g Z O £ 10*

EH Z O O W a

10 0 20 40 60 80 100 RATIO OF ELECTRODEPOSITED ZINC («) NaNO, 0.4M Flow rate 0.9cm/min Bed volume 3mL

Fig.3 ratie Effec th f electrodepositeo f o t d zin charo t c - coa n decontaminatioo l n factof o r

charcoal, whic s developewa h o improvt d mixine th e g condition of zinc and charcoal, relation between electrodeposition ratio of zinc to charcoal and decontamination factor of lO^Ru was studied e resultTh . e show ar sFig.3n i n . Decontamination facto f lOORo r u increased witelectrodepositioe hth n ratif o zinc and approached saturation at more than 30%. From this result it is recognized that a charcoal which has the electrodeposition ratio of more than 30% zinc could be utilized effectively. Decontamination factor attained saturation at more than 60% of zinc in the case of a mixture of zinc and charcoal appeart I . s that zinc-electrodeposited charcoal reached saturatio mixturw lo t a ne ratio, becaus f smaleo l particle sizd uniforean m dispersio particlesf o n .

68 10'

« g u S §

Z uO M Q

0 1 8 6 4 2 0 FLOW RATE (cm/min) Bed volume : 3mL Fig.4 Effect of flow rate of feed solution on decontamination factof o r

3.4.2. Effec f floo t w rate After adjustin o 5.2t H ,gp pretreatee th 30m f Lo d high- level liquid waste-I was passed through columns packed with a mixture of Ig each of zinc and charcoal at various flow rate. Figure 4 shows experimental results. Decontamination factor decreased with increasing flow rate, being -60 at Icm/min, -30 at 2cm/min, and 10-20 at more than 4cm/min. 3.4.3. Effect of pH After adjusting pH to 2.7-7.2, 30mL of the pretreated high level liquid waste- s passewa I d through columns packed wita h mixture of Ig each of zinc and charcoal. The results are shown in Fig.5. Decontamination facto f lO^Ro r s affecteuwa d with concentratio nitrif o n H cp acit becamd a pH6t a an d0 0 ~8 ,4 e 2-3, and it approached saturation when pH was lower than 4. Similar results were obtained in the experiments using the pretreated medium-level liquid waste.

• Zinc-charcoal column O Charcoal column

NaNO3 0.01M Flow rate 0.88cm/min Bed volume 3mL

Fig.5 Effect of pH of feed solution of decontamination factor of l°6Ru

69 3.4.4. Effect of differences in volume of feed solution The optimum condition for 106^U removai obtained from experimental results was as follows: Ratio of zinc to charcoal = >1:1 pH of feed solution = ~3 In this condition relation between volum f feeo e d solution and decontamination factor of 100RU was studied by passing 2.2L pretreatee th f o d medium-level liquid waste f whico s H wa h,p adjusted to 3.1, through a column packed with a mixture of 30g eac f zind o charcoalh an c e resultTh . e showar sn Fig.6i n . Decontamination factors of 10^RU Were >1C)3 at 0.4L feed and ~1() t 1.2a 2 L feede removaTh . l efficienc f thio y s colums nwa expected to attain to -102 provided that the volume of feed solutio s confinewa n d withif charo range g I - th nf 40m o er Lpe coal. Similar results were obtaine e experimentth y b d s using the pretreated high-level liquid waste-I. 3.4.5. Recovery effect by washing columns The zinc-charcoal column method required a large quantity of adsorbent for treatment of a large amount of waste, since decontamination factor decrease volume th f s feeo ea d d solution increased. With the condition above, 15kg (45L of apparent volume f adsorben)o necessars treatmene wa t th r fo yf tota o t l amount of medium-level liquid waste, 600L. The volume reduction ratio should, therefore, be about 13, since no effective proces s know swa o concentratt n e 106RU adsorbe n zinc-charcoai d l column. In order to reduce the volume of solid waste arising from the treatment, possibility of recovery of efficiency of adsor- e columth f repeaf no o bent were d us tan se examined using pretreated medium-level liquid waste. Immediately after passing waste lOOmth ef o Lsolutio n throug a columh n packed wita h

lü

g«

|io3

uo u

4 2. 0 2. 6 1. 2 1. 8 0. 4 0. 0 VOLUME OF FEED SOLUTION (L) NaNO, : 0.4M Flow rat: 0.8cm/inie n Bed volum : lOOme L

Fig.6 Effect of the difference in volume of feed solution on decontamination factor of

70 mixture of 4.5g each of zinc and charcoal, the column was washed repeatedly with wate r nitrio r c acid solutio d changan n f o e the removal efficiency was observed. It was found that although the column efficiency could be completely recovered by washing immediately after use with water or dilute nitric acid, washing with dilute nitripreferabls wa 3 c pH aci f eo d becaus o elutioen n of 106Ru was observed. The amount of washing solution necessary for the recovery of column efficiency was approximately equal to the volume of the column. The recovery of column efficiency by washing and repeat use e columth f no mad possiblt i e o increaset e amounth e f liquio t d waste treatee b colum th o d t eaccordingly b an nd o decreasyt e the volumresultine th f eo g solid wast s demonstrateea e th n o d

treatmen actuaf o t l waste below' . 3.5. RemovalR uof 106 from actual liquid wastes [19,20] For chemical species of ruthenium which were difficult to e removemedium-levee b th n i d l liquid waste, zinc-charcoal columns and zinc-electrodeposited charcoal column were used after treatment first by electrolytic flocculation, then with zeolite column, and finally with titanic acid column. For the treatment of the high-level liquid waste, after dilution, neutralization, coprecipitation with nickel ferrocyanide, and adsorption with a slurry of ortho titanic acid, the solution was passed through the zinc-charcoal column. A column wit diametea h volumf o L 5 f 149me o r d packean m d wit mixtura h f 2.5ko e f 60-8o g 0 mesh zinc powde d 1.5kan r f go 60-300 mesh charcoal was prepared for treatment of the actual medium- and high-level liquid wastes. The liquid waste diluted with equal s passevolumwa L watedf 30 o e througo rt e columth h n and then washed repeatedly with 5L of washing solution of pH2~3. Relations between the volume of feed solution and decontamination factor of lOoRu in the treatment of the medium-level liquid wastes were shown in Fig.7.

10" f

S 102 2

g Oy

10° solutioU ( n 30 60 90 120 ISO IBO 210 240 270 300 339 30} 390 420 450 f feeo dH p solution ,34,30,3^18^18^8^8^21,19,22.22,22,201 pHof effluent 8890 100)0 083 82 7 170 7 085837270 71 7 Ol Flow rate : 1.25cm/min

Fig.7 Removal of 106RU from medium-level liquid waste with zinc-charcoal column

71 e treatmenTh medium-levee th f 450o tf Lo l liquid wasty b e repeat use of a single column resulted in decontamination factor of 10-20 at pH3, and 102 at pH2. Similar results were obtained using 31% zinc-electrodeposited charcoal as adsorbent. In the treatment of 280L of the high-level liquid waste decontamination factor of 102 with a feed solution of pH2 was obtained by washing with dilute nitric acid.

4. Conclusion e conclusionTh se experimenta baseth n o d l resultd an s interpretatio s followsa e date th ar a f :o n ) Colum(1 n packed wit mixtura h metaf o e l powde d charcoalan r , especially zinc-charcoal mixture d zinc-electrodeposite,an d charcoal showed high efficienc chemical f removao yal r fo ll species of 106RUj in particular for the species that were difficult to be removed by the conventional methods such as coprecipitation, aggregatio n exchangeio d an n , with high decontamination factor. (2) Besides for lO^Ru, column packed with zinc-charcoal mixture was also highly efficient adsorbent for the removal of 239pu> U, 144Cej 15S d Eu125an > Sb. (3) Zinc-electrodeposited charcoal was preferably used for the preparatio f packeo n d column becaus particlee th e metaf o s l were dispersed uniformly. (4) The optimum condition of removal of Ru by the present method was to use zinc-charcoal column or zinc-electrodeposited charcoal colum f 30-80o n f zin%o c e contenth r fo t treatment of liquid waste within the acidic region of pH<3, e floath tw rat f ~2cm/mino e . (5) The efficiency of 106RU removal with zinc-charcoal column coul e recovereb d y washinb d g with wate r diluto r e nitric acid after use e colum.Th n could, therefore, repeatedl usee b y d for the treatment of a large amount of waste. Althoug usefulnese th h f zinc-charcoaso le columth r fo n waste management in the field of nuclear fuel cycle was demonstrated above, several problem e solveb e lef o ar sy t tb d further research as follows: ) Improvemen(1 efficience th procedure f o tth f o yr remova efo l f 106R.lo d othean i r radioactive elements. ) Extensio(2 modificatior o n presene th f o nt technique b o t e applicabl r treatmenfo e y batcb t h system. (3) Elucidation of the mechanism of radioactive ruthenium removal. Studies concerning these problems are now in progress and the results wil e soob l n published elsewhere.

Acknowledgements The author . Nakamura. H Sato T gratefu e . . ar s,Dr Mr o ,t l Mr. Y. Kawakami, Mr. K. Suzuki, and Mr. T. Abe of the Japan Atomic Energy Research Institute for their useful suggestions and discussions . giveus o t n References ] t al.e FLETCHE[1 , . InorgJ ,. M . RJ . Nucl. Chem (1955_ l ., ) 378. [2] ZVYAGINTSEV 0. E., et al., Peaceful Uses of Atomic Energy (Proc d Int2n . . Conf. Geneva, 1958) Vol.17Yorw Ne k, ,UN (1958) 130.

72 t al.e , , M. Peacefu] . Brow[3 G . P nl Use f Atomio s c Energy CProc d Intwn . . Conf, Geneva, 1958) Vol.17w YorNe k , UN , (1958) 118. [4] FLETCHER J. M. , et al., J. Inorg. Nucl. Chem. , j_2 (1959) 154. [5] KOYAMA M., Nippon Kagaku Zasshi, 82_ (1961) 1182. ] SUGIMOT[6 Radioisotopes, OS. (1979, ,28 ) 429. ] AKATS[7 Rep, E. U. JAERI-M 9159, JAERI, Tokyo (1980). [8] KRIEGER H. L., et al., Rep. ORNL-1966, Oak Ridge National Lab. N (1955)T , . [9] FEBER R. C., Process Chem., 2_ (1958) 247. [10] GARDNER E. R., BROWN P. G. M., UKAEA Rep. AERE-R 3551 (1960). [11] ISHIYAM al.t e , RadioisotopeT. A (19665 s_L ) 181. [12] LYNCH R. W., et al., Rep. SAND-75-5907, Sandia Lab., NM (1975). [13] KENNA B. T., Rep. SAND-78-2019, Sandia Lab., NM (1978). [14] SUGIMOTO S., SAKAKI T., Radioisotopes 2£ (1979) 361. [15 ]Energ. t al.At KANNe . , J y OT. Soc . Jpn. (1979_ ,2l ) 737. [16] NAKAMURA H., et al., JAPAN PATENT SH057-50698 (1982). [17] NAKAMURA H., MOTOKI R., et al., IAEA/WMRA/13-81/11 JP04 (1982). [18 Energ. ] At SEGAW. J y , SocAT. . Jpn. (19702 ,1_ ) 599. [19] IZUMO M., MOTOKI R., et al., Rep. JAERI-M 83-197, JAERI, Tokyo (1983). [20] MOTOK t al. e , IZUM , R. I, M. ORep . JAERI-M 84-015, JAERI, Tokyo (1984).

73 ION EXCHANGE CHARACTERISTICS OF HYDROUS OXIDES SEPARATIOANE DTH THEIN I E RUS URANIUF NO THORIUD MAN M

N.Z. MISAK, H.N. SALAMA, E.M. MIKHAÏL, I.M. EL-NAGGAR, N. PETRO, H.F. GHONEIMY, S.S. SHAFIK Nuclear Chemistry Department, Nuclear Research Centre, Atomic Energy Establishment, Cairo, Egypt

Abstract The ion exchange characteristics of several hydrous oxides have t.een studied. The kinetic studies done with zireonia and ceria have s'jown tha increase tth charge th cationf f o o eminoa s rsha effect on their mobility in the exchangers while the increase of charge or cornplexing ability of the anions decreases their mobility. A. dehydra- catione th tio f o ndeduceds s i mobilite .Th cationf o y d anionsan s necreases in the presence of alcohols. For alkali Cation exchange, it was deduced from thermodynamic Oata on hydrous zireonia and ceria that the Eisenman model is qualitatively sometimed ,an s quantitatively, applicablen ca thad ,an t t i L>e used for obtaining successful predictions of the trends in the thermodynamie functions, ïhis is not the ease with anion exchange. The thermodynamic studie hydroun so s zireoni methanolin ai c solutions show that the cation and anion exchange selectivity constants c-in be, at least in certain cases, predicted from the values in the aqueous medi consideriny ab g ion-solvent interactions« However, even in these cases large changes occur in the solid phase. The physicochemical properties (acidity, capacity and porous texture hydrouf )o s oxides hav profouna e d effec- ex thein to n rio change behaviour. Besides, water structure effects seem to be of i:reat importanc case th f aluminao e n ei e decreas.Th porositf eo d yan pore size seemincrease th leao st o t dhydrouf eo s oxide selectivity. Sorption studies of uranium and thorium from aqueous solutions on hydrous eeri frod an em both aqueou oix*d««olvend an s t medin o a hydrou n oxid sti hydrouf e possibilite e pointh us s o t e th f yo oxides for the separation of uranium and thorium from each other and from other elements.

INTRODUCTION Hydrous importann oxidea e sar t clas inorganif so exchangersn cio . Fields of their potential application are: chemical seprations and purification particulan ,i radioactivr rfo e materials exchangn ,io e catalysis at high temperatures, treatment of high-^teaperature water, and purificatio moderatof no cooland ran t wate nucleaf ro r reactors (1,2). Of course, a rational use of hydrous oxides in the various fields necessitates a good knowledge of their general and sorption properties and in particular their ion exchange selectivity behaviour. For this purpose, many hydrous oxides were prepared and studied in this laboratory. This paper review wore sth k done.

EXPERIMENTAL samplee th l sAl prepared were initially drie 50°t stored a d Can d in a nitrogen atmosphere over saturated ammonium chloride. Some of the samples were heated at different temperatures and the surface characteristics were measure nitrogey db n adsorption isotherms.e th Mos f o t samples were subjected to X-ray, thermal and IR analyses. When different sample same th e f shydrouo s oxid involvede ar e , thereferree yar o dt metasymboe e byth th oxidle f lo th ato ef o mfollowe Romaa y db n number.

15 Hydrous ceriCe-Id an Ce-: aII were prepare previousls a d y mentioned (3,4). Ce-I (3) is amorphous and its DTA curve shows a large endothermic peak (only endothermie peaks indicativ losf eo watef so hydroyr ro l groups will be mentioned) at about 125°C. Ce-II (4) is slightly crystalline. Its DTA curve haß a large endothermic peak at a bout 130 C, a very small one at about 100 C and a small one at about 24O C. The IE spectra show that in both samples, the peaks due to water and due to both water and. hydroxyl groups decrease continously with increase of temperature. Hydrous zirconi Bio-Raa s i Zr-a: dI HZO-1 hydrous zirconium oxide (5), whereas Zr-II was prepared as given before (6). Zr-II ie amorphous. Hydrated alumina ( or aluminium hydroxide) : Al-I, Al-II, Al-IV and Al-V were prepared according to the methods given before (4). Al-III ireadsa y JHDEX chemical designate Al(OH),s a dalumine th - l .ahy Al drates prepare semieryetalline ar d theid san r "degre crystallif eo - nity remains almost the same up to the heating temperature ?50 C. All the samples give endothermic peaks at 270-280 C. Besides, very small endothermic peaks around 100 C are present in the DTA curves of Al-III and Al-IV analyseE .I X-rayd an ) sho s(4 T , wD tha aluminae tth e sar generally composed ef one er more of the trthj-arxidee (gibbeit«», b..y- BT-it», norAct*ttna.ite; with varying amount amorphoun a f so s oxyhydro- xide. Ho significant changes occur in the samples up to 250 C. At the, transformee C ar y0 35 alumino dt a which sorbs water froe mth atmosphere, partly as hydroxyl groups. Ferric oxide gel : The methods of preparation of Pe-I. Pe-II and Fe-II givee Iar referenc n ni Fe-I^ virtualle »n Fe-, Iar I e4- y amorphous while Pe-II Bemicrystallins Ii e (4) erystallinite .Th l al f yo the samples Increases wit increase hheatinth e th f eo g temperature. The DTA curves show small endotheviiic peaks around 100 C in all the samples, large endothermic peaks in Fe-I and Fe-II at about 250°C, and two small endothermi Fe-III.Fron i C O c3O peakE d mI an about sa 0 t25 analysis deduces wa t ,i d thagele tth s begi loso n t group H eO temt sa - peratures ^,150 C. At 400 C, the samples still contain OH groups« Be- sides, molecular wate presents ri , whic probabls hi y sorbed froe mth atmosphere. Hydrou oxidn sti Hydrou : e oxidn sti e (B-stannie acid prepares )wa a da mentioned previously (7). B-etanni slightls ci aci) (4 dy crystalline. Its DTA curve shows a large endothermic peak at about 135 C. The IB spectra temperaturshoe watee th th w o r- , that C re los p tu 0 s ti e40 gained by sorption from the atmosphere.

RESULT DISCUSSIOND SAN S

1-Review of Previous Kinetic and Equilibria Studies on Zr-I, Zr-II and Ce-I Brit Kancollad Ban hav) s(5 e shown that Zr-I display selectie sth - vity pattern: Li> Na*> K, i.e., the cation of the smaller bare radius is preferred. These authors suggested tha alkale tth i eaction retainee sar d exchangee inth substantialla n ri y dehydrated form kinetie «Th c studies done later by Misak et al. (8,9) on Zr-II showed that the increase of the charg cationf eo generalla s sha y little effec thein te r mobility inside zirconia mobilite .alkalTh e th eart d f yio an h alkali cationn si zirconia decreases inthe order: C6>K>Ca~-Sr~B«>Li>Ia. This order support dehydratiosa catione th f no s sorbed, Kinetic studie hav) (3 Ce-n eso I shown tha mobilite tth somf yo e alkali and earth alkali cations decreases in the order: Ca~Na> E> Ba. This shows also that the cations are sorbed inside Ce-I in a dehydrated state monor .bivalend Fo -an t anions mobilite ,th Zr-In yi ) I(9 founs wa decreasano dt d) (? Ce-I e wit increase habilitth e th f eo y e anioo fth eomplexinr nfo g wit metae hth hydroule th ato f mo s oxide» presence th n I alcoholsf eo internae ,th l diffusion eoefficientf so the Na/H and Cl/OH systems decrease due to a stronger interaction of the counter iems with the exchange sites (6)* Accordin Eisenmao gt n (10) selectivite ,th alkalr yfo i- ionso f so lids containing fixed charges depends on two factors: ion hydration and electrostati bindingn cio . Whe hydration n io dominan e th s ni t factor,

76 the selectivity pattern IB: Cs> K >ffa>Li,i.e. , a préférence for the alkal witn iio largeha r bare radiu lesser so r hydratioa energ diss yi - played» The completely reversed sequence results when electrostatic binding is the dominât facto*. Partial reversals result la the inter« mediate conditions* The Eisenman model considers the sorption of ions in a completely dehydrated state. Besides, it takes into consideration only eoulombic electrostatic interactions, and the changes in the entro- anothey b t completeli e solie rar n th i replacin n df o pn o y io ye gon neglected. Nevertheless, this extremely simple mode bees lha n qualitatively successfully applie organio t d c resin Reiehebery sb g (11d )an to zeolites by Sherry ( !?.)• Sherry (12) found that the neglected entropy changes often rei'.forc selectivite eth y patterns predictey db Eisenman. We have trie teso applicabilite dt tth thif yo s mode hydrouo lt s oxides. When applying this mode catioo lt n exchang hydroun ei s oxides, reversed sequences may be expected to be favoured by the increase of capacity sjnd pH at which selectivity is measured, and by the decrease of the acidity of the oxide which increases the field strength of the anioni sites~ e0 thesl *Al e factors increas role electrostatief th eo c interactions. Increase of erystallinity may lead to this by increasing the lattice energy providin dehydration io r gfo n (5). Equilibria measurements were donmanr efo y cation saiond san n so Ce-I (3). '<-'iie result r alkalsfo i eationmanr fo yd anionsan eerin so a ina 3 other hydrous oxides were subjecte preliminara o dt y comparison. Regardles faee th tf sthao compariso e tth n often involved equilibrium distribution coefficients and that these coefficients may pertain only to a certain set of experimental conditions, the following observa« tion generaa sf o seemee lb natureo dt e sequenc.Sh preferencf eo f eo lessea f o rthn hydratioeio n energ observes yi d only wit more hth e acidic oxides: silica (13), alumina (14). titaaia (15 hydroud )am s manganese dioxide (16, 1?)» On the other hand. partially eat completely reversed sequences are observed on less aeidie oxides: «ireonia (5,8), oeria (3) and thoria (15)* For the anions, the résulta with the ma- hydroujorite th f yo s oxides (5,8,18-22) sho preferencn wa io e th o et of a higher eoaplexing ability with the metal atom of the oxide, which mean dominansa t rol electrostatif eo c interactions. This compariso mako urges t nha es du detaile thermodynamid d«a e studie severan so l hydrous oxidetestinf o m validise ai gth wit e hth - ty of the Eisenman model and examining the effect of th» various factors suc ariditys ha , capacit sitr y(o e density), porous texturd ean erystallinity ea their exchange behaviour, aad in particular the selectivity. The ultimate goal is to obtain a clear general picture for ion exchange on hydrous oxides* The first studies involved the thermodynamics of alkali eatioa exchange and aaion exchange in Ce-I (23,24-)* These studies involved exchangee come th th f o -d S saa £t , th H e &. déterminatio, G A e th f no t £tf , o SH &. , parison of the obtained data with similar studies oa Zr-^I (5) aad Zr-II (25)* these thermodynamie functions were resolved into two components pertainine :on solutioe th o gt n phas involved ean there sth - modyaamic hydration fuactions,^Gh, AH*1, £Sh, obtained from Rossein- •ky*a data (26), aad the other pertaining to the solid phase, £5*, £!!*,££*, aad is ebtaiaed from the relatioasi AG°« Gh

Th seamai eE n model considers onlfree yth e energy change (AG) of the exchange, aad this was the only parameter considered in the application of the model to resins (11) and zeolites (12)* We have tried to apply the ideas in the model also to the enthalpy ehaages ( A.H). The differen oxide replacemene t th otheth y date n b i r arfo d I alkalf to i ions for Ce-I (25), Zr-I (5) and Zr-II (25) are givea in Table I. considerabla s Zr-ha I y less water eoateat than var e Zr-Ith -n Ii ious cationie forms (23)* Heace Zr- Iapparentls i y mores densha d ean probably a higher site density due to a less average spatial separa- exchange tioth f no e sites capacitiee *Th oxidee th ver e f sar o y close and the ions seem to be substantially dehydrated in them since the water contents of the different ionic forms are rather similar. Ceria is lowef expecteo e rb acidito dt hencd yan havo et strongeea r field strength of its exchange sites.

11 Table I

Thermodynamic datd tor alkali cation exchange at 25°C on Ce-I*, Zr^I** and

•-t a 0> (D o l-l M V in U) -H 1 1 ' § )n l-t l-l S ü o o o o o r~*l Si Si ^i si o c .Ca K CO s o O O W CO 0) o> 0) o v ^ . ^ . Cl S <3 < < |'3 |U .U] a a ^

Li/Ea 5.69 0.00 -19.1 -100.0 -109.2 -30.9 105.7 109.2 11.7 Ce-I Li/Cs 9.08 6.0? -10.1 -227.3 -251.5 -81.2 236.4 257.6 71.1

Zr-I Li/Na 2.39 -0.71 -10.4 -100.0 -109.2 -30.9 102.4 108.5 20.5

Li/Ka 2.05 -9.60 -39.1 -100.0 -109.2 -30.9 102.1 99.6 -8.4

Zr-II Li/Cs _0<02 _2j.46 -78.7 -227.3 -251-5 -81.2 227.3 228.4 2. 0

frox m Referenc. 23 e aoe from referenc- 5 e soot from reference 25. Cosidering the above-mentioned facts and remarks« one may expect that a preference for the ion of a smaller bare radius would be more pronounce lese th s r acididfo c ceri porositf a (i secondar f o s yi y impe»- tancerelative to acidity), and for Zr-I relative to Zr-II. Besides, the contribution of £3* is expected to be higher for eeria than for zircon- Zr-r fo Id i athaan Zr-II r nfo . Since SÏÏ positiv s givei e 6 th - nex r *fo changee (Table I) .AH is expected to be less negative or more positive for ceria compared to zirconia, and for Zr-I compared to Zr-II, Table I shows thathesl al t e expectation fulfillee givee sar th nr dfo systems , which shows that the Eisenman model is qualitatively applicable to these systems. Tabl eshowI s that a£vers i >e y casth e smalth f e o n li Li/Cs exchange on Zr-II, which shows the quantitative applicability modee o fth thin li s ease. The study of the thermodynamics of NCC/Cl", NOT/Br" and NCC/SClT exchanges on Ce-I and the comparison of the results^obtained with those for Zr-I (27), Zr-II (25 anothed )an r zireonia »ample studie Ruvay db - rac and Trtanj (28) have shown the inapplicability of the Eisenman model in the case of anion exchange. This was deduced (24) to be due to solid entropy changes and aoncoulombic electrstatie interactions that hamper the applicability of the model in this «ase. The study of the thermodynamics of Li/Ha and Li/Cs exchanges on methano$ 50 d an l $ (29Zr-I 30 shows )n ha Ii n thavalue» H tth A f so and AS0, given in Table I, increase preressively on addition of methanol whereas the reverse occurs with AG . Thus, the selectivity constant entrope increaseth o t ye termsdu . Usinthermodynamice date th gth a a o s of transfe alkale th f rio ions from agueou methanolio st e solutions, ion-solveno t e du - S U tin d an th H eA change, valuee th G A f sn o si teraction addition so methanof no l were calculated. From these values, the changes in these functions due to changes in the solid phase were also calculated lattee .Th r changes were considérablefoune b o dt . How- ever, for the Li/Na exchange in J0# nethanel. ÛG calculated theore- ticall basie th ion-solven f sn o yo t interactions alom foums cwa o dt be very experimentae closth o et l value seemt .I s therefore thae tth solid phase changes lead to contributions to &H and UB° that largely cancel each other. The thermodynamics of halide ion/ nitrate ion ex- samchange th e n exchangeeo same th e n mediuri m (50) revealed similar trends. However, for the nitrate ion/ bromide ion exchange in 60$ methanol enthalpe th Q , A change o t y* ternddu .

78 2-Recent Studies on the Effect of the Pbysieochemieal Properties of Hydrous Oxides on their Ion Exchange Behaviour a-Kinetic f exchango s e The valuee diffusioth f o s n coefficient (trac* Cs f eo sion n i ) Al-I Al-d Van V (Na-form) fro ma 0.1 M NaOH mediu e givemar n Tablni I I e at different heating temperatures of the exchangers. This table contains e calculateals e datth oth f ao d e samplesmearadie th po nf o i »

Table II

Dirfusion coefficient n Al-Ii Al-d s t C V25°a an Vs f Co heated at different temperatures.

Heating teup., ilean pore DxlO Exchanger °C c Ae m radiusc A ,

50 44 861 Al-IV 250 27 67 350 22 6.9

50 77 398 Al-V 250 69 43.3 550 17 9 A

same th Table d sampleshowI an eI e son , thadecreaseD r tfo s with the decreas meae th n f poreo e radius. Sinc edoeD depent sno d onln yo pore dimension capacite alsth e exchangert sth n bu o f o y e strengt,th h of interaction with the exchange sites and the degree of dehydration(3l) diffusine ofth g species above-mentionee ,th d facshoy tma w thae tth pore width is the main factor governing the rate of ion diffusion in the exchangers. It must be mentioned here that the mean pore radius may not be well representative of the actual width of the pores in which diffusion occurs. This ie so because of differences in the geometries of pores becaus! an exchange eth e restrictee siteb y sma poreo dt certaif so n dimensions. These factors may be a reason of failure of comparison of diffusion rate samplen si differenf so t preparations one basi.th f so the mean pore radius ,above-mentionee eveth f ni d other factor- ta e sar ken into consideration. Other data on ferric'oxide gels (4) show also the major effect of pore size on the internal diffusion kinetics. b-Selectivity behaviour Before givin discussind gan selectivite date th gth f a e y behaviour of alumina and ferric oxide, it is important to show here the'anomalous' behaviour of Al-I heated at 350 C in respect of the variation of sorption of Na with pH. Figure 1 shows the variation of Borption of sodium from 0.1M solutions (NsjCl+HaOE )Al-n wito H Ihp heate different da t sorptioe th temperatures , C n0 increase25 o t p .U woulss a wit e , db hpB expected for an ion exchange process. For Al-I heated at 350 C, the sorption first decreases sharply and then increases with increase of pH. It has been proved that the sorption of Ha on Al-I heated at 35O C is an ion exchange process at all pH values. At 350 C, the exchange site presumable sar groupE yO s resulting from water sorptioe th y nb aluminium oxide frome thit da s temperature firse .Th t sharp decrease of sorption with increase of pH tb therefore surprising. A plausible explanatio thif no s situatio relatee b y n»a d with water structure effects and the porous texture,of the sample. At 350 0, Al-I is mainly compo-

79 senarrof o d w pores (4). Water structur expectes ei enhancee b o dt t a d the oxide/water interface anfl water structure effect more b ey sproaa - nouncenarrowee th n i d increasry porema d san e with increas, pH f eo resulting in a more ionisation of the surface protons that are water- structure makers. Therefore, at the relatively low pH values, Na ions wil excludee lb d fronarrowee mth increasesH rp porese th s .A , exclusion from the narrower pores increases and a decrease of sorption with the y occuincreassa H rp whef eo n this exclusion outweigh normae sth - lly -expected inoi-sie »orptiof o e a with i&ereaa Inothtl. pH f eo - expla- formatioe th natio e doubla b y f no nma e layer, whose potential increases with increase of pH , at the mouths of pores* This double layer may constitut barrieea r hinderin penetratioe gth porese ionth f n o n .si r savinFo g space, selectivity data wil givee lb n here onlr yfo Al- Al-IVd Ian selectivite .Th y coefficients were measure different a d t loading exchangere th valuee f so th d ssan given represen e selectitth - vity coefficient at 50$ loading of the Na-form of the exchanger with the enterinr alktdi ion. This coefficient, designate, ^ d Q herK s ea maquitequar e o yb o el t selectivite closth o et y constant s i (11 d )an anticipated to be a good representative of the overall selectivity displayed. The values of KQ ,- for the different alkali ion/ alkali ion exchanges in. different media are given in Table III for Al-I heated at different temperatures anfl Al-IV dried at 50 C. Table III shows that with the exception of the selectivity pattern displayed in 0.01M hydroxide solutions on Al-IV dried at 50 C, alumina show selectivitsa othee th ovea rl H alkalal ro yt iothee ionsth rr .Fo aluminas, a selectivity to Na over the other alkali ions is also displayed in all cases (4). The selectivity to Na over K and Cs may be attributed to the large effect of the latter ions in breaking the water structur probabld an e y strongea als o ot r electrostatic interaction of Na with the exchange sites. The selectivity to la over Li may be mainly due to the much higher hydration energy and hydration need of Li.

Table III

K« c for alkali ion exchange at 25°C on Al-I and Al-IV under different conditions. (l.!=alkale superscripth n e enterini th n i s io ioni t e g ,th ion).

P - 0 ,0 u i g 5 i s g 5 où . v S C - K - S C - K - l L - ^ , S S 3 3Ï g d I S 3 (C 0) W 0 0

50°C 0.0111 MOH 0.024 0.1? 0.16 0.15 0.94 0.94 0.88 50°C 0.10U MOH 0.240 0.18 0.13 0.08 0.72 0.61 0.44 250°C 0.01M UOH 0.025 0.16 0.1? 0.15 1.06 0.86 0.94 250CC 0.10M MOH 0.250 0.19 0.19 0.12 1.00 0.63 0.63 Al-I 50°C 0.10U MCI 0.029 0.028 0.032 0.048 1.14 1.50 1.70 (pH=8.3) 250°C 0.10M MCI 0.045 0.014 0.018 0.028 1.29 1.56 2.00 (pH=8.3) 350eC 0.10M «Cl 0.150 0.10 0.08 0.18 0.80 2.75 1.80 (pH=8.3) 350°C 0.10M MOH 1.01 0.35 0.29 0.32 0.83 1.10 0.91

D.01U UOH 0.085 0.62 1.03 1-26 1.66 1.22 2.03 0.10M MOH 0.342 0.1? 0.19 0.25 1.12 1.32 1.4? Al-IV 50°C O.Olil MCI 0.029 0.3? 0.50 0.57 1.35 1.14 1.5* 0.1011 MCI 0.117 0.09 0.12 0.13 1.28 1.08 1.44

80 chloride th n I e solutions affinite ,th d selectivityan mucs i he H o yt higher than in the hydroxide solutions. This seems to be due to a less ion exclusion from narrower formee poreth n rsi solutioas- na e th n .I rrower pores e selectivit increases th »C moro d t ovea an N e e rK o sdu yt profound water strucure effects, and that over Li increases due to a more role of ion hydration. With Al- 250°Cd Ian heateincrease 0 ,5 th t a d sitf eo e densit- yge nerally increase affinite smallea sth f o selectivitr yn o r io e th o yt erystallographic radius .takee b Thi indicato y nt s ma e small differences in the hyiration degrees (large dehydration) of the alkali ions inside Al-I. Thiease sth e doee witb seet so no hmt Al-IV. The increase of the heating temperature of Al-I f som 50 to 250°C leads to the increase of surface area from 27 to 190 m /g, the increana of total pore volume (porosity) from 0.08 to 0.54- c.e,/g, amd the decrease of the mean pore radius from 57 to 36 &. Xt 350 C, the surface area and porosity continue to increase to 437 m /g and 6.87 c.c./ whil meae eth n pore radius remains almossame th e t tvalua O 4 f eo In the hydroxide solutions, the site density is at 50 and 250 C. The increase of heating temperature from 50 to 250 C leads to generally lower selectivit case 0.1f th eo n y i M hydroxide solutions and to almost ao change in the case of 0.&1M solutioas. In the former case, this may be due to the effect of an increasing porosity, outwei- ghin effece gth decreasinf to g pore lattee sizeth n .rI o easetw e ,th effects balance each other probably because narrower pores that become too narrow at 250 C are involved in the exchange. For the sample heated at 350 C, the selectivity is in general lowest, which is presumably due to the much higher porosity of this sample, outweighing the effects of a higher site density (Table III) and acidity (Fig.1). In the chloride medium selectivite ,th y seem increaso st e with increase of heating temperatur Thi. e C probabl th s s 0 i o t 2^ ee o t froy du 0 5 m decreas porf o e e siz dominane ewhic b , tVv o t y » hinvolvemenma e tdu f to much narrower pores in this medium. With Al-II and Al-III. the results O) indicate that in the conditions favourable for comparison, the decrease of pore size increases the selectivity.

Û 50 s t-' «0

7 « 9 10 11 12 U EquilibriuH mp Fig.f 1 »Variation with pH of V. uptake of Na*fromO-1M Na* solution Al-n o s l heate t differena d t temperatures I0m!/= m ) V/ g (

The capacity-pH curves (not given here for the sake of brevity) show that the acidity of the samples decreases in the order: Al-IV •> A1-I1I>Al-I>Al-II. Table III shows that the selectivity to Na is considerably hip.her in Al-I than in Al—IV. Comparisons for other alu-

81 minas (4) show also that this selectivity is higher in the samples of lower acidity. Thus, the selectivity of alumina to Na increases not only wit decrease hth porf eo e sizt alsebu o wit decrease hth sitf o e e acidity, Fe-I and Fe-II have shown the same kinetic and sorption behaviour

in agreement with their simila poroud a& r A X-rays DT textur, ,IR e data f (4)e selectivit.Th y dat Fe-In ao Fe-IId sa I I under different condi«. tion givee ar sTabln ni , eIV

Table IV

r alkalfo Kc n Qexchang io i t 25°a e Fe-In Co Fe-IId an I I under different conditions. '(M»alkal ti.n i en i superscripio ion e ,th e enterin th s i t g ion).

v bo u l a *»> p . o a f o o bo a -H.-HO-H -H. ... „ 0} 4* Ot r-l iH -P OUÛ v^i TfK V^® V^ v*US „US o S I "30 3 §,> T*a *Na *Na *Li *K *Li

50°C O.ODi MOH 0.110 4.40 6.80 9.30 1.55 1.37 2.11 50°C 0.1011 MOH 0.310 8.60 5.80 3.40 0.67 0.59 0.40

j,e_IZ 200°C 0.0111 UOH 0.091 1.20 1.40 1.90 1.17 1.36 1.58 200°C 0.10U MOH 0.220 1.48 0.63 0.49 0.43 0.78 0.33 400CC 0.0111 MOH 0.058 1.60 2.80 4.10 1.75 1.46 2.56 400°C O.lOli MOH 0.190 1.68 0.48 0.38 0.29 0.79 0.23

50°C 0.01U UOH 0.104 p.80 7.90 10.10 2.10 1.28 2.66 50"C 0.10M MOH 0.410 7.00 6.30 4.10 0.90 0.65 0.59 Fe-III 200°C O.OUU 130H 0.090 1.70 2.20 2,?0 1.29 1.23 1.59 200°C 0.10M MOH 0.32 1.40 1.00 0.62 0.71 0.62 0.44 400°C 0.0111 UOH 0.056 3.90 3.80 4.70 0.97 1.24 1.21 400°C 0.1011 MOH 0.170 2.64 1.40 1.19 0.49 0.85 0.42

Tabl V showeI heatinsl al tha t a tg temperatures increase ,th f eo site densit botr yfo h Fe-I Fe-IId Ian I leadlarga o st e iacreaef eo the affinit smallea f o selectivitr n yo r io bar e eth radiuso yt . This probably shows tha e alkaltth i ionlargele sar y dehydrated. Selectivi- ty reversals occur smallea f o , wherebn r io bar e eyth radius becomes preferred at the higher site density. This is probably due to the fact that more weakly acidic sites are involved at the higher site density. Before comparin selectivite gth y sequences oust , i mentionee tb d that Fe-II has a porosity of 0.14 0,36 and 0.24 c,c,/g and mean pore radius of 23, 68 and 69 A at the heatint g temperatures 50,200 and 400 C, respectively Pe-IIIr .Fo , these value 0.29e sar , 0,5 0,1d 3aa 9 e.c./g respectively, A 0 8 d aa 5 acidite 6 . Th , an Fe-IIf 40 dyo highes Ii r than thaheatin l Fe-If al o t t Ia g temperatures acidit» th C f 0 yo 20 t .A

both samples tis almost the same as that at 50 C up to a capacity of about 0,1 meq/g. At higher capacities, the average atlëity is apprec- iably less at 200 C, At 400 C, the acidity decreases considerably for both samples. In 0.01M solutions site ,th e densit ïe-f yo Pe-II d Han I dried at 50 C is almost the same. In these conditions, the selectivity displayed for the ion of larger bare radius is generally somewhat higher for Fe-III. Since both the porosity and pore sice are higher for

82 Fe-III, this may b« Aue to the higher acidity of this sample. For both »amples heated at 200 C, The capacity in the 0.01M solutions differs little from that at 50 C. Besides, the acidity is, as mentioned above, the same» Table IV shows that in these solutions, the selectivity of both samples becomes e mucThi, th C clearl s h si o 0 t lowe 20 e t ydu ra larger porosity and pore size at this heating temperature. In 0.1M solutions, the site density and acidity at 200 0 are considerably lowe rbotr thafo h nC samples 0 thos5 t e a thes n .I e solutionse ,th selectivity sequences are largely different at these two heating temperatures and with the exception of the Li/Ha exchange they show a generally muc hsmallee higheth f o r r n bar io affinit ee radth o y-t ius at the higher heating temperature, which is presumably due to stronger electrostatic interactions as a result of the lower acidity* However, the lower site density and higher porosity and pore size at 200 C reveal theaselves mainly in the much lower selectivity for Li over Na. The selectivity pattern displaye 0.1n i d M solution Fe-Iy sb I Li>Na>K^Cs: iß C O heate2O t a d . This sequenc reinforces ei e th n di same solution at the heating temperature 4OO°C where, compared to the temperaturacidite th porositd , C yan 0 lowee e20 yar r whereae sth site densit pord yan e sizalmose ar esaaee th tdecrease .Th acif eo - dity reinforces this pattern, showing a preference for the ion of smallea r crystal]ographie radius increasiny ,t role electrof gth eo - stativ interactions, wherea decrease sth porositf eo y reinforcet si by increasing the selectivity. 5-Sorption of Uranium and Thorium by Hydrous Tin Oxide and Ce-II Uraniu thoriud an m e sorbear m hydroun o d n oxidsti e from acidic solution aca2 df o s0.01 (7,52)H p M a urany t . A thoriud lsa m salt solutions, the capacity of this oxide is 0.36 and 0.44 meq/g for uranium froaqueoue mth s nitrat sulphatd ean e media, respectively 0.2d ,an 6 and 1.1 meq/g, respectively, for thorium. In the aqueous medium, the U/Th separation factors are rather poor. However, these factors ea» become very high in the mixed solvents. Thus, for example, for O.OO1M nitrate solution # aceton80 0.01n d si an e M nitric uranr acidfo - ,K ium is about 600 and for thorium it is about 25 (V/m-25 ml/g). For O.O01M sulfate solutions in 20# acetone and 0.1N sulphuric acid, K, for uranium is about 320 while for thorium it is about 1 (V/m « 25° ml/g). Schemeh separatioUA r sfo n have been suggested (7,32e th n )o basis of these and other results. Relatively large uranium and thorium sorption capacities are also displaye thoriue th » 3 Ce-IIy mdb H p capacit t .A y from 0.01M nitrate solution is 0,9 meq/g, and at pH 3.2, the uranium capacity from the same solutio 0.4s ni 4 meq/g. However, while uraniu thoriud an m m seem sorbee tb o hydrouy n exchangdb io n n oxida sti y eeb mechanisme ,th sorption mechanis cerin i m a quitseeme b o est complex. Beside rate sth e of attainmen equilibriuf to vers mi y slo cerian wi , requiring more than -i month. Nevertheless e majorit,th f exchangyo e takese placth n ei first hour. Experience shows that many hydrous oxides do not sorb any significant amounts of alkali, earth alkali or rare earth ions from acidic solutions. It is therefore anticipated that hydrous oxides can be used of uraniu thoriud an m m separation from many other elements.

ACKNOWLEDGEMENT - The authors wish to thank deeply the International Atomic Energy Agency for financial supprt under the Research Contract No. 2716/RB, 2716/R1/RB, 2716/R2/RB.

REFERENCES 1.A.MPHLETT,C.B.,"Inorgani Exchangersn cIo , McGraw-Hill York(l964)w ,Ne . 2.RUVARAC,A.."Inorgani Exchangn cIo e Materials",Editor: GLEftRl-'IELD.A., Chap.5, CRC,Florida(l982). 3.MISAJC,N.Z.,MIKHaIL,E.M.,J.appl.Chem.Biotechnol.28(1978)499. 4.KISAK.N.Z..et al.,Progress reports submitted toTfcEA. .Research Con- trac 2716/RB,2716/R1/R. No t B ,2716/R2/BB(1981-1984). 5.BRITZ.D.,NANCOLLAS.G.H.,J.inorg.nucl.Chem.31(1969)3861. 6.MISAK,N.Z.,GHONEIMÏ,H.F.,J.Chem.Tech.Bioteehnol.32(1982)709. 7.MISAK,N.Z.,SA3AMA,H.N.,EL-NAGGAR,I.M.,Chemica SerTpta 22(1983)74.

83 8.HALLABA,E. ,MISAK,N.Z.,SAIAMà,H.N,,Indian J. Chem.11(1973)580. 9.MISAK.N.Z .,SALAMA,H .N.1Inorg.Nucl.Chem.Lettersl1Tl975)559. 10.EISENMAN.G. Biophys. — J.2(1962)259. 11.REICHENBERG,Gt . "Ion ExeEange"»Editor:MARINBKY.J..Marcel Dekker. New York(1966)227. 12.SHERRYjH.S.,"Ion Exchange"»Editor:MARINSKY,J.,Marcel Dekker,New York (1969)89. 13.TIEN,H .T.,J.Phys.Chem.69_(l963)350. 14.CHURMS.S.C..J.S. JLfr.Chem.InBt.19(1966. )98 15.HEITKER-¥IBGUIH.C.^.LBtU-YARON,AT. J.inorg.nuel.Chem.28(1966)2379. 16.GEREVINI,T.,BMOGLIaBA,R.tEnergia Nucl. Milano 6(19597^39. 17.LEONT «EVA.G.V.^VOL'KHIN^.V^Izv.Akad.irauk SSBR",N*org.Mater.^ (1968)728). 18.FLOOD,H.,DiBeuBsions Faraday Boe.7(19^9)180.

21 IAHRIAND , B eAUESO. H ,0 ., J. inorg .nuel .Chêm. 33(1971 ) 2229.

22.DONAIJ)SON . tPTmLER,MJ) ,J t J.inorB.nucl.C5«m.32(1970)1703, . .J . 23.MISAK,N.Z .t MIKmiL.E.M. lJ.inorg.nucl.Chem.43TT981)1663. 24.MISA.K,N.Z. .MTKmilL.i.M., J.inorg.nucl.Chem.^Tl981)1903. 25.GHONEIMY,HiI.f M.Se.. Ain-Shama Univareity, Cairo(19?8). 26.ROSSEINSEYtD^.jChem.Rev.65(1965)467. 27.NAHCOLIAS,d.E. JaEH),D.S.,77inorg.nucl.Cham.31 (1969)213. 28.RUVARA ,0?RTANJ.M.. CJl , J.inors.nucl.Chem.3i(1972)3893). 29.MISAK.N.Z., GHOHEIMY,H.P.,Colloid d Burfacean s B 7(1983)89. 30.MISAK,N.Z., GHOKEIMY,H.?.,J.Chem.Teeh.Bioteehnol732(1982)893. 31.HELFFERICH,F.1"Ion Exchange",MoGraw-Hill,New YorkrT962). 32.MISAK,N.Z.,SAIAMA,H.N.,EL-NAGGAE,I.M.,Chemiea Bcripta (in press).

84 INORGANIC SORBENTS IN SPENT RESIN INCINERATION - THE ATOS PROCESS

C. AIROLA, A. HULTGREN Studsvik Energitekni, kAB Nyköping, Sweden

Abstract The spent resins from the operation of nuclear power plants form a dominating waste category, by activity contents and by volume. Volume reduction by incineration is technically feasible but introduces problems by activity release in the off-gas and in ash solidification. By a chemical pretreatment that transfers radioactive nuclidés to inorganic sorbents a considerable improvemen obtaines i te incineratio th n i d n route. The paper summarizes efforts at STUDSVIK on development work along this line over the last years, resulting in the ATOS process. Included are results frot activitmwe y transfer experiments, incineration of pretreated resins, and ash immobilization. The overall conclusio tha s ATOi e n tth S process offers clear advantage o cleat s nsa off-gases, significant volume reductio d stablan n e products.

1. INTRODUCTION Spent ion exchange resins from power reactor operation contai ne tota th mor lf eo radio% tha 5 9 n- activity of wet reactor wastes. Cementation and bituminization are the two methods applied in Sweden e immobilizatioth r fo w no thesf uo npo t e resins, both methods leading to final waste volumes by far exceedin volumoriginae e th th g f eo l resins. The reactor wastes in Sweden will be disposed in a rock cavern under the sea bottom of the Baltic, 1 km from the shore. In order to minimize disposal costs it is desirable to reduce the volume of the wastes before disposal /!/. One process develope discussed an d Sweden i d n for volume reduction of spent resins is the PILO process PILe /2/th O . n I process resine ,th e sar eluted and the eluted activity (99.9 % of Cs-Sr) is sorbeinorganin a n i d c material, suc zeolites a h e .Th eluted resi they nincineratee ma n b possibly, or d , disposed without further treatment. ATOe Inth S proces chemicasa l pretreatment enables tha resine t th incineratee sar d wite hth activity retaine e ashe th wele s a n sth mosi s dl a f to sulphur from the cation resin.

85 2. THE ATOS CONCEPT acronyn ATOa s Activiti r S fo m y Transfer from Organic resins to inorganic Stable solids. Process flowsheets for ATOS and PILO are shown in Figure 1. In the ATOS process the spent resins, beads or powder mixee ,ar d simultaneously wit elutinn ha g agent and an inorganic sorbent. The slurry is dried and incinerated/calcined under controlled conditions

PILO ATOS

SPENT SPEN1

SORBENT

FIG. 1 Process flowsheets for PILO and ATOS e incineratioTh n residue mixtur,a ashf eo , inorganic sorbent and metal sulphates, may be immobilized in different ways. It may be sintered or melted with additives to a stable end product, or solidified by cementation, depending on local conditions at the site where ATOS wil appliede lb . The main objectives of the ATOS process are the following: transfe cesiuf ro othed man r radioactive elements froorganie th m c resininorganin a o st c sorbent by a pretreatment process significant volume reductio waste th y eb f n o incineration of the pretreated resins easy adaptation to already existing processes for incineration and/or solidification simplified cleanu off-gasef po s a minimum generation of secondary waste.

86 treatmene Th t stage reaco t s h these objectives include a simple transfer of resin-water slurries from spent resin storage tanks to a mixing vessel, e additioth pretreatmenf no t chemicals, removaf lo excess water after treatment, transfer of the dewa- tered resin to a dryer/incinerator, incineration under controlled redox conditions thereby ensuring retainment of volatile activity, mainly cesium, and containment of most of the sulphur as metal sulphates.

3. PROCESS DEVELOPMENT /3-13/ 3.1 Pretreatment One main problecombustioe th f mo radioactivf no e spent resine releas volatile th th s i sf eo e activity, principally Cs-compounds. Studies were thus initiated at STUDSVIK assessing different way whicy sb h Cs-compounds coul e retainedb d in the incineration residue during combustion where different additives such as zeolites and other inorganic sorbents were tried. In the drying-calcination process cesiu trappes e structuri m th n i d Na-Alf eo - silicate systems, which gives a good retention of cesium during this stage of the process. Studies of the thermodynamic stability of various Cs-compounds under typical redox conditions encountered in a furnace were also conducted. Bench scale tests with different eluting agents such as oxalic acid, Na-oxalate, tartaric acid, Na- tartrate, citric acid, Na-citrate, phosphoric acid, various phosphates and various formates showed that elutine amountmeq'4 th 2- f f so f o g so l agenm r tpe resin is an optimal amount leading to almost exhaus- tive activity elution from the organic resin. The different sorbents tested include Na-titanate, bentonite clay, mordenite, Ca-phosphate- ,Zr oxide, Zr-phosphat wels a es speciall la y tailored sorbents like OXTI, PHOZI PHOMI, RA POLYA, XA , M N PERTITANIC and FE-CO-K (Applied Research Sprl, Belgium). The performanc e abovth ef eo mentione d sorbents showe PERTITANIe bese th dt alsth e tbu e oth b o Ct most expensive one. Second choic Na-titanats i e d ean bentonite clay and mixtures/blends of these. As Na- titanat gooa s di e sorben Co-6r bentonitd fo t 0an e claCs-13r fo y standar7a d sorbent consistin% 0 8 f go Na-titanate and 20 % bentonite clay was used as a reference retentioe .Th n coefficien thif o s, tK mix - ture was for Cs, K,=200 and for Co, K,=600 (Kd=Csorben Solution' t ml/n 'i ^' The result of these studies led to the develop- procesmene a conditionin th f to r fo s f spengo t resins prior to combustion. A simplified process flowsheet is shown in Figure 2. The spent resins are transferred batchwise to a mixing vessel. From each batc sampla h withdraws i e n

87 ELUTING INORGANIC CALCINATION AGENT SORBENT ADDITIVES

TO OFF-GAS CLEANING SYSTEM

PRECONDITIONED RESIN STORAGE TANK

Simplified flowsheet for the ATOS pretreatment process FIG. 2

and analyzed. Based upon this analysis eluting agent, inorganic sorben calcinatiod tan n additives/sulphur retainer addede sar . After treatmen e conditionetth d resi transferres i n storaga o t d e tank. From this tank the resin is fed to the incinerator.

3.2 Incineration markeA t survey mad n lati e e 1980 showed that ta that time no proven commercial incineration system for resin s availableswa e developmen.Th sucf to ha system has been in progress in the USA, where e.g. the Aerojet compan demonstrates ha y d pilot plant scale incineratio exchangn io f no e resins /ll/. At STUDSVIK incineration tests were performed, at first in laboratory scale to evaluate and optimize the pretreatment process. Later a prototype resin incinerator was built and tested on the kg/h scale. Studieincineratioe th f so n proces y thermasb l analysis showed that: untreated resins temperat decomposa r ai -n i e purn I e. oxygeC gasee 0 nth 30 so turet p su ignite explosively at about 300 C if dried resin usee sar d t untreateifwe d resin usee ar sd volatile decomposition products are distilled with the water. Ignitio oxygen i n takew nno s plac about a e t 400 °C t pretreatewe d resins ignite explosiveln i y oxygen at about 260-280 C and are fully com- busted at about 500 C

88 unloade untreated an d d resint ignitno o sed even in oxygen decompost ,bu e slowlabouo t p tyu . 700°C Bench scale furnace tests were then made to verify the above mentioned observations, and results from thes e show ar ethin Tabln I i n s. 1 etabl e results from combustion in normal air and in oxygen are compared. The reason for using an oxygen atmosphere was to get a clean and complete incineration. Cesium may also be retained as Cs^SO. which is formed in the presenc sulphuf eo r under oxidizing conditions. e seeb nn Fro e tablca tha mth t i te during combustion of untreated resins in oxygen some 7 % of the activity are emitted and that this figure can be lowered to about 0.1 % by preconditioning of the resin.

TABLE I Activity measurements from incineration tests

Resin Activity Emission treatment measured % of initial activity Oxyge tina .r n Ai atm. Untreated Total 6.6 0.02 Cs-137 5.5 0.04 Cs-60 6.8 0.02 Untreated + Total 1.6 0.004 inorganic Cs-137 3.3 0.004 sorbent Co- 60 1.3 0.006 Pretreated Total 0.4 0.02 Cs-137 0.6 0.05 Co-60 0.4 0.02 Pretreate+ d Total 0.04 0.06 inorganic Cs-137 0.08 0.02 sorbent Co- 60 0.2 0.1

Combustio r showeai n variatioa di n e th f no emission of activity between 0.02-0.06 %. These w tha e accuraclo tth o levels measuremenf e o y sar f to w amountthlo e activitf so y usee somewhaar d t uncer- tain. The tests also showed that good combustion characteristics wit higha h combustion efficiency indee s obtaine dwa oxygen a n i dn atmosphere whereas the amount of unburned resin was appreciable after combustio plain ni n air. A prototype pilot plant resin incinerator witha capacit abouf o y kg/1 t h (dry weight basis s desig)wa - ned and built on the experiences obtained through the bench scale tests. The pilot plant incinerator is shown in Figure 3. e incineratoth o t d e resife r Th s eithei n r through a screw feeder (powdered resins) or through a funnel feeder (bead resins). Incineration takes place oheatea n d turnable gratin oxygen a n i gn enriched atmosphere. The off-gases pass through an afterburner to a condenser and from there through a HEPA filte activitd an r y monitorin e laboratorth o t g y stack.

89 FIG. 3 Pilot plant incinerator

90 e resultTh s fro firsa m t serie f inactivso e tests are summarized in Table II. From this table it is seen that a considerable volume reduction e achievedfactob y r ma off-gasee .Th s contain normally about 3500 ppm S02. This is of the same order of magnitude as reported by Aerojet. By adding suitable calcining additives preliminary tests indicat conten_ ereducee S0 b tha e n tth beloca o t d w 100 ppm. Further optimization of the incineration process is required, but the results so far are quite promising.

TABLE II A summary of the inactive test runs

Range Average

Resin feed 0.4 - 0.7 0.5 kg/h Incineration 700 - 850 800 temp/ C Incineration 50 - 170 80 time/min Incineration residue/ml 80 - 250 150 residue/g 40 - 170 85 Volume reduction factor by vol 4-19 11 by weight 5-21 14 Offgas comp. oxygen/% 7-28 17 , SO /ppm 1000 -5000 3400 0 75 0 -25010 0 T NC /ppm

By additio f appropriato n e calcination additivee th s , contene off-gaseSO th f to s decreasem pp beloo t d0 10 w

After modificatio piloe th tf no plan tseriea s of active test runs will be performed. A preliminary test to incinerate radioactive resins pretreated according to the ATOS process was e fulth mad ln i escal e waste incineration facilitt a y STUDSVIK. 25 kg of spent resin containing 20 mCi Cs-137 and 5 mCi Co-60 were mixed with medical radioactive wast d incineratedean . Activity emissions measured showed that only 0.6 ppm Cs and 2 ppm Co were emitted. Traces of unburned resin were found in the ash which stresses the point that the combustion air shoul enrichee db oxygenn i d .

3.3 Ash solidification e asheTh s from resin incineratione b hav o t e solidifie r transporfo d disposald tan .e Oneth f .o methods use Sweden i d e solidification th toda r fo y n

91 of various wastes including spent resins is cementation. This method is proven and accepted by the authorities. A main drawback of this method is the waste volume increas factoa least a y eb f tro two. In orde maintaio rt volume th n e reduction advantage achieved by incineration of the resins, the residue should be solidified with a minimum volume increas e solidificatioth n i e n stage Figurn e I . th 4 e change volumf o differenso t e edu t treatmente sar shown. From this figure it is seen that cementation of the incineration residue still provides an overall reduction of volume by a factor of about 5, compared to the initial volume. Screening tests made at STUDSVIK show that the resin e solidifieasheb y sma y sinterinb d temperat a g - tures about 1100-120 e addinTh f suitabl. o C g0 e sin- tering aid y somewhasma t reduce this temperature.

DIRECT: CEMENTATION BITUMINIZATION

CEMENTATION INCINERATION

MELTING

FIG. 4 Volume change differenn i s t treatmentf so spent resins

The tests conducted at 1200 C showed a Cs- emissio voluma d an abouf e no % reductio 5 t n factor of about two. Lowerin e sinteringth g temperaturo et 1100 °C showed that the Cs-emission decreases to e additiveth t bu % sabou require5 0. t d prohibit further volume reduction.

92 Tests have also been made to solidify the ash by cementation. The results so far show that a compara- tively good obtainee b soli n ca d dcontenh as wit n ha t of about 35 weight %. The compressive strength of this solid was 12 MPa and the volume increase factor only 1.3 in the cementation stage. There are two problems encountered when cementing the ash metae :th l sulphate e carboth d nsan content. e e solveformeb Th n usiny db ca r g sulphate résistent cements or special cements like the US Gypsum cement. The second problem should be solved by an optimization of the incineration process, with a sufficiently long residence time under oxidizing conditions. Work is now in progress at STUDSVIK to optimize solidification of the incineration residue by cementation. Result expectee ar s d durin e fallth g , 1984.

4. CONCLUSIONS A review of the results of the STUDSVIK studies supports the conclusion that incineration of resins pretreateATOe th Sn i proces s feasibla a d s si e means to achieve a considerable volume reduction of the spent resins, while at the same time minimizing secondary waste generatio activitd nan y emission. The use of appropriate inorganic sorbents in the pretreatment process ensure completsa e combustion and minimized activity emissio e subsequenth n i n t incineration step, provided that oxidizing combustion conditions are maintained. This means that the off- gas cleaning system can be relatively simple, thus reducing process and operating costs.

la Eluting agent Ib Sorbent 2 Mixer-dryer 3 Buffer tank 4 Metering screw 5 Furnace conveyoh As 6 r 7 Ash storage tank B Afterburner 9 Scrubber 10 HEPA-filter 11 Fan

FIG 5 . Layou ATOn a Sf to pilo t plant

93 o secondarAt s y waste generation promising results have been obtained by the addition of sulphur retainer incinerator/calcinere th o st 2 emissioS0 . n is below 100 ppm, but more work is needed in order to bring this figure further down. As the overall volume reduction capability decreases with efficient sulphur retentio compromisa ne foun b o dt betwees eha n volume reductio sulphud nan r retention. Pilot plant studie completa f so e ATOS facility have also been made and a layout of such a facility is show Figurn e cosi n Th f suc to . e5 planha s tha been estimate o about d capacita MSE0 1 tt a K f o y aboum /yea 0 1 tf spenro t resins.

REFERENCES / HULTGREN/! , "Treatmen,A. spenf t o exchang n tio e resin practices- developmentd san Sweden"n si , Studsvik Technical Report NW-83/428 (1983) Contributio e IAEth A o nt Thir d Research Coordina- tion Meeting, Cairo, Egypt Feb-8 ,2 4 March, 1983. /2/ HULTGREN, A., AIROLA, C. , FORBERG, S., FÄLTH, L., "Preparation of Titanates and Zeolite Theid san r Use Radioactivn i s e Waste Management, particularly in the Treatment of Spent Resins", Studsvik Technical Report NW-83/475 R 83-23T S 1983y ,,KB Ma . /3/ AIROLA HULTGREN, ,C. , "Fina,A. l treatmenf to spent resins witATOe hth S proces studa s- f o y principles" (in Swedish), Studsvik Technical Report NW-82/237 (1982). /4/ RINGBERG Unpublishe, ,H. d results. /5/ RINGBERG , "Incineratio,H. n experimentr ai n si and oxygen with spent resins under various conditions" (in Swedish), Studsvik Technical Report NW-82/312 (1982). /6/ RINGBERG , "Tria,H. l incineratio ATOf no S pre-treated resins in the STUDSVIK incineration plant" (in Swedish), Studsvik Technical Report NW-82/383 (1982). III OLLINEN AIROLA, ,M. , "Solidificatio,C. n experiments with ashes froATOe mth S processn (i " Swedish), Studsvik Technical Report NW-83/524 (1983). /8/ BRODËN, K., NIRSCHL, N., "Results from trials with incinerated resin ashes solidifien i d concrete Swedish)n (i " , Studsvik Technical Report NW-83/539 (1983). /9/ AIROLA, C., "ATOS-AMOS plant, technical description Swedish)n (i " , Studsvik Technical Report NW-83/488 (1983). /10/ HULTGREN , "Developmen,A. ATOf to S technique. Status report September 1983 Swedish)n "(i , Studsvik Technical Report NW-83/545 (1983). /ll/ LINDGREN , "Incineratio,E. e resinf no th n i s ATOS pilot plant. Results from trials performed during summer 1983 Swedish)n "(i , Studsvik Technical Report NW-83/547 (1983) .

94 /12/ MICHLINK, D.L., MARCHALL, R.W., TURNER, V.L., SMITH, K.R., VANDERWALL, E.M., "The Feasibility of Spent Resins Incineratio Nucleat a n r Power Plants", Waste Management '83, Tucson, Arizona, 1983. /13/ PATEK, P.R.M., TSYPLENKOV, V.S., HULTGREN, ,A. POTTIER, P.E., "Treatmen Spenf to Exchangn Io t e Resins". International Conferenc Radioactivn eo e Waste Management, Seattle USA, y ,16-20WA ,Ma , 1983. IAEA-CN-43/455.

95 PRECIPITATED MAGNETITE IN ALPHA EFFLUENT TREATMENT

K. HARDING Atomic Energy Establishment Winfrith, UKAEA, Dorchester, Dorset, United Kingdom

Abstract

Iron presen a fortuitou s a t s componen f moso t t nuclear plant effluent e precipitateb n ca s s magnetita d e after suitable chemical treatment. The spinel crystal lattice of the magnetite is abl o incorporatt e a wide e rang f foreigo e n ions which includes many important radioactive species.

Provided the operating conditions are well chosen the precipitated magnetite provides an excellent scavenger for many radionuclide d givean s a denses r more granular solid produco t t be separated froe effluenth m t thae morth ne visual scavenging agent ferric hydroxide.

e conditionTh s under which magnetit e formeb y n b dca e precipitatio s propertieit d an n e presentear s d comparean d d with those of ferric hydroxide. Because magnetite has a high magnetic susceptibility its separation by high gradient magnetic separation become a spossibilit d datan ya relevan o thit ts i s presented.

1. INTRODUCTION

Nuclear plant effluents frequently contain traces of_j.ron typically at concentrations in the region of 10-200 g m in association with radionuclide contaminants which vary from plant to plant. In reactors these are., mainly activated circuit corrosion products Co, Fe, Mn and Zn with only minor amounts of fission products and alpha active nuclides. In fuel reprocessing plant^ they, „are reversed, fission products u R predominat , d Cs signifian particularle , - Sr y cant amount f alpho se b a y activma m A e d nuclidesan u P e i , present.

H adjustmenOp n t these effluents for a precipitatm f o e ferric hydroxide which also carries with it many of the trace radioactive species as co-precipitates or by adsorption. Because these incorporation mechanism H dependentp e ar se th , efficienc e variouf removao th y r sfo l active species varies y givewidelan d alsr nan yfo o specie s consistent t resultno e ar s eas o achievet y . However satisfactory remova f thio l s ferric flon producca e e significant decontamination factors (DFr fo ) such effluents. Unfortunately ferric floe is notoriously difficult to handle on mechanical filters being gelatinous, thin films quickly give ris o higt e h pressure drop n conventionao s l filters resulting in low capacity and causing problems with back washing. Because of this effect settling only is often resorted to, the plant then becoming large and expensive and can often give rise to variable DFs because of small particle carry over.

97 n efforIa n o improvt t e overalth e l efficienc f efflueno y t decontamination with particular emphasi n reducino s g coste th s application of High Gradient Magnetic Separation (HGMS) to ferric floe filtration was investigated [1]. This work concluded that although HGMS would give process improvements it was not economically viable because the magnet system needed would be relatively large. What was required was a precipitate which had a higher magnetic susceptibility than ferric hydroxide and so reduce the size of the magnet, provided it was at least as effective as ferric hydroxide at removing activity from solution .

The obvious choice was to precipitate magnetite from solution instead of ferric hydroxide. That magnetite can be produced from cold solution f ferrio s c ferrous iros beeha n n demonstrate y othee resultb dTh r , workers 3] showe , [2 s d that the precipitated magnetite was strongly influenced by the presence of other ions in solution, some like sulphate being beneficial other like silica preventing its formation all together excep t higa t h temperatures.

However, the potential value for effluent clean up of precipitatin a densg d stronglan e y magnetic solid fro n iroa m n containing waste stream was considered to be worth detailed investigation. This paper reviews the progress made so far with the identification of the boundaries within which magnetite can be formed from cold solutions, its properties and potential for effluent clea p operationsu n .

2. MAGNETITE - METHODS OF PRODUCTION

Magnetit s ferroui e s ferric oxide, havin e chemicath g l formul e simples th a alarg s f i FeO.Fe^o te d familan , Of o y mixed oxides having the same structure of cubic close packing spinee knowth s la n structure e generiTh . c terr thesfo m e mixed e ferritesoxideth s i s . Many examples exis n minerali t n i s whic e divalen nickels th hi n io t, copper, cobale th r zin o td can trivalen s aluminiui n io t r chromiumo m s rari r thes fo et I e. minerals to be fully stoichiometric, ie all the ferrous iron to be replace y nickeb d r coppeo l d consequentlan r y these minerals presen a bewilderint g rang f compositiono e d propertiesan s .

Many ferrites have interesting and useful electrical and magnetic properties and are now prepared commercially on a large scale. These include compound t foun no snaturen i d .

In general these synthetic ferrite e preparear s y b d precipitation of mixed metal hydroxides by the addition of alkal carefullo t i y formulated mixed salt solutions, followey b d careful calcinin d finallan g y sinterin o product g e regular sized crystallites with the required mixed cation array.

r applicatioIou n e requirw n o product e a similae r ordered arra f o ferrouy d ferrian s c ions with relatively small proportion f boto e ferrous th h d ferrian s c ions being replaced by the radioactive species also present in the effluent stream. Provide e siz f th thesdo e e ion s withii s e acceptablth n e range for the spinel structure they will be locked into the crystal

98 D onl lattic CE e e removeb precipitaty th d an f ei d s fulli e y dissolved, unlike the situation with ferric hydroxide where such cations are often held by adsorption and are released by relatively small change n solutioi s n composition.

Reporte successfuth f o s l remova f vero l y large ions such s a cadmium d leamercuran d y b incorporatioy n into ferrite precipitate ] 5 indicate , [4 s s tha a vert y wide rangf o e radioactive species could well be held in this form. More recently work at Rocky Flats [6] has demonstrated that plutonium and americium e ferriteb prepare n ca sd under appropriate conditions.

There are three methods whereby a solution can be made to yiel a mixed d oxide precipitate whic s capabli h f conversioo e n to a ferrite structure in which the predominant material is iron.

e ferrou ) th Iro (a n i ns stat d ferrouan e s hydroxids i e precipitated followed by partial oxidation usually with air.

(b) With iron in either the ferric or ferrous state an appropriate amount of the counter valance state iron is adde o e t givnecessardth e y stoichiometrr fo y magnetite production, followed by pH adjustment.

(c) With iron in the ferric state partial reduction is carrie t followe ou dH adjustmen p y b d yielo t mixea d d oxide precipitate. e effluentA surveth f o y o whict s h this process coule b d applicable showed that the iron was predominantly present in the ferric statd thereforan e) woul(c e b der o metho ) (b d applicable. Method (b) involves the addition of further iron to the system thereby increasing the volume of solid waste to be handled and since this is highly undesirable, this method has not been investigated. Our work is being concentrated on method (c), ie the reduction of part of the iron present to yield a ferrous to ferric ratio of 0.5 ±0.1.

Reduction of ferric ions to ferrous can be accomplished by a numbe f reagento r d alsan so electrolytically. Howevery an , process designed to treat effluents needs to be both simple and economica d thereforan l e only sulphit d hydrazinan e e have been studied in detail so far in the programme. However dithionite (sodium hydrosulphite) is an interesting reagent suggested by Streeter et al [7] because in addition to its use for reduction in n solutioundeca t ri n some conditions convert ferric hydroxide to magnetite direct in a process analogous to the oxidation of ferrous hydroxide.

3. EXPERIMENTAL

1 Effect3. f Iroo s n Concentratio d Dissolvean n d Oxygen

A reducing agent, .suitabl r thifo e s process shoul e stronb d g n acceptabla t a e n eF a enougrato r t fo eo reduc t h e F e efficient process, but not strong enough to react with other

99 oxidising agents particularly nitrate which is always likely to be present in typical effluents. The presence of dissolved oxygea complicatin s i n g y likelfactoan n i yr process since o known ther e n ar ereagent e F s o t whic n reducca h e F e in the presence of oxygen and therefore it must be removed by a prior deoxygenation operation, ie nitrogen sparging or reduced along wit e ironh th par f .o t -3 At high iron concentrations, ie above 350 g m the amount of sulphite needed to react with the dissolved oxygen is relatively small, approximatel e totae thirth on ylf o dneeded , and doet seriouslno s y interfere wit e controe iroth hth n f o l reduction reaction. Jftowever, as the iron concentration is reduced below 100 g m the reduction of oxygen becomes the major sulphite consumer and control of the ferrous reduction reaction becomes more difficult .e concentratioth Als s a o f o n iron in solution is reduced the reaction with sulphite becomes slower .and consequently the position where the optimum e /F Ferati s reachei o s mori d e difficul o determinet t . Clearly some lower concentratio f ferrio n c iron wil e reacheb l d where satisfactory formatio f magnetito nt practicableno s i e . o examinT e this aspec e formatioth t f magnetito n s beeha e n examined over a range of concentrations in the presence and absence of oxygen all at ambient temperature (20-25°C). -3 A stock solution containing 100 m 0 g iro s ferria n c nitrate with excess nitri cf 1.5-2. o aci H o givp t s used 0a ewa d to prepare by dilution a series of test solutions containing a range of iron concentrations. The solutions were either purged with nitrogen to remove dissolved oxygen followed by the additio f sufficieno n t sodium sulphite solutio o reduct n e on e third of the iron to the ferrous state, or sufficient sodium sulphit s adde wa eo reac t d t wite dissolveth h d oxyge d alsan no reduce irone th thiron ef . o d After stirrin o ensurt g e th e reduction reactio s substantiallwa n s wa y H completp e th e adjusted to 10.0. On standing the properties of the precipitate formed were assessed visually and with a magnet.

Tables 1 and 2 show typical results from these experiments.

Because of the small scale of the experiments, problems were encountered with the exclusion of atmospheric oxygen and some variation in results was obtained. The data presented represents average resuts. From thes e seeeb resultnn ca t i s e absencth tha n f i oxygeto e a magneticn w volume_»roduclo , s i t e th t producebu t m iroa d g n 0 concentration3 s a w lo s a s process is impractical below that. The presence of oxygen inhibits magnetite production gooa ; d quality product only being forme t iroa d nabover o concentration .m g 0 10 f o s

3.2 Effect of Ferrous/Ferric Ratio on Magnetite Formation

e essentiaTh e lreactio th ste n i p.foro t n m magnetite th s i e developmen e F e optimu th o iont n i f so tm e F rati f o o the solution prior to precipitation. From the composition of magnetite FeO.Fe„s cleai t ri 0 that this ratio shoule b d clos o 0.5t e . However precise contro f thio l s ratiplana n o ot scal ee simpl b wil d t thereforan lno e f thii e se b proces o t s i s considere r industriafo d l application t wil i se necessar b l o t y

100 Table 1 Formation of Magnetite from Ferric Nitrate Solution Free from Dissolved Oxygen

Iron Mixing Time (mins) Final Concentration before alkali pH Observations m Fepp 3"1" addition 150 5 10.8 Immediat e- densheav t pp ye black/brown - magnetic

100 7 10.4 Immediate granular, colour - black - magnetic 80 10 10.5 Immediate granular, colour - black - magnetic 70 10 10.6 Immediat es gran a opt t -no , ula s befora r magneti- e c 50 10 10.2 Instant fine - black preci- nitate - magnetic 30 15 10.2 Fine ppt, black/brown - formed afte minut1 r - e magnetic 15 15 10.2 Initial dark green solution forme vera d y fine suspension after 1 minute, which did not start to settle until after minutes3 , black/brow- n less strongly magnetic

Tabl2 e Formation of Magnetite from Ferric Nitrate Solution With the Presence of Dissolved Oxygen

Iron Mixing* Na2So% 2 2 Final Concentration Added Time + Observations ppm Fe3+ cm3 (mins) pH 150 10 5 10.6 Dense brown/blact pp k formed immediately - magnetic 100 8 7 10.5 Black granular precipitate - magnetic

80 6 10 10.7 Dark brown floe quico t k for weakl- m y magnetic

70 5 15 10.8 Brown floe, becoming dense n standino r weak- g - ly magnetic 50 4 11.2 Light brown floe faintly 15 magnetic

30 3 15 11.2 Yellow/brown floe, non- magnetic

»Mixing time considered just sufficient for reaction to be complete. +Before alkali addition

101 demonstrate that some degree of freedom can be allowed in the adjustment of the Fe /Fe ratio and still produce a satisfactory product.

Table 3 shows the results of a series of experiments to determine the effect of varying this ratio on the production of a magnetic produce ove a rangr f iroo e n concentrationsl Al . solutions were purged free from oxygen.

Table 3 Magnetite Formation from Cold .Aoueous Solutions

Ferrous/ Ferric Total Iron Concentration Ratio 50 100 150 300 600 0.2 Brown Brown Brown Brown Brown KM KM KM NM KM 0.3 Dark Brown Dark Brown Dark Brown Dark Brown Dark Brov. n WM WM V;M V;M WH 0.4 Dense Dense Dense Dense Dense Black Black Black Black Black SM SM S M SM SM 0.5 Dense Dense Dense Dense Dense Black Black Black Black Black S M SM S M SM SM

0.6 ff ii n it "

0.7 Brown Dark Brown Dense Dense Dense WM S M Black Black Black SM SM SM 0.8 Green Brown Green/Black Green/Black Dark Brown KM WM WM V;M SM 1.0 Green Green Green Green Green KM KM NM NM N»'

Kon-magneti- Kote M K : c WWeakl- M y magnetic S- StronglM y magnetic

At low ferrous concentrations the formation of a magnetic precipitate shows no iron concentration dependence. However at e /F Feratio e sformatio th beyona magneti6 f 0. o dn c product is dependent on_ iron concentration. At high iron concentration sa ) eve t ratio8 (60 a nm 0. s hig g a 0s a h magnetic product is readily produced.

These results were encouragin e developmenth r a fo g f o t possible industrial scale e scopprocesth f thes o ed an se tests were widene o studt d y possible plant control systems.

In a solution containing both oxidised and reduced forms of an ionic species the potential taken up by an inert electrode has a constant value under standard conditions of species

102 concentration known as the redox potential of the system. This effec n easilca t e use b yo determin t d e whe n oxidisatioa n r o n reduction reactio s beeha nn take o completiot n n becaus a share p e solutioth chang n i e n potentia s i e obtainelaslth t s a d e fractiospecieon f s o i convertee nsother e th Th o t . d s clearlsituatioa t no y s i definen d whee requireon n o takt s a e reduction reactio n intermediata o t n e position.

Experiments were carrien whic i e redo t th h ou dx potential odeoxygenatea f d ferric nitrate solutios s followewa nwa t i s a d titrated with sodium sulphite solution; sample f solutioo s n being withdrawn periodically and the Fe /Fe ratio being determined colorimetrically. Figur 1 showe e changeth s n i s red_o_x potential against sulphite addition, the region where the Fe /Fe ratio was between 0.4 and 0.6 (ie magnetite is . ThimV 0 sproduced 35 grap d e betweean h li s see i 0 )o t 40 nn clearl ye contro th show w f conditiono lsho s neede magnetitr fo d e production is easier at higher iron concentrations. Another featur f intereso e t show n thii n s grap s thai e amounh th t f o t sulphite require o reduct e diro s e th greatethiri n on e f o d r thae theoreticath n l amount.

ooo l l \

I I I . 1 5OO 1 Iron I Concenlrotion (ppm)

4OO Ferrou sFern/ e Ratio between I i O 4 and 0-6

3OO

2°°

ICO

6 8 1C 12 M lu 18 20 22 Sodium Sulphite Fig 1 :-Change of Solution Potential as Ferrous / Ferric Ratio Changes.

It should also be noted that the presence of nitrate ion change e positio e solutioth s th f o n n potential band within which magnetite is formed but not the overall width or general shape of the titration curves.

3.3 Effect of some Interfering Ions on Magnetite Precipitation

The most comprehensive work on the precipitation of magnetite from ambient temperature solution s beeha s n carried e treatmenth r ou fo f tminerao t l waters [2]. More recently reference [8] has also addressed this problem.

103 Table 4

Effec f Magnesiuo t n Magnetito m e Forman o ti

Stirring Fe(III) Mg(II) 2% Sodium Time Cone Cone Mg/Fe Initial Final Ratio Sulphate Observations Mins ppm ppm Added pH pH cm 3

10 15 5 0.33 0.75 2.1 10.7 Very fine black ppt - 5 mins to settle around magnet - magnetic

15 10 0.66 0.75 2.2 10.5 15 y minutesettlinan r fo o st g 10 occur around magnet, most settled within 30 minutes.

10 15 15 1 0.75 2.2 10.7 Fine suspension visible after 15 mins. 1 hr to settle ccmpletelv. Formed overnight - weakly magnetic

10 15 20 .33 0.75 2.2 M Fine suspension visible after 20 mins. Yellow solution afte 0 mins3 r . Light brown floe afte hrs} 1 r . Non-macnetic 10 15 25 1.67 0.75 2. 2 11 Non-magnetic floe afte min5 1 rs Immediate formatiof o n 7 30 5 0.167 1.5 1.9 10.7 black ppt - macnetic 7 30 10 0.33 1.5 1.9 10.7 Fine suspension at first, precipitation on magnet after 4 minutes - magnetic.

7 30 20 0.66 1.5 1.9 11.0 Floe-like suspension, settle s magnetita d 0 2 n i e i.iin - magnetis c 7 30 25 0.833 1 .5 1.9 10.6 Very fine immediate suspension. Fint arounpp e d magnet after mins0 2 . Full settling after ' 0 minute4 weakl- s y magnetic.

7 30 30 1 1 .5 1.8 10.6 Greenish-brow t aftepp n0 2 r min - weaklv magnetic. Dark brown magnetic ppt after 30 min. Floe-like mixture of mapnetic/non-magnetic ppt formed overnight Green floe-like ppt formed 7 1.167 1.5 1.9 10.8 30 35 . Formehr I . i< na floed - like mixture, as above. overnight . 2.0 1 Fine floe suspension - 7 30 40 1.33 1.5 formed overnight, non- magnetic 7 30 50 1.667 1.5 2.2 1 1 Fine green floe, forming brown floe. Non-magnetic

These workers report results from systems where the iron concentrations and those of the interfering ions are much higher e casth tha e s e likeli b nwit o t hy typical nuclear industry effluents. However e maith ,n ions which suppress magnetite formatio e identifiear n s magnesiuma d , aluminiu d solublan m e silicate whereas sulphate enhance productions it s .

104 Table 5

Effec f Aluminiuo t n Magnetito m e Formation

Stirring Fe(III) AlCIt) Al/Fe Initial Final Time Cone Cone Observations Ratio pH pH Min s ppm ppm

7 15 0 0 2.2 10.9 f magnetito Fint pp e e immediately - solution test.

7 15 1 0.066 2.2 10.5 Instanf magnetiteo t pp t .

7 15 2 0.133 2.2 10.8 Fine suspension showing after 3 mins - settled in 10 mins.

7 15 3 0.2 2.2 10.8 Magnetic ppt settled in 30 mins - brown/black

7 15 5 0.33 2.2 10.95 Fine suspensio o formt o ,sl n settled as brown floe - not magnetic

7 15 10 0.66 2.3 10.9 Floe-like brown ppt, not magnetic

5 30 2 0.066 1.9 10.9 Fine black ppt, settled after 3 mins - magnetic

5 30 3 0.1 1.9 10.7 Fine suspension visibln i e 5 mins. Settle s faina d t ppt afte min0 3 r magneti- s c

5 30 5 0.167 1.9 10.4 Tine suspension visible in 5 t?in s. Settle s faina dt pp t in 20 mins - became denser after 2 hr - magnetic

5 30 10 0.33 1.9 10.5 Fine suspension sloo t w form, settled as dark brown floe, non-magnetic

o assesT e effectth s f theso s e eformatio th ion n o s f o n magnetite at relatively low iron concentrations and where the ferrous ione produce ar sn sit i uy b dreductio a serien f testo s s have been carried out in the presence of varying concentrations of magnesium, aluminium and soluble silicate. Again all solutions were deoxygenate y nitrogeb d n purginge th prio o t r test.

6 presen d e resultan Table th t5 , s4 s obtained from which the following conclusion e drawnb n .ca s

(a) Magnesium interferes least strongly, with solutions containing Mg/Fe ratios up to one able to form a magneti w iro lo t even t pp ca nconcentrations e th s A . iron concentration increase e proportioth s f o n magnesium which can be tolerated also increases.

) Aluminiu(b m interferenc mors i e e serious than thaf o t magnesium. At low iron concentrations magnetite formation starts to be suppressed at Al/Fe ratios of 2 s witbuta 0. ,h magnesium e iroth n s ,a concentratio n increase highea s r Al/F etoleratede b rati n ca o .

105 Table 6

Effec f Silicato t n Magnetito e e Formation

Stirring Fe(III) Silicate Si/Fe Initial Final Observations Time Cone Cone Ratio PH PH Mins ppm ppm

5 30 1 0.033 1.9 1 1 Instant black ppt of magnetit - strongle y magnetic

5 30 2 0.066 1.9 10.8 Paramagnetic green floe instantly formed - darkened slowl magneti- y c

5 30 3 0.1 1.9 10.8 Creen flo - darkenee d slowl becam- y e magnetic afte 0 min2 r s

5 30 5 0.167 1.9 1 1 Dark brown floe - not magnetic even on standing

5 30 10 0.33 1-9 10.3 Light brown floe - not magnetic on standing.

7 15 0 0 2.1 10.5 Fine blac t immediatelpp k y - solution blank test

7 15 1 0.66 2.0 10.1 Faint black ppt after 10 min magnetis- c Fine suspension coes not 7 15 2 0.133 2.0 10.2 settle after 30 mins. Non-magnetic . Yellow solution - ferric _- 5 2.0 10.4 15 0.33 floe ppt very slow to settle. Non-macnetic 15 10 1 2. 0.66 10.7 Yellow solution, ferric floe slow to settle as before.

(c) Soluble silicate prevents the formation of magnetite at mole ratios above 0.1. However, unlike aluminium, increasing the iron concentration does not lead to a greater toleratio r silicatefo n .

3.4 Activity Uptake by Magnetite

Following the identification of a wide region of solution conditions where magnetit precipitatee b n ca e d from solutioe th n capability of this material to remove trace .activity from solutio s comparewa n d with tha f ferrio t c floe under similar conditions. -3 A stock solution of 300 g m iron as ferric nitrate containing trace f typicao s l radionuclide contaminant s usedswa . s compositioIt s givei n Tabln i n . 7 e Aliquots of this stock solution were taken and either treated directly with ammonia to precipitate ferric floe (pH 10.5) or dissolved oxygen was removed by purging with nitrogen followe y sulphitb d e additio e /F o product rati ne F oa e

106 Table 7

Analvsis of Active Stock Solution

Solution Component Concentration

Iron mg/litre 300

Magnesium " 10

Plutonium (238, 240, 241) pCi/litre 15

" Americiu 1 24 m 1.4

Cobal0 6 t 2.1

Caesium 137 " 2.9

Cerium 144 1 .4

Europium 154 " 8.0

" Rutheniu 6 10 m 4.2

Strontium 90 " 69

Table 8

Compariso Radif o n o Kuclide Uptak y Ferrib e e Flod Precipitatean e d Magnetite

DF Radio Isotope Ferrie Floc Magnetite

Plutonium 100 150 200 150 800 150 1000 1000 Americium 150 250 100 150 500 800 700 600 Co 60 10 15 10 15 100 80 100 150 7 13 s C NIL NIL NIL NIL NIL NIL NIL SIL 4 14 e C 50 40 12 20 50 90 100 80 Eu 154 20 20 15 10 100 80 100 100 Ru 106 10 6 10 8 50 100 40 60 Sr 90 2 2 1.5 1.8 10 20 15 10

of 0.5 before making alkaline with ammonia to pH 10.5. After allowing both precipitates to settle overnight the supernatent liqui s decantewa d d off, filtered throug a 0.4h 5 micron millipore filte d thian rs clear liquo s analysewa r r fo d radionuclides. The results from A pairs of experiments are give Tabln i n . 8 e

These results confir e expectatioth m n that precipitated magnetite shoula mor e eb d effective gette r tracfo r e active species than ferric r floealphFo .a active species this improvemen s onli t y marginal since ferri cn effectiva flo s i e e getter for the transuratiic elements. The main advantage seen to be obtaine magnetitf o more e dth us e fro s e i econsisten th m F tD for alpha coupled witVu,a signifiant improvement in performance . Ru bete th a r d emitterfo an r S s

107 4. PROPERTIES OF PRECIPITATED MAGNETITE COMPARED WITH FERRIC FLOC

4.1 Settling Rates and Densities

Although magnetite has been shown to give improved DFs for a number of radionuclides it is its high magnetic susceptibility wite attendanth h t possibilit f efficieno y t solids separation using HGMS which originally prompted this work. In much of the scoping tests directed at finding the boundaries within which magnetite coul e formeb d e facth dt thae precipitatth t s wa e black and attracted by a magnet was taken as sufficient evidence that magnetite was being produced.

Having established the operational conditions necessary to produce a "magnetite" precipitate, experimental work was directed towards the characterisation of this material and the comparison of its separation properties with those of ferric floe precipitated from the same solution comp_osition. The composition of the solution used was 300 g m of iron as ferric nitrate with nitric acid to pH 1.5. Treatment used was as before. The settling rates of the two precipitates were measuree plottear e magnetit d n Th Figuri an dd . 2 e e settles significantly faster thae ferrith n c floe despit e facth et that the magnetite particles are a good deal smaller than those of the ferric floe (se e. Scannin4) Figure d an 3 sg electron photo micrographs (SEM o type) tw f o particlese picture th f o ss indicat a possible e explanatio r thifo ns behaviour e ferrith , c floe particles havin a sponge-likg e appearance whereae th s magnetite particles were smooth but "stringy".

T> 1 V I

10 20 30 40 50 60 Time minutes

- SETTLIN: G FIRS E CURVETH TR HOUSFO R

108 18

14 o ^ 12 £ 10

_ 8 o I 6

2 5 10 2O 3O 4O SO Particle Size

Fig 3:-PARTICLE SIZE DISTRIBUTION FOR MAGNETITE.

2 10

I 8

S 6

£ 4 û

2 5 IO 2O O 5 O 4 3O Particle Siz- e Fig 4:- PARTICLE SIZE DISTRIBUTION FOR FERRIC FLOC.

The settled densities of the two precipitates varied with time. At 30 minutes the settled magnetite had five times the density of the ferric floe, at 20 hours it was only twice as dense and by 200 hours it was only 20% greater at 8.5% total solids. This relatively low final settled density is disappointin d clearlan g y more wor s needei k o develot d e th p condition e productioth r a morfo sf eo n crystalline forf o m magnetite.

4.2 Magnetic Susceptibility

As indicated in the scoping test results given in Table 4 the formation of magnetite or at least the development of strong magnetic propertie t instantaneouno s i s n o precipitations .

109 cgs Units r molpe e

0 10 20 30 40

Time a*ler Precipitation ("inutes)

Magnetic Susceptibility of Magnetite Directly after Precipitation Figurl 5 e

Figure 5 shows the change of magnetic susceptibility of two magnetite precipitates with time. Both develop strong magnetic properties of the same value as mineral magnetite at 2.6 and 3.0 CGS units per mole (ground mineral magnetite had a magnetic susceptibilit S unit CG r mole) e pe presencs7 Th 2. . f f o o ye sulphate ion retards the development of magnetic properties initially but results in higher values being developed ultimately. This material was also more crystalline as shown in electron micrograph e particlesth f o s . These e b value n ca s compared with a value of 0.01 CGS units per mole for ferric floe.

5. CONCLUSIONS AND POTENTIAL APPLICATIONS

5.1 Precipitation of Magnetite

s beeIha tn demonstrated tha magnetite-lika t ee b soli n ca d produced from effluents containing iron on suitable treatment of the effluen d thaan tt this solid:

(a) has high collecting power for radionuclides

) settle(b s ultimats rapidlit t bu ye settled density require e increaseb o t s d

110 It is not possible to precipitate magnetite directly from all solution compositions n thai , t a numbether e f commoar eo r n ions which suppress its formation direct from ambient temperature solution. However e additio,th morf o n e iro o that n t naturally present will in the great majority of such cases enable the interferenc e overcomeb o t e .

The use of the cheap reducing agent sulphite to produce the necessary conditions for magnetite precipitation without adding e bul tth f wastoo k e produced when sufficient iro s naturalli n y present in the effluent has distinct advantages for subsequent waste disposal [9],

5.2 Applications

(a) Gravity Settling - the increased density of magnetite gives rise to faster settling of this material than with ferric floe enabling smaller settlers to be used for a given throughput. However e smalleth , r particle size gives rise to some problems with carryover.

(b) HGMS - the very high magnetic susceptibility of this material provide a potentias l applicatio r solidfo n s separation usin a magnetig c filter directl n placa i y f o e settler. Alternatively, a gravity settler to handle the bulk of the solids followed by an HGMS unit to polish the last trace f solido s s appear e attractivb o t s f maximui e m e neededar s DF .

(c) Magnetic Assisted Gravity Settling - again because of the strong magnetic properties of the precipitate a relatively weak magnetic gradient which will both flocculatd an e attract the magnetite particles has been shown to produce high rate f settlingo s d highean r settled solids densities of up to 14%.

REFERENCES

[1] HARDING, K., BAXTER, W., "Application of HGMS to Ferric Hydroxide Filtration" paper 14-4, Intermag, Grenoble (1981).

S Environmenta"U ] [2 l Protection Agency Water Quality Office Studie n Densificatioo s f Coao n l Mine Drainage Sludge", Report 14618 MJT (August 1971).

[3] OKUDA, T., SUGANO, I., SUJI, T. T., "Removal of Heavy Metals from Waste Wate Ferrity b r e Co-precipitation Technique", Filtratio d Separatioan n n Journal (January/ February 1976).

[4] TAKADA, T., "Removal of Heavy Metal Ions from Waste Wate Ferritisation"y b r , Kag Taisako 7 (1977)t o3 , 13 i, RFP translation 281.

] IGUCHI[5 KAMURA, ,T. , INOUE,T. , "Ferrit,M. e Procesr fo s Treatment of Waste Water Containing Hevy Metals", PPM 10, 49 (1979), RFP translation 284.

Ill [6] KOCHEN, BOYD, PRICE, "FerritE. L. Y. . ,T . . , R M e Treatment of Actinide Waste Solutions", RFP 3299 (July 1982).

[7] STREETER, R. C., McLEAN, D. C., LOVELL, H. L., "Reduction of Hydrous Ferric Oxide to a Magnetic Form with Sodium Dithinite; Implication r Coafo sl Mining Drainage Treatment", Am. Chem. Soc. (September 1971).

[8] VAINSHTEIN, V. I., et al., "Influence of Salt Composition e Kineticoth n f Magnetito s e Formation Magnetis It , c Susceptibility and Mechanical Dehydration", Zhumal Prikladnoi Khinii, f>5_ 1 (1982) 133-138. [9] HARDING, K., WILLIAMS, R., Provisional UK Patent 2091135 "A '(1981)

112 THE APPLICATION OF INORGANIC ION EXCHANGERS TREATMENE TH O T ALPHA-BEARINF TO G WASTE STREAMS

E.W. HOOPER Chemical Technology Division, Atomic Energy Research Establishment, UKAEA, Harwell, United Kingdom

Abstract e DepartmenTh e Environmenth f o t B fundinI t a generig c programm f woro e k t AEREa , Harwel examino t l e applicatioth e f Inorganio n n exchangerio c o t s e removath f radionuclideo l s from aqueous waste streams.

A literature survey has identified six exchangers of potential use, they are:- Titanium Phosphate, Zirconium Phosphate, Manganese Dioxide, Polyantimonic Acid, Hydrous Titanium Oxide and Copper Hexacyanoferrate.

A programm f batco e h contact experiments using spiked solutiond an s solutions of Irradiated fuel has examined the extent and rate of absorption f radlonuclideo s from solution Mola1 s d witn nitrati ran hn hydrogeio e n io n concentrations between 1 Molar and pH *• 10; 2hi and 4M nitric acid solutions have also been examined.

This paper present e absorptioe datth th s a n o obtaine r f fa o n o s d plutonium, americium and neptunium from spiked solutions and of total a-activity removal from solution f irradiateo s d fuel.

1. Introduction

n K 198DepartmenI U e 2Environmene th th f o t t initiatet a d AERE Harwell a programme of work with the objective of developing processes which would provide efficien d safan te concentratiof o n the radioactive constituents of waste streams into solid materials which can be consolidated into forms suitable for storage or disposal.

e A literatursearcth f o h e showed tha a considerablt e volume of informatio f inorganio e s availabli us n e cth absorbern o e r fo s the decontamination of waste streams but the reported data are ofte f limiten selectino ni e us dn absorbea g a particula r fo r r

113 waste stream sinc e candidatth e e materia s generallha l y been examine r individuao w fe d a onl r l fo ynuclide s ove a smalr l range of operating conditions. Als e informatioth o n availabln o e a-decontamination is considerably less than that for ß-y removal.

An assessment exercise" e , literaturbaseI th - - n o d e search, identifie n exchangerio x si d s wit a knowh n affinitr fo y actinides and/or fission product n experimentaa d an s l programmf o e work is now in progress to obtain a comprehensive knowledge of the capabilitie f theso s e exchangers absorbere Th . s selecte r thifo d s study are:- Titanium Phosphate, Zirconium Phosphate, Manganese Dioxide, Polyantimonic Acid, Hydrous Titanium Oxid d Sodiuan e m Copper Hexacyanoferrate.

This paper present e e datth th s an o obtaine r fa o s d absorption of plutonium, americiuc and neptunium from spiked solutions and of total a-activity removal from solutions of irradiated fuel.

. 2 Experimental

1 2. Absorbers

Hydrous Titanium Oxide (HTiO) was prepared by precipitation from titanium tetrachloride solutio additioy b n f sodiuo n m hydroxide solution e precipitatTh . s washe wa ed the an dn driea o t d glassy gel.

Manganese Dioxide (MANOX A,Mn02), Titanium Phosphate (PHOTI D.TiP), Zirconium Phosphate (PHOZI , ZrP)RA , Polyantimonic Acid (POLYAN S, SbA) and Copper Ferrocyanide (FeCu) were obtained from Applied Research SPRL, Belgium.

Polyantimonic AciTitaniud an d m Phosphate absorbers have also been prepare t Harwella d .

l absorberAl s were sieve o providt d e material wite th h particle size range 212-425nm.

114 2.2 Waste Simulates

Simulate f salt-freo s e acid wastes were prepare y spikinb d g 1-AM nitric acid solutions with radionuclides. Simulates of salt bearing wastes were prepared fro sodiuM I m m nitrate solution.

Solutions of Am-241 in 3M HN03, Pu-239 in AM HN03, Np-237 in AM

HN0 and irradiated fuel in HN0 ( ~ 3M) were used as sources of 3 3 radionuclides.

The pH of the waste simulate was adjusted to the required

value by addition of HN03 or NaOH.

2. 3 Conditioning of Absorbers

Conditioning is necessary in order to maintain a constant solution pH during the course of a kinetic run. Absorbers were washed in order to remove adherent fines then contacted with water r severafo maintaineH p l a daye require t th a s t a d de valuth y b e

addition of either HN0 or NaOH solution from an autotitrator. The 3 conditioned absorbers were washed with distilled water and, with the exception of HTiO, allowed to air dry for several days. It was observed that HTiO undergoes dissolution at a solution pH < 1.0, and

e conditionethab o t t presence Mn0d th ha 2n i dNaN0f o e 3 sincen ,i solutions of low ionic strength, it tended to break down.

A . Kinetic2 Absorptiof o s n

e rat Th f absorptioo e f radio-iono n s followeswa y batcb d h contact experiments. f conditioneo 0.2g I 5o t d absorbe s addewa r d to 50 mL of waste simulate which had been adjusted to the required e flaskTh s. pH were then shake fixea t a ndhoursA 2 ratr fo e, samples being removed periodically for analysis.

The percentage removal of an isotope as a function of time was calculated from the following equation: A A % 2remova ~—— 0 ( 10 — - l) n.o

115 where,

Ao « initial activity in (unit volume of) solution

At * activity at time t in (unit volume of) solution

5 . Distributio2 n Coefficient Measurements

A 50 mL aliquot of waste simulate solution was added to an accurately weighed amount of ion-exchanger. The solution was then allowed to stand at 25°C, with intermittent shaking, for periods of o weekstw o u.t p

Distribution coefficients (KD) were calculated froe th m following:

l _- _^2-..»- , ,_ V - D A w œ

where,

V = total solution volume (mL) w = weight of exchanger (g)

3= Results

1 . Plutoniu3 m Absorption

The absorption kinetics of Pu(IV) (28 uM) on a variety of exchangers under different solution condition e show ar sn figure i n s 4 inclusive o t 1 . Initial rates were fairly rapi equilibriut bu d m was only achieved after several days contact, and possibly particle diffusio e ratb ey nma controlling n goin2.0o I t . g0 , 1. fro H p m the hydrolysis of Pu(IV) produces species with less positive r l charge1- 2 J , which could account for the observed decrease in uptake kinetics . (figureTher2) d ean shoul1 s d a decreasals e ob e ir. competition for exchange sites by H"1" ions, which, in the case of MnU2, appear moro t s e than compensat y changan e r n chargi eth fo e f o e u specieP s presen n solutioni t .

116 Figurel Figure 2 Kinetics of Pu (iv) removal by inorganic ion Kinetic (ivu P ) f removao s y inorganib l n io c exchange. exchangers.

Sb» 100 TOM No NO 3 pH 7-0 90 25'C

% 70 i removal 60

3 4-5 5 4 3 Time (hrs) Time (tirs)

Figure 3 Figure^ Kinetic f Pu(ivo s ) A uptakSb y eb Kinetics of Pu (iv) removal by TiP

100 solutions »pH 1C contain VOM NaNDj 25'C 90

60 pH l C 70 Ve removaa!l 60

.50

345 24 24 Time (hrs)

117 Figure 5" pH protile for Pu (ivj removal by SbA ) removadv u profilFigurH P y inorganip b r l fo e6 e n io c 2Pu;25°M 5p C exchangers. 25uM Pu;25°C

10 : contaisolution0 1 DnH 1 Msïp NaNO;

tolutuns ï pH 10 contain 1 OH

10'

MnOj 10'

10' -1 54321 PH !HN03]M pH

KD values for Pu(IV) are listed in Table I. Plots of log

KD vs pH are shown in figures 5 and 6. Over the range of nitric acid concentration e grapth _ s expectea i h1M e sb o t froo t dM 4 m linear functio witH p slopa hf o n e e e chargequath th n o o t el

species being absorbed. The limited data in Fig. 6 suggest that

+ + 2

Pu(IVabsorbee b y )ma Pu*«+s a d , [Pu(N0 [Pu(NO)d an )] 3 )froJ m 3 3 2 acid media by different absorbers. With the exception of ZrP, Na~*" ions have little effect on the uptake of Pu. Above pH l colloidal u interfereP s with absorption A producinSb d ,an botmaximua gP hZr n i m the Kn value at pH ~ 1. This inflection point can be calculated r l from the expression for the solubility product"- 2•> :

+ - [PU* p s ]I K

6i++ sp 10-x 7 ID"x " [Pu, 8 M 2. ]-

hence, the pH at which the [OH~] is sufficient to reach Kgp is approximately 1.35.

118 TABLE I values for Pu(IV)

Acidity/ ABSORBERS pH CSbA ZrP Mn02 TIP HTiO

U M A. 660 x 101 5.720 x 10° ° 1.6110 x 0 2.740 x 102 n .m.

2 hl 1.220 x 103 1 3.B310 x 0 1.000 x 10° 7.720 x 102 n .m.

1 M^ 1.340 x 101* 2.160 x 102 3.100 x 10° 3 2.8910 x 0 n .ID,

pH 1 5.16 x 010* * * 1.1510 x 0 * 1.300 1 0x 8.970 x 102 1.020 x 103

pH 2 4.74 x 0lO 11 3 3.4210 x 0 1.400 x 102 5.62 x 0lO 1* 1.160 x 101*

pH 7 5.130 x 102 n. m. n. m. n .m. n. m.

pH 10 3 2.3710 x 0 n. n. n. m. n. m. n. m.

pH 12 7.890 x 103 n. m. n. m. n. m. n. n.

All Solutions with pH > 1.0 contained l.0 M KaNO measuret no • n.md. 25°C

The increase in K^ with increasing pH observed for TiP,

HTiO and Mn02 can be interpreted as being due to the decrease in competitio + whicH y b hn out-weigh e effectth s f colloio s d formation.

2 . Amerlciu3 m Absorption

The exchange kinetics for Am(III) are shown in figures 7 to 9 inclusive. Exchang s fairli e y rapi r solutiofo d d an 1 n wit> H p h shows a dependence on solution pH. These data can be interpreted in terms of the formation of hydrolysis products of Am- •'J. Decrease in competition by IT*" for exchange sites account for rate increases whereas rate decrease e attributear s e formatioth o t d f o n hydrolysis product m witA f ho s reduce r zero d o charge.

KD values for Am(III) are listed in Table II. The

variatio Am(IIIr fo Iogn i n D K 10) with solutio s showi n H i np n e generao increast Th D . K l11 r e figuretren d fo wit an s i d0 h 1 s o reducet solutio e du d, competitiopH n H"*"y b n. Ther s littli e e

119 Figure 7 Figure 8 Kinetics of Am removal by POLYAN S Kinetic [inm A ] f removao s y inorganib l n io c exchangers.

removal 60

Time (hrs) Time (hrs)

Figure 9 Kinetics of Am (in) removal by inorganic ion exchangers

VOM NoHOj pH 30 25« C

345 Time (hrs)

120 TABLI I E valuer fo s

Acidity/ ABSORBERS pH PolyaS n ZrP Mn02 TIP HT10

4 M 1 2.3210 0x ° 8.9310 0x 1.180 x 101 1.010 x 101 n.m.

2 M 1.370 x 102 2.170 x 101 1 1.1510 0x 1.780 x 10"l n.m.

1 M 3 2.8010 0x 2.710 x 101 7.900 x 10° 1 10 1.06 x 0 n.m.

pH l 5.590 x 103 2.360 x 101 4.400 x 101 7.900 x 101 2.150 x 101

pH 3 5 2.7210 0x 1.910 x 103 6.13 0x 10 1* 2.230 x 101* 1.470 x 105

8 p8. H 9.28 x 010" 1 n.m. n.m. n.m. n.m.

pH 10 1.430 x 105 n.m. n.n. n.m. n.m.

pH 12 1.330 x 10 11 n.m. n.m. n.m. n.m.

Ail solutions with pH > 1.0 contained 1.0 M NaNO3 n.m. » not measured 25°C

Kigure 10 pH protiie tor Am im) removal oy inorganic ion (in m profilH )A Figurp removar . fo e1 e 1 POLYAy b l NS 6 exchangers.0.76yM Am;25°C 0.76 y M Am; 25°C and HTiO 10 * pH VO tontan 1 0£ SoNO

iHT.OlH) C»

10' I 10 12 51321 [HN03]M

121 interference from competing Na+ ions for solution pH up to 3-4 but figure 11 shows a decrease in Am absorption above pH 6 which may be due to Am hydrolysis or increased Na+ absorption.

The slope of the graph for SbA over the HN03 concentration range 4M to IM (figure 10) is about 3.2, which corresponds to the uptak a tripositiv f o e m specieseA .

3 . Neptuniu3 m Absorption

mose th t s Np(Vi stabl) e oxidation stat f neptuniuo e t i d an m

s usualli y presen n solutioi t t NpOa n t ?disproportionate"*bu ~ t a s •\ *~ _ _ —.„._.._.._-___„. . J „ [

Np"+ 24-

The exchange kinetics of Np(V) (0.42 mM in IM NaN03) on various exchangers are shown in figures 12 and 13. Rates increase with increase in pH, which is interpreted as being due to decreasing competitio n generalI n fro. Ir m , uptak s slowi e , presumablo t e du y the absorbed species having a relatively large radius and bearing only a single positive charge; particle diffusion is probably rate determining.

Figur2 e1 Figure3 1 Kinetics of Np(v) removal by inorganic ion Kinetics of Np (v) removal by inorganic ion exchangers. exchangers.

, NO o N H O 1 100 100 10M NoNOj pH 1-0 pH 20 90 90

80 80

70 70

60 » 60 % , removarvall removal 50 50

30 30

20 20

10 10

48 Time

122 Measured KD values for Np(V) in acid and 1.0 M solutions are listed in Table III. The variation of Np(V) with solution pH is shown in figure 14. It should be noted

that M0 solutionNaN0n _i 1. e ther0 3. Abov ar - 1 seH p e> wit H p h igeneraa s l increas witp K n hi e increas n solutioi e H p n becaus decreasinf o e g competitio exchangr fo n e . + siteBeloH y b sw + pH = 0, the disproportionation of Np02 to species with higher charge t effec produceminima ne s e i tTh an increasa s. Kp n i e K^-pe ith n H curve.

TABLE III value Np(Vr sfo s )a -.

Acidity/ ABSORBERS pH CSbÂ ZrP Mn02 TiP HTiO 4 £ 4.380 x 101 1.530 x 101 1.110 x 101 7.060 x 10° n. m.

2 M 1.980 x 101 1 1.0310 x 0 6.850 x 10° 4.700 x 10° n. n.

1 hl 2.920 x 101 1.170 x 101 1 " 7.8010 0x ~ 0 n.ni«

pH 1 2 1.5410 x 0 ° 8.4510 x 0 1.130 x 101 5.470 x 101 5.940 x 10°

pH 2 1.800 x 102 1 5.0610 5x 1 1.7410 x 0 5.310 x 102 3 1.9810 x 0

All solution 0 containe1. s M NaN> wit 0 H Op h1. d n. m. « not measured 25'C

) remova(v p profilFigurH N inorganiy -p b lr 4 efo e1 n io c exchangers

tolutioncontai0 NoNO1 H O H n1 »p ij 25'C

123 3 »4 Tota g Absorptiol n

Figure 15 illustrates the extent of a-activity removal by different absorbers from solution f irradiateo s d fuel dissolven i d nitric acid and then adjusted to various pH values. The experimental data on which these graphs are based is listed in Table IV.

Figure 1S Tofaloc removal

Antimomc Acid Hydrous Titanium Oxidi Manganese Dioxide

Zirconium Phosphate Sodium Copptr Huacyono Titanium Phosphate fcrrotclll)

0 1 6 6 4 2 H C 2 4 0 1 8 6 420- 4 Z 1 4 2 01 24 6 6 10 pH (HNOjlM pH [HNOjjM

TABLV I E - stutters Removal Percentage Recovery

PH / Time Manganese Antimonic Zirconum Titanium Copper Hydrous molarity (hrs) Dioxide Acid Phosphate Phosphate Hexacyano Titanium ferrate (II) Oxide

4 M 49.33 0 12.36 0 6.43 / / 2 M 47.50 7.51 54.45 25.95 45.42 / /

1 M 47.75 1.83 70.17 16.09 26.56 / /

0.1 M 91.08 6.23 99.10 35.08 40.85 / /

pH 2 90.00 18.82 99.21 55.90 53.64 3.56 18.91 8 p3. H 48.83 99.00 97.75 69.53 99.61 18.86 /

PH 6 89.50 99.61 99.80 98.18 90.44 98.08 98.84

pH 8 92.42 99.74 98.32 92.54 68.39 77.69 98.06

pH 10 1.50 61.26 98.37 65.80 26.54 / 98.13 -11

124 s apparenIi t t from vera figur s yi 5 thaeffectivA 1 e St t e absorbe a-emitterr fo r s ove wida r e rang f solutioo e n H"*"

concentration. HTiMn0d an O also exhibi strona t g affinitr yfo 2 actinides at solution pH values betweenA-10 and 4-8 respectively.

ZrP, TiP and Copper Hexacyanoferrate show good affinity for a-emitters at solution pH of 5-6 but the absorption of actinides falls off rapidly at higher or lower pH values.

Effece Th f Interferino t 5 3. g Ions

Some waste streams contain inactive salt f elemento s s sucs a h

3 1 + 2 fe ,+ ~ Al"" and Kgwhich arise from magnox or stainless steel cladding treatments or are introduced as process reagents such as ferrous sulphamat u reductioP r fo e Alr o nr complexin3" 1fo " ~ ionsgF . These ions may be present at high concentrations and could seriously interfere wite absorptioth h f radionculideo n n exchangersio y b s .

Exchange kinetic uptake th f Pu(IVr o e fo s ) from l.OM NaN03 solutions containing 0.28 g IT1 A13+, 40 g L"1 Mg2+ or 7 g IT1 Fe3 + are shown in Figures 16, 17 and 18 respectively. With the

Figure 16 Figure 17 Kinetics of Pu dv) removal by inorganic ion Kinetics of Pudv] removal by inorganic ion exchangers exchangers.

» Sb 0 p1 K

P Ti 0 p H1

PH20 HnO, f If 0 p H1 pHVO HTiOINoJ

Time Ihrs)

Time (hrs)

125 Figur6 1 e Kinefic (ivu P ) f removao s inorganiy b l c ion exchangers.

3 4 5 Time (hrs)

exception of Mn02 (pH 2.0), solution pH had been adjusted to 1.0. It should be noted that degradation of HTiO interfered with kinetic measurements and therefore these results are not completely reliable because of change in absorber particle size during the experiment.

The presence of Fe3* markedly reduces the rate of absorption

and the effective capacity of the absorber for Pu. The effect of

3 Al "*muct 3 a " h lower concentratio experimente n th tha n i n s , wit + Fe h o lowe t e absorptio s th ri e effectivnth ratd an e e capacit somo t y e extent. When Mg+ is present at 40 g/L (1.66M) the rate of absorption is reduced but generally there is little effect on absorber capacity for Pu.

Pu K~ measurements in the presence of Fe2"1", Fe3"1", A13+ and Mg2"1" are listed in Table V. Because of the reduction of Pu(IV)

by Fe2"*", Pu Kp measurements are for the Pu (I II) oxidation state. Fe"*2 " has very little effect on Pu uptake, presumably

becauss loweit rf o echarge , which suggests tha u decontaminatioP t n

1 of Fe"3 bearing waste streame improveb y sma y pretreatmenb d t witha reducing agent.

126 TABLE V K values for Pu(IV) - (interfering ions)

ABSORBERS Ion pH

Polyan S ZrP Mn02 TIP HTiO

0 2 8.3310 x 0 3.750 x 101 1.730 x 1011 10 6.41x 0 n.tn. Fe3+ 1 3 10 1.58x 0 1 3.4910 x 0 2.460 x 101 1.765 x 102 * 3.290 1 x 0 2 1.990 x 103 3.100 x 101 3.130 x 1012 10 1.03x 0 3 10 1.44 x 0

Fe2+ 2 k 4.3010 0x 1.953 x 103 ~ 0 2.352 x 103 1.853 x 101*

0 1.465 x 101* 1.950 x 1021 10 1.69x 0 2 5.4310 x 0 n.ra. A13 + 1 3 5.2310 0x 4.970 x 102 ° 8.9010 x 0 2 8.4510 x 5 9.130 x 101 2 9.930 x 101* 4.420 x 102 ' 10 1.07 x 0 4.130 x 1023 0 1 1.50x 0

0 2 6.2410 0x 6.970 x 101 2.760 x 10 l3 10 1.20x 0 n «n« Mg2 + 1 6.280 x 103 5.580 x 102 1 10 1.84 x 0 3 2.0810 x 0 2.910 x 101* 2 1.940 x 10**3 10 1.23x 0 3 10 1.38x 0 2 4.5710 x 0 1.590 x 10 3

gL'7 IFe - ]1 [Al0.2- ] 8 g!'1 [Mg] « 40 gf1

l solutionAl 0 containe1. sNaN M I > wit O0 H p h1. d measuret Ti.rono " . d 25"C * Pu 86 Pu(III)

Am(III) exchange kinetics at pH 3 from solutions containing

either A1+ (0.28 g L"), or Mg" (40 g L"") are shown in figures

3 1 2 1 1

+ 19 and 20 respectively. Experiments with Fewer3 e carried out at

pH orde2.0n i preven,o t r formatioe th t Fe(OH)f o n ;Fe 3 +

completely interfere3 d thereford an wit uptake m A th h. f no eo e results can be shown.

Thera decreas uptake s i eth n ei e kinetic l exchangal n o s e P beinmaterials mose Ti th gd t an affected P ,Zr .

D valueK Am(IIIr presence fo s interferine th th n ) i f o e g ions Fe3+ , Fe2 + , A13+ and Mg2* are listed in Table VI. Whereas Fe3+ is absorbed in preference to Am3"*", Fe2+ is not. It should be

127 Figur9 e1 Figure 20 ) remova(m Kinetic m A inorganiy b f l o s c Kinetic r o Am(nis ) remova y inorganib l c n exchangerio s ion exchanger

1-OJJ NoKO

6 24 46 Tim hrsI e) Time (hrs)

TABLE VI values for Am(III) - (interfering ions)

ABSORBERS PH

Polyan S ZrP Mn02 TiP HTiO

0 1 2.1710 x 0 5.580 x 10° ~0 ~ 0 n«ni. Fe3+ 1 1.460 x 101 ~ 0 ~ 0 ~ 0 ° 1.4010 x 0 2 ~ 0 ~ 0 ~ 0 ~0 ~0

Fe2+ 2 3.426 x 102 1 2,3510 x 0 ~ 0 3.970 x 101 4.810 x 101

0 2.100 x 103 8.450 x 10° 2.000 x 10 * 6.810 x 10° n.m. A13+ 1 1.570 x 103 5.050 x 101 1 5.4810 0x l 6.700 1 x 0 5.470 x 10: 3 6.20 x 010 1* 5.750 x 101 3.100 x 103 3 1.1110 x 0 7.070 x 10 3

0 1 3.7810 x 0 1.960 x 101 1 1.4910 x 0 ~0 n »in« Mg2+ 1 2 9.7210 0x ° 6.1410 x 0 * 7.520 1 0x 3'. 78 01x 0! 3 1.020 1 x 0 3 5.180 x 103 ï 3.380 1 0x 1.380 x 10 5 2 2.4010 x 0 5 2.940 1 x 0

l solutionAl 0 containe1. s > wit H p hd l.OhlNaNO, [Fe] 7 gL-1 measuret n.mno • . d [Al] 0.28 gL'1 25C C [Mg] 40 gL-1

128 noted, however, that the solution pH at which these Fe measurements were carried out, was not optimum for Am3+ uptake. In contrast to Mgd 2an this+ + hav3 ,A1 e very little effecremovam A n o t l ovee th r solution pH range 0-1 but as the pH is increased above 1, the effect

of Al and Mg on KD for Am is different for different absorbers. Mg2* has little effect on the Am Kp values for MnC>2 and HTiO and may enhance them slightly.

KJJ measurement r Np(Vsfo ) uptake from solutions containing Fe34 (7 glT1 ) , A13 + (9.28 g L'1) or Mg2* (40 g L'1) are listed in Table VII. The presence of these ions in the waste simulates reducer Np(Vfo J e valu)KJ th s becaus f eo f eo competition for exchange sites. The affinity of these exchangers for NpC>2+ is greater than that for Na+ since, due to its larger ionic size, its hydration sphere is held less firmly than that of d henc an coordinatios eit + Na 0 moleculen1^ e displacesar d more readily. TABLE VII valueK. Np(Vr sfo NpOs a ) - (interferin g ions)

ABSORBERS Ion pH

CSbA ZrP Mn02 TIP HTiO .

0 ~ 0 ~ 0 ~ 0 ~ 0 rum. Fe3 + 1 1 4.8510 x 0 ° 4.3210 x 0 8.940 x 10° 2.970 x 10° 1.02 x 10° 2 4.070 x 101 ~ 0 1 1.6810 x 0 ~0 l 2.270 1 x 0

0 2.980 x 101 ° 9.6410 x 0 1.990 x 101 7.400 x 10° n «m« A13 + 1 1.120 x 102 ° 8.1310 x 0 4.820 x 10° 6.060 x 10l ° 7.5010 x 0 2 1.730 x 102 1 1.2510 x 0 1.340 x 10° 2 1.1710 x 0 2 6.1810 x 0

0 1 2.2110 x 0 1.900 x 101 ° 7.8210 0x 1 1.4610 x 0 n »nu Mg2* 1 2.350 x 102 1 4.9910 x 0 3.370 x 10 1 5.430 x 102 3.330 x 10 3 2 1.640 x 102 1.160 x 101 2.860 x 10 1 2 1.4010 x 0 8.67 1x 0:

[Fe] - 7 glT1 [Al 0.2]« 8 gL-1 [Mg] - 40

l solutionAl 0 containe1. s > wit H p hd n.m. « not measured 25 °C

129 With exceptio e extenMn0f o nth 2, f competitio o t n l (ani pH d t a n _0 M NaNO1. decreaseg) sequence th n i s e

Fe3+ > A13+ > Mg2*

at the concentrations examined.

Thes e rationalizeb daty ma a ionie n termth i d cf o scharge d an s e competinradith f o i g ionse smalleTh . r charg Mgn o e2"* " compared with that for both Fe3+ and A13+ is reflected in its smallest interfering effect. Fe3* and A13+ possess ionic radii of 0.64 and 0. 51A respectively, and as such, the hydration sphere around A13+ is more firmly bound than that for Fe3*. It is therefore easier for Fe3+ to lose its water molecules and enter the ion exchange structure.

The experimental data collected so far in this study of exchanger s i showins g that ther e severaar e l inorgani exchangern io c s thaprovidn ca t e high distribution coefficient r actinidfo s e ionsn I . general, the value of KJJ is greater for Pu(IV) than for Am(III) and is lowest for Np(V). This behaviour might be expected because of 1 1 3 + + the different charges on the actinide ions i.e. Pu *" ", Am and Np02 . The presence of di- and trivalent metal ions, such as Fe3+, Al3"1" and Kg2"* in the solution, reduces the value of Kp for the actiniae* because of coopetitio r exchangfo n e cites.

The performance of these selected absorbers under flow condition beinw no g s examinei s d together wit a hstud f recovero y y of absorbed radionuclides by elution and absorber recycle.

5. Acknowledgments

The experimental data presented in this paper was obtained by B.A. Phillips, S.P. Dagnal N.Pd .an l Monckto Separatiof o n n Processes Group, Chemical Technology Division, AERE, Harwell. Theie r th hel n i p preparatio f thio n s pape s alsi r o gratefully acknowledged.

130 This work was commissioned by the UK Department of the Environment as part of its radioactive waste management research programme. The results e formulatiowil th e use b ln i d f Governmeno n t t thipolica t s bu y stage they do not necessarily represent Government policy.

References

[l] Hooper, E.W.; Phillips, B.A; Dagnall, S. P; Monckton, N. P. DoE Report No DoE/RW/83. 171, Harwell Report No. AERE-R11088 [2] Coleman, G.H. The Radiochemistry of Plutonium, Atomic Energy Commission. ] [3 Schulz, W.W. Chemistre Th , Americiumf o y , Technical Information Center, ERDA(1976). [A] Burney, G.A; Harbour, R.H., Radiochemistry of Neptunium. NAS- NS-3060, U.S.A. E.G. Washington (1974).

131 SEPARATIO PURIFICATIOD NAN FISSIOF NO N PRODUCTS FROM PROCESS STREAM IRRADIATEF SO D NUCLEAR FUEL

S.A.ALI.H.J.ACHE Institut für Radiochemie, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republi f Germanco y

Abstract New results are presented of studies of the application of inorganic exchangers in the following fields: - Concentratio d removaan n f fissioo l n product molybdenum from solutions of irradiated, short-cooled fuel element target y attachmensb water-containino t g metal oxide d subsequenan s t thermal desorptiof o n MoU3 from the stationary phase. The volatilization yields were clearly improve e additioth y b df o n wate rmobile vapo th phases o ega t r . - Chromatographie remova cesiuf lo m froe aqueoumth s medium-active waste streams of nuclear fuel reprocessing plants on ammonium molybdatophosphate, ammonium hexacyanocobalt ferrat zirconiud ean m phosphate. In this case, stationary phases of ammonium molybdatophosphat mose th t r e fa wer y b e e founb o t d efficient exchangers among those studied.

Inorganic exchangers, becaus f theieo r stability against oxidizing and reducing media and their resistanc o radiationt e , have found extensive application in radiochemical separation techniques and as stationary phasee radionuclidth n i s e generators used world- .wide, especially in nuclear medicine [1]. Another interesting applicatio bees nha n indicate mory b d e recent studie f cesiuo s m removal froconcentratee th m s of medium-level waste and directly from the cesium- bearing partial streams acidified with nitric acid, respectively, in nuclear fuel reprocessing plants. Cesium removal from these solutions would clearly reduce the required additional thicknesses of so- called "lost concrete shields" and thus generate considerable savingcoste th f fina so n si l waste conditioning. Metal oxides, because of their extensive use in generator applications, are among the best proven inorganic fission product adsorbers.

133 e generatoTh r used mose tth frequentls i r fa y b y technetium-99m generator. Technetium, because of its low gamma energy of approx. 140 keV and its halflife of 6 hours, has optimal characteristics for organ functioning tests. The worldwide success of Tc-99m generators began only after suitable exchangers had bee e parenn th foun r t fo dnuclide , molybdenum-99, which is increasingly being isolated from short- cooled fuel element targets irradiated specifically r thifo s purpose, because extremelth f o e y high specific activities produce y thib d s techniquee .Th most successful stationary phase used in technetium generator s beeha s n acid aluminum n oxideca t I . immobilize molybdate anions with high retention coefficients from weakly acidified solutions, while the technetium passes throug columne th h . Several development studies have been conducted in order to utilize the favorable properties of aluminum oxide in fission product molybdenum removal and concentration e result.Th s achieve dato t d o et havd le e inorganic exchanger columns having become established parts in most fission product molybdenum removal processes. Normally, molybdenum decontaminated to a high degree from solution w aciditlo f immobis o i y - lized on Al2C>3 columns and then eluted by ammonia after the completion of the washing processes [2]. In some production plants all organic impurities are removed in a time consuming process in which the molybdenum containing solution is evaporated to dry- ness and molybdenum is sublimed as molybdenum tri- oxid n thest 10OI a e bees . eOha °C nstudiet i ] [3 s attempted for the first time to streamline the proces y combininsb e immobilizatioth g molybdenuf no m to hydrated metal oxides and the subsequent direct sublimatio molybdenue th f no m froe adsorbeth m r matrix. If this modification can be realized, it would clearly reduc e expenditureth e n termi s manpowef o s d timan r e respective oth f e production campaign would - an s de d crease the amount of radioactive waste solutions accumulating without entailin y lossean g qualitf so y finae th lf o product. In order for the desired development to be applicable to full-scale production processes, volatilization should be feasible at the lowest possible temperatures .majoA r reductio sublimatioe th n i n n temperature of molybdenum in volatilization from the exchanger matriachievee b e n th ca xf makiny o b d e us g increased vapor pressure of molybdenum oxohydroxide, Mo02(OH)2; this is a gaseous compound observed in the pure Mo03-water vapor syste temperaturet a m s around 650 °C [4].

In the first few experiments the steady-state exchanger behavio molybdenuf o r s studiemwa medin i d a acidified with nitric acid and containing the respective inorganic exchangers, A12Û3, Sn02/ Mn02d ,an Sb205. Fig. 1 is a plot of the molybdenum retention

134 100 oe l 90 o> °l BÖ 70 aq 60 Mn02 • oq

50 Sn0q a • 2 CO Sb205-oq 30 20 10 0 5 C O C 5 3 0 3 5 2 0 2 5 1 0 1 5 Loading (mg Mo/g Exchanger] e oxideth Figur n s o sv : ] Retentio1 e [% o M f no Mo-loadin 0.0n i gHNO.,M 2 .

of water-bearing metal oxide0.0n a M HNÛ2n i s 3 mediu functioa s a m f feeno d loading e fee.Th d load is the ratio of the amount of molybdenum in the solution to the amount fed into the exchanger. As the diagram shows, aluminue m th oxid s ha thit e a H sp highest retention coefficients with 30 mg of Mo/g of exchanger e comparativel.Th y high retentio A120f no 3 in this pH range can be explained by its high porosity. Exchanger experiment highet a s r concentrations f nitrio c acid produced mainly decreasin o retengM - tion capacitie exchangersl al n i s . In view of the intended use under hot conditions numerous experiments were conducte exchangee th n di r systems mentioned above to define their kinetic and wear behavior. Aluminu mdioxidn oxidti d ean e exhibited the most advantageous preconditions. Antimony pentoxide, becaus s sloit w f eexchango e kinetics, and manganese dioxide, because of its pronounced wear sensitivity, were excluded from further study. To optimiz e retentioeth molybdenumf no - in e ,th fluence of the pH on sorption to A1203 and Sn02 was studied over a wide range at constant molybdenum feed loads (3 mg Mo/g of exchanger) and a molybdenum concentration of 2 x 10"3 mole/1. Fig. 2 is a plot of the amount molybdenuf so m immobilize e Sn0d th 2 an n o d A12O3 exchangers, respectively functioa s H ,a p f no afte hour0 2 r f contacso t time. A12Û3n a act s a s anion exchanger already from a pH of 9 onward, completely immobilizin Mo/f o f exchange go g m g3 r from a pH of 8.5 on. When the pH is reduced, complete retention is maintained up to pH 1, but afterwards

135 100 •: BO g 70 I 60 ~- AliO, j o-o-O n oS

30 20 10

0 1 9 B 7 6 5 , I 3 2 1 0 1 • - valupH e Figure 2: Retention of Mo [%] on the oxides vs. pH

thersteea s i ep declin n capacityi e . SnC>2 showa s similar behavior in principle, but the entire curve is shifte de aci th mord o et region t .i Abov8 H ep n anioa actt achievei s na s5 exchanger H p s t a d ,an complete retentio molybdenumf no , whic maintaines i h d up to a pH of 0. This is followed by a slight decline much less pronounced, however, than in the case of A12Ü3e SnU e shifth e reaso 2th n Th -i tanior - fo n ex n change towards lowe e H lowelevelrp th rs i sbasicit y of the -SnO~-group compared with A^O^. The electron densit oxygef o y deformes i n d strongl posie th -o t y e tivtetravalenth e y metab n io lt oxidation levef lo tin, which binds more firmly the OH~-group to be exchange d allowan d s anion exchang o takt e e place only at higher H+-concentration casse th thaf e o n i n A1203. After optimizatio e exchangth f no e conditiont i s was possible to start the volatilization experiments. s importanItwa o elucidatt dependence eth e th f eo Mo03 desorptio n e loadin e stationyielth th n f o do g - ary phase; for this purpose, A12Û3 with different subsequentld an loadC ° predries s0 wa 10 yt a dsub - jecte 30-minuta o t d e thermal treatmen temperat a t - . tureo 120t °C Op su

sublimee ploth a f s to i Figd3 .Mo0 3 fraction oveexchangee th r r loads t indicateI . s thatr ,fo high volatilization yieldattainee b o st d under these conditions, A1203 mus e loadetb d witmolybdenue th h m carrier almost to the limit of capacity (dashed line) t lowe.A r loads volatilizatioe ,th n yields decline drastically to levels below 10%. This is one major disadvantag A^OSf eo largr ,fo e quantitief o s inactive molybdenum carrier would entail low specific activitie producte th f so . Another reason for avoiding high loads is the marked decline in the exchange rate. Clear advantages

136 2 100 !5 "».. I 9U "o

I «"B

r*» l 7D

0 10 15 20 25 30 -Looding |mg Mo/g AI203]

Figure 3: Yield of the sublimation of Mo03 [%] vs. Mo-loading. Temperature = 12OO °C.

in this respect were foun n SnC>2numbea i d n I .f ro serie f experimentso e influencsth molybdenun eo m sublimation yield volatilizatiof o s n temperatures, the duration of thermal treatment, and of water vapor in the mobile phase passing through the molybdenum- loaded adsorber s studiedswa . Fig show4 . comparable sth e thermal molybdenum desorption yields from systems containing A12Ü3 and e presencth SnC> n i 2watef o e r vapor plotte . subvs d - limation time e characteristi.On developmene th f co t of sublimation in tin oxide is the steep initial rise in 11003 sublimation yields which then gradually declines with the decrease of MoOß on the exchanger surface concentratioe th e droe reaso th f Th .po s i n n gradient, which is the prime mover. The much flatter sublimatioe th ris n i e n yiel A^C^n i d , howevers ,i indicative of volatilization being determined by the slow diffusion step through the pores in the surface. In another program of specific technological orientation conducted withi framewore th ne th f o k Reprocessin Wastd an g e Treatment Project, Chromato- graphie techniques are developed mainly for the specific decontamination of individual medium-active waste streams arising in the reprocessing of nuclear n fuei fuel ld elemenan s t fabrication [53 e underTh . - lying concept is based on the fact that these solutions contain largely similar chemical species of the

137 O B d i 0 2 100 120 KO 160 1BD 200 Time [mm]

Figure 4: Yield of the sublimation of MoCU [%] vs, tim f sublimatioo e loadina t a n f go Mo/g [m g 4 oxide]. Temperatur sublimatiof eo n 105= ni C 0° the presenc vapor0 H_ f .eo

carrier f activityo s . Their remova directla y lb y connected specific decontamination step woul muce db h simpler and more effective than the procedure devised whicn i r hsfa o first nearl aqueoul al y s process wastes consisting mainly of nitric acid and sodium carbonate solutio co-evaporatede nar . Decontaminat- e concentratinth g e carrierth f eo f activitso s i y to be achieved by cumbersome procedures comprising essentially a series of precipitation steps. Institute th t A Radiochemistrf eo y (IRCH) modern, simple and effective decontamination procedures by extraction chromatography have been developed within the Reprocessin Wastd an g e Treatment Projec r sucfo th important partial process streambasie th cs a scar - bonate stream [6, 7], the solvent wash in reprocess- oxalate th d ean ing bearin, g mother liquordu P f so oxalate precipitatio nuclean i n r fuel fabrication. e activitieTh thin i s spresent a fiel e ar dt being concentrate n studieo d f cesiuso m removal froe th m aqueous medium active solution nucleaf so r fue- re l processing plants. If this project were to be imple- mented, it would offer a possibility for removal of the longer-lived Cs-137 source of activity, which would entail clear savings in shielding costs of the waste after final treatment. The process to be developed should permit rather wide variation compositioe th n solutioni e s th f no s from whic he removedb cesiuo t s i m, thus obviating the need for additional expenditures to set optimum process conditions e ,salinit th suc s d a acidith an y y solutione oth f knows i t nI . from experience that the cesium distribution is limited to the process

138 streams acidified with nitric acid enteran de th s evaporator concentrate in this way; a powerful method therefore would have to be able under all conditions to allow high cesium decontamination levels to be achieved in nitrate-bearing solutions strongly acidified with nitric acid. In steady-state and dynamic experiments the distributio retentiod an n n behavior, respectively f cesiu,o m from aqueous solution variouf o s s aciditied an s basicities, respectively varioud ,an s sodium nitrate content e inorganith n o s c exchangers ammonium molybdatophosphate (AMP-1) made by Bio-Rad and ammonium hexacyanocobalt ferrate (NCFC), hydrated antimony pentoxide (HAP) d zirconiu,an m phosphate (ZPH) mady eb the Carlo Erba compan studies ywa d [8] e dat.Th a determined so far underline the clear superiority of AMP-1 under almos experimental tal l conditions.a Figs i 5 . plot of the cesium retention percentage determined dynamicall functioa e s loadina yth f AMP-nf o o g n i 1 systems of different pH levels.

100 -.ftD

HKDH 6 j A

• 1 H NN03 KK0H 1 30. D 7 (HjO H p )• 5 9. H p o BO

0 10 O S O B 0 7 0 6 0 5 O t 10 0 3 0 2 Loading [gCs/kg AMP-1]

Figure 5: Retention of cesium on AMP-1 vs. loading for various H+-concentrations.

e diagraTh m shows that cesiu s retainey i mb % 99 > d AMP-1 in solutions ranging between higher HNC»3-concentration weakld an s y basic solutions wit capacita h y of > 60 g/kg of exchanger. The uncommon versatility of this exchanger is demonstrated by the comparison below of data determined under steady-state conditions in systems whose stationary phases in each case were the NCFC and ZPH exchangers. Fig. 6 is a plot of the cesium retention percentage determined under dynamic conditions from aqueous phases of various acidity and basicity levels, respectively, on the ammonium hexacyanocobalt ferrate (NCFC) inorganic exchange functioa e s loadine solia rth th f dnf o o g phase.

139 pH 12.5

pH 7.0

0.1 M HND3 1M HN0 25 3

conc. HN03

10 20 30 40 50 60 70 Loading [gCs/kgNCFC]

Figur : 6 Retentioe f cesiuo n NCF n mo . loadin vs C g for various H+-concentrations.

The diagram shows the very low retention levels in acid media; even in neutral solutions the use of this exchanger canno recommendede tb . Acceptable valuen ca s e achieveb d . onl9 yf o abovH p a e

Comparable data tak clearlea y different turf i n zirconium phosphat stationare th uses i es a d y phase In neutral systems zirconium phosphate is found to be particularly powerful. Under these conditions, approx 100 g of cesium can be immobilized quantitatively on f exchangero g k 1 . However e hig,th h retentio- nde creases rapidly with increasing shifte s th bot n i h acid and in the basic directions; already at a pH of 1 only some 8 g of cesium/kg of exchanger can be retained basin I . c system lossee th s capacitf so e ar y also intolerably high .comparisoA e retentioth f no n levels attainable in NCFC and ZPH systems with those of AMP-1 indicate e cleath s r discrepanc e rangth n ei y of applicability possibls i t .I e with AMP-1 columns to retain cesium quantitatively from highly acidified, neutral and basic systems without any pH adjustment.

140 1^ 100

pH 1(0.1 MHNDjl

HN03 pH 12.5 25

SMHNOj

50 100 150 Loading IgCs/kg ZPH]

Figure 7: Retention of cesium on ZPH vs. loading for various H+-concentrations.

•1100 f.* 90 fc\1 eo

30 Cs

Cs.3.6MHaN03 1MHN03 20 MAW-simulai«

3 5 10 30 50 100 300 500 1000 Loading [gCs/kg AMP-1]

Figure 8: Retention of cesium on AMP-1 vs. loading in 1 M HN03 for Cs in (a) pure, (b) NaNO-,-contaminated ) simulate(c , d MLW solutions.

141 Another important criterion of the practical applicabilit n exchangea f s behavioo yit e s th i r n i r presenc highef o e r nitrate concentrations, sucs a h those whic W concentratehML n occua n i r . Figshow8 .cesiue th s m retention levels obtain- functioa abl s ea exchangef no r loading from 1 M HN03, 1 M HN03 + 3.6 M NaN03, and 1 M HN03 + MAW simulate. While hardl y negativan y e effect coule db observed in the AMP-1 system, the retention capacities of all the other exchangers broke down even under their most favorable exchange conditions. References [1] "Radioisotope Productio d Qualitan n y Control", IAEA, Vienna (1971), Technical Report Series . 128No p 659-74p , 3 [2] Sameh A. Ali, H.J. Ache, Fourth Intern. Symposiu n Radiopharmaceuticamo l Chemistry (Abstracts), Jülich (1982) 168 [3Biirck. J ] , Same Ali. A h , H.J. Ache, KfK-3754 (1984) [4. Glemser0 ] , R.V. Haeseler, Zeitschrif r anorganischfü t d allgemeinun e e Chemie, Band 316 (1962) 168-181 [5] W. Faubel, P.-M. Menzler, Sameh A. Ali, KfK-3617 (1984) . [6J Haag] , Same . AliA h , KfK-3015 (1981) [7] J. Haag KfK-3460 (1983)

[8] W. Faubel, P.-M. Menzler, Sameh A. Ali, KfK-3742 (1984)

142 EFFICACY OF ARGILLACEOUS MINERALS USED AS A MIGRATION BARRIER IGEOLOGICANA L WASTE REPOSITORY

E.R. MERZ Institute of Chemical Technology, Kernforschungsanlage Jiilich, Jülich, Federal Republi f Germanco y

Abstract The current strateg isolatinr fo y g nuclear wasten i s an underground geological repository reliea n o s serie multiplf o s e barrier o restrict s e releasth t e f radionuclideo e biosphereth o t s . Waste ford an m caniste e backfil th wels a rs a l l material use o filt d l the space between the waste canister and the host rock, provide engineered barriers, whereae th s repository itself and the overlying geologic medium represent natural barriers. A suitable backfill material, preferentially argillaceous minerals, exhibits the potential for providin multifunctionaa g le overal th rol n i e l performance of the waste package, significantly retarding radionuclide migration via solution flow. The backfill barrier will serve physicaa bot s a h l barrier, preventing convective solution flod allowinan w g radionuclide penetration only by diffusion, and as a chemical barrier capabl f chemicallo e y interacting with the radionuclides via sorption, ion exchange or precipitation. The paper presents measurements of the hydraulic propertie e diffusioth d d sorptioan s an n n abilities of various bentonite claysd an s . They includin e vestigations dealing wite sorptioth h d retardatioan n n o mosbehavioutw te importanth f o r t fission product radionuclides strontium and cesium. The Kd-values determine batcy db h distribution measurements shoa w strong dependence froe salth mt concentratioe th n i n solutions. Brine solutions bring about a displacement action upon sorbed ions. In highly compacted bentonites the transport of radionuclides through the backfill may be dominated by diffusion through the inter- stitial water. The rate of diffusion will thus depend on normal diffusion and on the sorption properties of the clay minerals.

Introduction e generaTh l strateg n isolatini y g nuclear o wastt s i e impose a series of barriers between the radioactive materiabiospheree th d an l differene .Th t barriers

143 considered are in their order of proximity to the biosphere: geological repository including overlying strata, backfill and sealing material in the repository, canister including overpack, waste form. s possiblIi t e thae firsth t f teitheo three on re barriers, if appropriately choosen, could by themselves provid e necessarth e y isolatio wastesf o n , irrespectiv e qualitwaste th th f eo f e o y form. Thus, the main function of the waste form would be to contain and safely isolat e radioactivth e e materials during handlin d transportatioan g n procedures until proper depositio e geologicath n i n l repository. Howevert i , is common understanding that the waste form should be made as chemical resistant as reasonable achievable to provide the maximum possible protection within an acceptable cost range. In this contribution the attention is directed towards the usefulness and efficacy of backfill materials on the basis of argillaceous minerals. Emphasis is put only on salt evaporites as the geological host media since salt domes are the favorite choice for a high- level waste repository in the Federal Republic of Germany [1]. Functio backfilf o n l materials Various backfill material e considerear s e useb do t d igeologicaa n l repositor o filt ye spac lth e between the waste canister hose th t d rockan s . Suitable backfill materials may also be used to fill the mine drifts and shafts when the repository is sealed. Proper selection of backfill- and displacement materials, respectively, and methods of emplacement create n additionaa s l barrier betwee e radioactivth n e waste and the biosphere in case an aqueous solution is presen e transporth s a t t mediue repositoryth n i m . A comprehensive review related to backfill performance of various material bees sha n prepare WHEELWRIGHy db T l e[2]a t . Most promising material e argillaceouar s s minerals, preferably expanding clays like sodium or calcium bentonites. The backfill has the potential for providing a multi- functiona e loveral th rol n i e l performance th f o e waste package, significantly retarding or minimizing waste package degradatio d subsequenan n t radionculides release. The primary function is threefold: - to retard or exclude the migration of ground water or salt brine between the host geology and the waste canister;

144 - to buffer the pH value and chemical composition of the aqueous solution and the canister surface in order to suppress corrosive attack; to retard the migration of selected chemical species frowaste th m e matrix into the ground water after breaching of the canister has started. In addition, supportive function e self-sealinth e ar s g of cracks in the backfill or interfacing host geology, the provision of resistance to applied mechanical forcea goo d d an shea t conductibility froe canisteth m r hose th t o t rock system. Experimental Information The objectiv e investigationth f o e s focuseo swa tw n o d items : the hydraulic and e sorptivth - wels a emigratios a l n retardation propertie f selecteo s d backfill materials. Only expanding clays wit higa h h conten bentonitf o t e and montmorillonite, respectively, have been considered as potential candidates, since they combine the two most important required properties, namely, of solution convection suppressing and ion sorption capabilities. They include natura wels a l s activatea l d bentonites and different salt clays with high contents of alcalin r alcalino e e earth bentonites. The ability of these materials to reduce significantly or to stop the flow of brine and ground water between the hos e wasttth rocd e an kcanister primarf o s i s y importance, because it inhibits the single viable mechanis r transportinfo m g radionuclides froe th m repository to the biosphere. Migration tests include studies of the ability of the clay materials to retard the outward migratio f radionuclideo n s leached from the waste form after the structural canister barriers are breached, and secondarly, to slow down the inward migratio f chemicao n l species tha y enhanctma e corrosion of the canister. Experimental Results Materials The materials used are:

1. Natural bentonite (NB), Kärlicher Ton- und Schamottewerke, 2. Activated I (AB)bentonitII , ,II , I e Kärlicher Ton d Schamottewerke-un ,

145 3. Grey salt clay (T3), Werk Niedersachsen- Riedel, m level 0 50 , 4. Red salt clay, brown fraction (T4B), Werk Niedersachsen-Riedel, 13O5 m level, sald Re 5t. clay d fractio,re n (T4R), Werk Niedersachsen-Riedel, 130 m level5 , . 6 Brick clay (ZT), claAltwarmbuchet pi y n near Hannover. e compositioTh varioue th f o ns materials determined by chemical analysie depictear ] n tabl[3 si d . I e The solutions used and their composition are listed in tabl. II e

Table I: Composition of materials used

Component Mass Fraction in %

NB ABI ABU AB III T3 T4B T4R ZT

Si02 60.1 58,6 56.0 52.5 46.1 47.7 45.9 58.1

TiO2 0.88 0.51 0.3 0.7 0.68 0.29 0.43 0.83 Al,0, 17.3 17,6 20.6 16.9 14.4 6.1 12.9 19,4

Fef03 9.5 5.1 4.7 8.8 1.5 1.9 3.0 5.2 MgO 1.8 4.1 3.4 1.9 17.1 9.3 11.9 2.5 CaO 1.4 3.0 2.0 1.3 1.0 19.0 3.1 3.0 SrO < 0.009 < 0.007 - - 0.043 0.11 0.069 < 0.006

Na20 0.21 1.9 3.0 8.4 0.46 0.75 9.9 1.4 K2O 2.1 1.8 1.4 1.5 2.0 1.3 2.2 3.1 CSjO 0.0008 0.0006 - - 0.0005 0.0004 0.0007 0.0006

H2O 7.1 7.8 7.5 7.6 14.6 3.1 3.4 1.8 (total)

Table II: Composition of solutions used

Component Destined Quinere Brine Quarternere Brine Saturated (Quit 1) (dual) NaCI-Brine Ig'U [g/i] tg/i]

Na+ _ 44,3 44,3 14,5 K+ - 31,8 31,8 - Mg?+ - 43,4 42,2 - CI' - 220,4 220,4 21,3 sol- - 4,7 — —

Densityf 1,00 1,233 1,226 1,197 [g/cm3]

146 Hydration Rate The rate of water intake was measured for several bentonite specie t rooa s m temperature _applyinapa g p method originally propose ENSLIy b d N [4J. The samples were compacted to a density of approximately 2,0 g/cm3. The water intake velocity as well as the valu water fo e r saturation sho a dependencw e th n o e sodium contenbentonite th f o t e samples e highe.Th r their share, the greater their water intake capability [5]. The results obtained are illustrated in figure 1.

0 14 0 12 0 10 0 8 0 6 O i 0 2 0 Time [ hours ]

FIG. 1: Temporal progress of water intake of various bentonite specimen

Swelling Pressure Icompactea f d bentonite bods accesha y externao t s l s confinei wate d an ro tha s s dtota it t l volume remains constant ,swellina g pressure Th buils i e. tup mechanis f thio m s phenomen s fairli a y well understood knows i t nI . tha swelline ] th t[6 g pressur smectitf o e e claya functio s i s f theio n r initial compaction density. The swelling pressur f compacteo e d high-density bentonite 0 -2.sample2. 1 = p g/cm3( s s measure)wa d n apparatuia n s show n figuri n . 2 e t allowI e determinatiosth e swellinth f o n g pressure by measuring the force applied on a load cell by a water exposed sample, restraine constana o t d t colume

The results obtained applyin o differentw g t large areas expose wateo t d r acces e illustratesar n figuri d . 3 e In the lower curve by far no saturation has been e facth achieve to t tha e tdu donl y abou e fiftton f o h

147 Screw Rod :-

Bentonite Receiver

Force Lood Cell

FIG. 2: Schematic view of the swelling pressure measuring device

0 10 20 30 40 50 60 Time [days] FIG : Swellin3 . g pressur highlf o e y compacted sodium bentonite Curv : wateA e r exposed surface areO rnm^2O a ; water content after 60 days exposure approx% 4 . 2 Curv : wateB e r exposed surface area 1OO; m Om water content afte dayO 6 r s exposure approx. 16 %

148 the sample surface has been exposed to water access. The upper curve represent mora s e realistic picture. But even in this case no complete saturation of water intak bees ha en reached afte 0 day 6 rexposuref o s .

Permeability Some material transport will take place towards the canister wall and vice versa from the eventually breached canister backwards throug backfile th h l material. The predictive theory which describes the flow of brine or ground water, respectively, in a porous materia DARCs i l laws Ye directl' ;th y derived paramete hydraulie th s i r c conductivity. Q = -K • i • A = discharg Q e rate [cm-^/s] K = hydraulic conductivity [cm/s] i = hydraulic gradient [cm/cm] A = cross sectional area [cm^j e formalisIth n f DARCo m s law e hydrauliY' th , c conductivity, K is a constant relating discharge to the hydraulic gradient. It is a function of both the solid materiae flui th s indicate a d an e l th y db relationship

K =

•y k = permeability (intrinsic) [cm ] p = fluid density [g/cm3] = fluin d viscosity [g/cm-s^j = gravitationag l acceleration [crn/s^] The permeability k is a function of the solid media and depends upon variables such as grain size, grain shape, porosity and path tortuosity. The density and viscosity factors depend entirely upon the fluid phase; and the dependence on viscosity clearly indicates that hydraulic conductivities vary strongly with temperature. After saturatio backfille th f o n , chemical species mae transporteb y d throug backfile th h l eithey rb movement of liquid, or if the backfill' performs as expected y diffusiob , n throug e interstitiath h l water. In either case, the ability of the backfill to retard the migration by sorption, precipitation, or any other mechanism is one of the more important attributes of the backfill. The hydraulic conductivity of brine through compacted bentonite (densit g/cm^1 2. ybees )ha n measuret a d room temperature in a constant-volume flow cell. The backfill materia s confinei l d triaxially between porous stone plates d liquian , d brin s pumpei e d through

149 the material at a constant hydraulic head pressure until saturatio achieveds i n e timTh . e elapsed until saturation yields the materials hydration rate; the flow rate after saturation yield materiale th s s liquid permeabilit n uniti y f [uarcies]o s , ,or equivälently, the hydraulic conductivity in units of [m/s]. The hydraulic conductivity dropped rapidly with time before stabilizin muca t ha g lower value th tha t a n beginnin s originait g f (abouo 4 l 1/ tvalue) . Tablelll summarize measuree th s d data. e higTh h permeability coefficients observed give rise to make the backfill material practically impervious to brine percolation. However, radionuclide transport is still possible - though at a very low rate - by another mechanism, namely diffusion. Sufficient quantitative data of radioisotope diffusivity are still lacking [7,8].

Table III: Permeability measurements of activated Na-bentonite samples. Depicted values after stabilisation (time span elapsed ~ 500 h)

Material Density Hydraulic Head Hydraulic Hydraulic Conductivity Activated Pressure Gradient [cm/s] Na-bentonite [g/cm3] [MPa]

ABI 2.11 15 -103 1. 5 4.8 • 10-13 ABU 2.10 15 1.5 -105 4.0-10-13 AB III 2.09 15 -106 1. 5 5.4 • 10't3

Diffusion Measurements In a series of experiments the apparent diffusitivity of cesium, sodium and chlorine ions have been measured usin staca g highlf o k y compacted bentonite disks forming a diffusion cell and applying radio-tracers of Na-22, Cl-36 and Cs-134 in a saturated NaCl brine. bentonite Th s compacteewa densita g/cm9 o 1. t d 3f .o y After a period of 2 months the cell was dismounted, and the individual disks measured separately with regard to their radioactivity; the y-radioactivity with a Ge(Li)detector and the ß-emitters applying liquid scintillation counting. The diffusitivit r one-dimensionafo ) (D y l diffusios i n given by the equation dc dt d-x dx e solveb o whict y applyindb s hha g appropriate initial and boundary conditions [9].

150 With a plane source containing a limited amount of substance which is diffusing in a cylinder of infinite lengt assumind an h concentratioa g n independen, D t the solution is

c 1 M 1/2 2(7T.Dt) c concentration M total amoun diffusinf to g specieg s X distance from source [cmj| D diffusitivity [cm2/s] t timj [s e The diffusitivity obtained applying this equation to measured concentration profiles yields an apparent diffusitivity (D~), includin effece th g sorptiof to n e solith dn o [10J e relatio.Th n betwee e apparenth n t diffusitivity diffusitivite th d an , a t affecte,D yno d by sorptions i , ,D

D = Da(1 + KDp) = densitsolie p th df o y [g/cm] w datfe aWite availablth h yets a e , show tabln i n, IV e s impossiblii t o drameaningfut y e an w l conclusion. More elaborate measurement needede ar s .

Tabl : MeasureeIV d diffusivities

Element D, (best estimates) [cm2/s]

Cs 2-10-9 Na 8 • 10-7 CI 5- Id'8

Nuclide Migration physico-chemicae Th l mechanismns underlying radionuclide .migration through argillaceous clays are rather complex. In most cases several effects may be super- impose n attainini d g fixatio e releaseth f o n d radionuclides to the solid backfill. Important are adsorption and exchange phenomena, complex formatio hydrolysisd an n , precipitatio surfacd an n e reactions, mixed crystal formation and recrystallisation, filtration effects if suspended particle e presen e liquiar sth n i td media.

151 In order to determine the retardation capabilities of candidate backfill materials, batch distribution measurements of Kd values for strontium and cesium have been performed with 6 different argillaceous clays and bentonites using the various solutions ) g liste0 2 Sample. ] d tabln i d[3 an eacf I 0 so eI (1 h were contacted with 50 ml solution for sufficient time (between 1 and 20 hours). The quinere brine was applied only in the Cs system but not in the Sr system in order to avoid precipitation of Variables have beeequilibriue th n m concentrationf o s Sr- and Cs-ions in the respective solutions, temperature (25° and 100°C) , and pressure (1 and 10O bar).

103

102

ZT.Sr 10° ~H8.Cs

10" 1(T2 10'° 10 1 101 Cs, Sr-Equilibr. -Cone, (g/ll

FIG : 4 Batc. h distribution K

s werThC eed contentdeterminean r S f o s d applying X-ray fluorescence analysis. The results are shown in figures 4 to 8. The results show that strontium and cesium are retained by the bentonites and clays. The substance which diffuses throug e backfilth h l material wil e retardeb l d n transpori t owin o sorptiot g n until equilibrius mi established. Only thee substancth n e will migrate further. Retardation increase e valueth s f a o ss increases.

152 Kd

HR.Sr 102 UB.Sf

AB.Cs

10° T3.Sr

H8.Cs

10,-1

I I I I I 1 I 10-3 10'2 10"1 10° 101 Cs, Sr -Equiliûr. -Cone, lg/1]

FIG. 5: Batch distribution K^-values as a function of equilibriun mi concentrations C d an r S r fo s salt brine, quiner quarternerd an e e brinet sa d 25°r differenan Cfo r ba 1 t argillaceous clays

101

AS.NnCI 10°

- £1 '5

10-

10' i i i i i i I__ l I 10" 10° 101 102 Sr-Primary ] Cone l g/ [ .

FIG. 6: Adsorption isothermes of Sr at 1 bar, 25°C

153 10° AB. Qui 11

c o 10-

10-z -

I I _I I I 1 10 10,-2 ° 10 10'1 102 Cs -Primary Cone, i g/l]

FIG. 7: Adsorption isothermes of Cs at 1 bar, 25°C

103

101

bar1 , (1Sr ,) 25°C bar0 10 ,, 100°(2Sr ) C 10° (3) Cs. 1 bar, 25°C 0 bar10 , , 100°Cs ) C(4

10" J___l I I 10'3 2 10'1 10° 10' Cs. Sr-Equilibr.-Cone lg/1]

FIG : K^-value8 . functioa s sa equilibriuf no m concentrations for Sr and Cs in aqueous solutions at different pressure temperatured an s s

154 However e sorptiv,th e capacity decreases strongly with increasing salinity of the solutions. The cations in the brine give rise to a strong displacement action. An influence of increased temperature and hydrostatic pressur measurabls i e t comparablebu y smal d onlan ly of minor importance interestinn .A g behavious rha been found for the strontium retardation, which is abnormally high under certain circumstances. Obviously, it is due to a precipitation of SrS04, originating from a small e contencla th r iony S n si mineralsf t o . The rat f diffusioo e n throug backfile th h s controlleli d by normal diffusion processes and on the sorption properties of the backfill materials. The migration of various ions exhibit differena s t pattern dependinn o g the sodium content of the bentonite. It seems highly probable that, ovee lonth rg term, transporf to solution through the backfill will be so slow that diffusion will be the main mechanism of radionuclide transport and release. In addition to sorption and ion exchange, precipitation as well as remineralisation may pla n importana y e tretentio th rol n i e f radioo n - nuclides.

References [l] HERRMANN, A.G., Geowissenschaftliche Probleme bei der Endlagerung radioaktiver Substanze n Salzi n - diapiren Norddeutschlands. Z. dt. geol. Ges. 131 (1980) 433-459 ] WHEELWRIGHT[2 , E.J. al.t e ,, Developmen Backfilf to l Material as an Engineered Barrier in the Waste Package System. Report PNL-3873 . (1981p 5 1O ) ] BRODDA[3 , E.G. ,Wirksamkeir Zu MERZ , ,E. Tonn tvo - mineralie s Rückhaltebarrieral n Endlar de i -ebe gerung radioaktiver Abfälle in Salzgesteinen. . geolZdt . . Ges4 (198313 . ) 453-466 ] ENSLINe chemisch[4 Di , ,0. e Fabri (19333 kJ_ ) [5] MIRSCHINKA , Eigenschafte,V. n Behälterwerkvo n - stoffen und Verfüllmaterialien für die Endla- gerung hochradioaktiver Spaltprodukte. Report JÜL-Spez-226 (1984) 1OO S. [6] PUSCH , Highl,R. y Compacted Sodium Bentonitr fo e Isolating Rock-Deposited Radioactive Waste Products. Nuclear Technology <45 (1979) 153-157 ] NOWAK[7 , E.J., Diffusio Cs(If o Sr(IId n an ) n )i Liquid-Saturated Beds of Backfill Materials. Report SAND-82-6750 (1982. p 2 5 )

155 [8] PUSCH, R., ERIKSEN, T., JACOBSSON, A., Ion/Water Migration Phenomen Densn ai e Bentonites. Scientific Basi Radioactivr fo s e Waste Management V, Vol. 11, Elsevier Science Publishing Co,, Amsterdam-New York (1982) 649-658 ] Mathematice CRANK[9 Th , ,J. Diffusionf o s . Oxford university Press, London (1956) [10] TORSTENFELT al.t e , ,TransporH. Actinidef to s through a Bentonite Backfill. Scientific Basis for Radioactive Waste Management V, Vol. 11, Elsevier Science Publishing Co., Amsterdam-New York (1982) 649-658

156 YEARN TE EXPERIENCF SO EXTRACTION EI N CHROMATOGRAPHIC PROCESSES FOR THE RECOVERY, SEPARATION AND PURIFICATION OF ACTINIDE ELEMENTS

C. MADIC, J. BOURGES, G. KOEHLY Sectio transuranienss nde , Institu recherche d t e technologiqu développemene d t ee t industriel, Centre d'études nucléaires, Commissaria l'énergià t e atomique, Fontenay-aux-Roses, France

Abstract An extraction Chromatographie process has been developed for the recovery of actinide element 1 24 2"Îs 2 (Cm87 24 2^ , Am, " Pu, Np, U and others) from irradiated targets, from alpha liquid wastes and from aged actinide stocks n inorganiA . c adsorben stablea usee b s a dn ca t, inert supporinorganin a r fo t c extractant, e.g., TBP- impregnated large th eo silicat variet e Du .f organi o y c extractants available, this technique can be applied successfully to solve difficult actinide separations in quantities ranging from micrograms to kilograms. A comparison of extraction chromatography is also made with resin chromatography and liquid-liquid extraction.

I INTRODUCTION Since about fifteen years the French Atomic Energy Commission has develope da progra e productioth r mfo f isotopeno f actinideo s s elements for frenc d internationaan h l usesovervien A e isotope. th f wo s produced, the accumulated amounts since the beginning of the program and the metho f preparatioo d presentes ni d Tabl . BasicalleI y typeo thertw f so s ei programs. I.I. Major programs (i) Involvin e preparatioth g f speciao n l targets, their irradiatioy b n neutron n nucleai s r reactore chemicath d an s l treatmene th f o t irradiated targets productio f *33no irradiatioy ub treatmend nan f 232ThC>o t 2 targets productio f 238po n irradiatioy b u d treatmenan n f 237(>jp/Ao t r o i 237NpO2/MgO targets productio f 2^2pno u> 243Am, 2^C irradiatiomy b treatmend nan f o t 239pu/Al targets or 2*2pUQ2 targets. (ii) Involvin e milkingth f larggo e amount f pareno s t materia : lproduc - tion of 241 Am by chemical treatment of kilogram amounts of PuO2 metau oP r l stocks ric 2*lpun hi . (iii) Involvin e chemicath g l treatmen f largo t e volume f alpho s a active liquid waste srecover: 24f yo 1 Am. 1.2. Minor programs Which correspon o t productiond f speciao s l isotopes eithey b r milking of stocks of aged parent (228ih, 229th, 23*u, 237y, 239Np, , 2

157 TABLE I ISOTOPES PRODUCE RECOVERER DO D ELEMENT ISOTOPE METHOD Scal productiof eo n OF PRODUCTION since the beginning of the program T 22STh 232uS. 228Th lO^g H milkin f larggo e amounts O of aged 233u rich in 232u R I 229-rh 233u^ 229-Th mg U milking of large amounts M of aged"3u 233u Irradiation of 232yh by neutrons hundreds Chemical treatment of of grams U irradiated 232jhO2 targets R A 23*u 238pu2, 23*u grams N milking of large amounts I î ageo d 2J8puO2 U M 237(j 2Pu 10 of mg O milking of aged 2»

A 2»lAm 2<*IpuC2«lAm kilogram M Milking of aged PuOj or E Pu metal stock s- Chemica l treatment R of alpha active liquid wastes I C 2*3Am Irradiatio f 239pno r 2*2po u u by neutrons 100 g U Chemical treatmen f Pu-Ao t l M r PuC>o 2 irradiated targets

202Cm Irradiation of 2*1 Am by neutrons 10 mg Chemical treatmenf o t C irradiated AmO2 targets U R 2<*«Cm Irradiation of 239pu Or2'»2pu I by neutron sChemica- l treatment too g U of irradiated Pu-Al or PuO2 targets M 2<»8cm 252CfflL248Cm 10 0*Mg Milkin f agego d 232d stock

For the production and purification of all these isotopes chemical separations mus achievee caseb w t fe sn i thes d; e separation e easar so yt be performed due to the large difference in chemical properties of the two element n presencei s e productioth e cas,r th thifo es i f s239fvjo n p (239Np(IV))/(243Am(III)) separatio r 240o n Pu (2Wpu(iv))/2Wcm(iiI)) separation ; for the other isotopes produced difficult separations are needed s . recover tracef yo 233u(VIf so largn i ) e amount 232f so Th(IV) recover f traceso y f 228o , 229rh(IV n i larg) e amount f 233u(VIo s ) (containing 232y) . separation 238pu/237Np . separation 2«Am/2**Cm.

158 At the beginning of the programs these separations were realized using liquid-liquid extraction (L.L.E.) techniques operated in mixer-settlers but ten years ago the extraction Chromatographie technique was developed for preparative purposes and is now applied for all chemicals separations needed for the production of actinides isotopes. That technique appears to e simple productio b th usee kilograo flexiblet d b r dg ean fo n fi ca f mt no I . amount f actinido s e isotopes. This paper focusee experiencth n o s e gained after ten years and describes some peculiar production of actinide isotopes solve usiny db g extraction Chromatographie techniques.

H EXTRACTION CHROMATOGRAPHY TECHNIQUE More than ten years ago extraction chromatography technique was principally use n analyticai d l laboratorie e separatioth r fo s f smalo n l amount f metallio s s fielcIt f application. o dion] 1 s[ extendew no n i so t d process chemistry especially for nuclear applications £21 to [31. The main advantage f extractioso n chromatography technique comparee b n sca d with those of more conventional liquid-liquid extraction or ion exchange chromatography methods. Table II presents such a comparison j in spite of a lower capacity than that possessed by ion exchange chromatography, extraction chromatography seems preferable due to the large versatility of extracting functions availabl simple th d ee an metho d usee whicr b dfo n hca the recovery of actinide trapped in the extracting material. The main advantag f extractioeo n chromatograph s liquid-liquiyv d extraction lien i s simplicite th device th f yo e use perforo dt separationse mth f coursO . e eth small exchange capacitie e columnth f o s s use extraction i d n chromatography make that technique only ideal for the production of figrams to kilograms amount f actinideso s e productioth r Fo . f tono n s amountf o s actinides, extraction chromatography cannot compete with liquid-liquid extraction.

^*\METHOD EXTRACTION ION EXCHANGE LIQUID-LIQUID CHARACTERISTICS^. CHROMATOGRAPHY CHROMATOGRAPHY EXTRACTION Versatility of extracting functions large mean very large

Exchange mean large mean capacity (adjustable) (adjustable)

Extracting simple simple complex device (1 column (! column (mixer-s,ettlers) pump1 + ) * 1 pump) » pumps)

Numbe f plateo r s large large small

Sensitivit f pero y - formanc operao t e - ting conditions small small large

Recovery of the simple very difficult simple actinide trappen i d A diluent washing of A possible method is Technique of e th extracting e columth n allows to calcine the resin solvent washing the recovery of the and to dissolve material ex tractant and of the ashes in acid actinide

Eliminatiof o n easy easy difficult alpha contami- (quasi solid (solid waste) (large volumes of nated material waste) alpha organic liouid wastes)

TABLE n Comparison of extraction chromatography, resin chromatography and liquid-liquid extraction technique actinider sfo s separations

159 n.l. Extracting molecule theid sran uses The choic extractinn a f eo g molecule suitabl r extractioefo n chromatography application depends on several criteria : . affinity for the actinide to be recovered simple condition loadinr sfo elutiod gan needee nar d impregnation conditions of the extracting molecule on the inert support lowest molecular weight possible allowing the highest exchange capacity of the columns low solubilit aqueoun yi s solutions.

MOLE- TYPF EO ° X FORMULA ABBRE- STATE CULAR APPLICATION VIATION (weight)

C^HgO,. /238pu(iv)/237Np(v) (D C^HcO— P— O TBP liquid 266 CfcHgO"' *2*3Am/2*«Cm

Qcl^P-O POX.ll liquid 346 2*lAm(!H)/transition C7H15-0-CH2 elements

C6Hl3 Q. ,C2H5 © P - CH2 - C - N DHDECMP liquid 363 2»lAm(IH)/transition elements Ö "C2HC56 Hli Ö" 3

(o), ,CHbH9 N - C - 2 CH - ,P © o 0 *C<»H9 elements

237 «v)/238 (ni) ©CgHl7-NH*, N03- TOAHNO3 solid «16 Np Pu CgHI7'

I(CH3)2-CH-CH2]2-CH-0> /S © P. HD(DIBM)P liquid 350 2«lAm/238p (iv) D2EHDTP liquid 3i4 u C8H17-0 SH TABL m ExtractinE g molecule theid san r use actinidn si e separations (D tri-n butylphosphate 0 di-n-hexyloctoxymethylohosphine oxide 0 di-n-hexyl diethyl carbamoymethvlene phosphorate © di phenyl di-n-butyl carbamoyl méthylène phosphme oxide © tri-n-octylammomgm nitrate © bis(2,6 dimethyl 4 heptyOphosponc acid di-2-ethy@ l hexyl phosphoro dithtoic acid

Tabl presentI eII s some characteristic f extractino s g molecules usen di our laboratory meetin criterie gth l somal ar eo describe d above. e simplesTh t r whicmolecul fo e behavioP e th h th TB r ee fo ruseth s i d extraction of actinides ions is well known from liquid-liquid extraction (L.LcE.) data. The transposition of the knowledge from L.L.E. to L.LoC«, is simple and easy. The high affinity of TBP for U(VI) and Pu(IV) present in acidic aqueous solutions makes L.L.E. technique using TBP extracting agent ideal for purification of uranium and plutonium. To obtain high distribution coefficient r trivalenfo s t actinides (Am(III Cm(III)d an ) ) with s necessarTBi t Pi e highl us o yyt salted aqueous solutions which induces liquid wastes difficul o managet t a researc y , wh tha f extractins ho i t g molecules havin ghiga h affinit r trivalenfo y t actinide acidin si c medis awa initiated firsa n I t. step POX.l a phosphin( l e oxide) whose propertiee ar s clos thoso et f f POX.lTOPeo o s studiede Or us wa fo le P insteaTh . TB f do the extraction of Am(III) allows a decrease of the concentration of the saltin reagent gou t neede aqueoun di s solutionM s: 3.68 insteao t M . 7 f do LiNO3. The main advantage of POX.ll vs TOPO lies in its liquid nature

160 allowing simple metho r loadinfo d e extractingth g molecul e inerth n to e support. In a second step search for molecules able to extract Am(III) from nitric acid solutions without salting out reagent was pursued. For this application DHDECMP, an extracting molecule developped in the U.S.A. [6] C7] andv^fe new molecule) were selected. At first the use of Win L.L.E. was considered difficult because of the solid nature of this compound involving difficulties for the homogeneous loading on the inert support, but s founwa dt i tha e solvatth t e between nitrid an c aci s liquidi easd dan y to impregnate on the stationary phase. case Ith nf TnOAHNC>3 eo , preparatio loadee th f no d materia s easi l y: first liquid TOA is loaded on the support and when the column is filled with this material, a flow of nitric acid through the column converts the amine s saltie nit hig Th .h affinit f TOAHNOo y r tetravalen3fo t actinides from nitric acid solutions makes this extracting agent suitable purifith r -fo e catio f Pu(IVno r Np(IV)o ) . f HD(DIBM)Po e us e a Th highl, y stericaliy hindered extractans i t restricte e separatioth o t d n between Am(VI d trivalenan ) t actinider o s lanthanides e selectivitTh . e separatioth f o y s veri n y highe speciaTh . l propertie f di-2-ethylhexylphosphorodithioiso c acid abl extraco et t U(VIn )i a synergetic combination froP witmTB h acidic solution o reduct d an es Pu(IV), allow a very special separation U(VI)/Pu(IV) without the need to reduce plutoniu Pu(IIImo feede t th n .i ) Ü.2. Inert support The properties inerneedee th r t suppordfo t use L.L.Cn di : e ar . commercial availability low prices high specific area hydrophobic properties highest specific gravity possible. Three different inert supports were use d: Celit es Chro 5*5Ga ,m Q (Applied Science Labhydrophobid an ) c silica (Merck) s mad. Celitwa e5 e5* hydrophobic by treatment with dimethyl dichlorosiiane. The particule size d theian r n 0 1* o t distributio0 s 11 Chro s Ga i md Q f Celito an n 5 e5* specific arem2.g-l1 . ass e highes Th . t specific gravite hydrophobith f yo c silica makes this product more attractive than the two others, this permit highest extracting capacities for the column per unit of volume. Ü.3. Preparatio characteristicd nan stationare th f so y phases Impregnations of the stationary phase with the extracting molecule are carried out as follows t a certain mass of material is placed in contact with a solution of extractant in hexane or acetone, the solvent is then evaporated under reduced pressure by means of a Buchi Rotavapor rotary evaporator. Impregnation levels in mass in the final mixtures are respectively : TBP (27 %), POX 11 (30 %), DHDECMP (30 %), *V (30 %), TOA , (2HD(DIBM) . 5Batche%) %) , HD 0 %) f (3 2mixtur0 EHDTo s P P(3 TB P+ e can be prepared at the 3 kg scale. Table IV presents the exchange capacities of the different stationary phases used, for the extraction of actinide t differena t oxidation states from nitrate solutions. Ü.4. Chromatographie equipment The chromatograhic equipment includes : plexiglass columns packed with the stationary phase impregnated with the extractant Prominent proportioning pumps storage tank r variousfo s solutions gauge for in-line detection at the outlet of the column.

161 STATIONARY EXCHANGE PHASE EXTRACTED COMPOUND CAPACITY rn.moJe.g~l

.U02(N03)2. (TBP)2 0.507 TBP (27 %) 4Pu2 0.507 Um(N03)3. (TBP)3 0.338

,U02(N03)2 (POX.11)2 0.133 POX.ll (30%•|piXNO3) ^ (POX.U)2 0.433 l Am(N03)3 (POX.U)i, 0.216

DHDECM) % 0 (3 P Am(N03)j (DHDECMP)3 0.275

«V (30%Am(NO3)3 fcM3 0.269

TOA (25 %) Np(N03)6(TOAH)2(TOAHN03)2 0.177

HCKDIBM)) % 0 (3 P AmO2(D(DiBM)P))2 (HD(DiBMP))2 0.21*

' D) 2% EHDT 0 (2 P (HA) 1 UO2A2-TBP 0.282 TBP (10%)

TABLE rv Exchange capacities of the different stationary phases used

The column is filled by successive additions and repeated packings of the dry stationary phase. Once the packing is completed and the top cover cemented colume th , t s placecelle nglove-boi ho th ,e n di wherth r xo t ei undergoes pre-equilibrium designed to produce the chemical conditions require r fixatio expeo dfo t aire d th ln an . Two kinds of gauges are used for in-line selection at the outlet of the columns . small Geiger-Muller tube for detection of 241 Am (60 keV) cerium doped glass scintillator for alpha detection. Dependin amoune th n gf o actinid o t extractee b o et e siz dth f colume o n f stationaro g k varie9 o t y s g phasfro0 10 me contained e mosTh . t used columns contain 2.8 kg of stationary phase and possess a 0 ; 82 mm and a thei, mm r voi0 heigh70 d ; volumh t abous ei liters5 1. t . Degradatioe th f no colum n occu nca washiny b r r radiolytigo c damag e extractantth f eo e Th . column e mosth s t sensitiv o thest e e phenomen e thosar a e loaded with DHDECM e extractioP th use r dfo f americiuno m 2*1. Columns loaded with D2EHDT e sensitivar P o oxydatiot e y nitrib n c acid media, s i thu t i s n anti-nitrita necessar d ad o t ye agent (NH2SO3H e feedth n .i ) Used columns can be eliminated as alpha solid waste after removal of the solution contained.

m SEPARATION AND PURIFICATION OF ACTINIDES ISOTOPES BY EXTRACTION CHROMATOGRAPHY Instea overaln a f do l description speciaa f o l proces productioe th r sfo n actinidn oa f e isotop wile ew l describ presene th n ei t chapter some peculiar points correspondin differeno gt t processes: treatment of irradiated targets recovery of actinides from alpha active wastes recover f decayo y products from aged actinide stocks. OLL Treatment of irradiated targets NeptuniuIIL LL irradiate7 m23 d targets This program was initiated in France in 1969 in order to prepare plutonium 238 batteries suitable for nuclear pacemakers. That production of plutonium 238 was based on the irradiation of Np/Al or NpU2/MgO target n nucleai s r reactors e chemicath , l treatmen e irradiateth f o t d targets aftex monthsi r f coolino s g wit a doublh: recoverm ai e f o y plutonium 238 and of neptunium 237 for recycling. The overall process was based on liquid-liquid extraction technique operated in mixer-settlers after dissolutioe th targete th f no nitrin si c acid solutions [83-tlOl.

162 In 1975 we started to test extraction chromatography as an alternative for the L.L.E. process. Here we will focus on two experiments concerning e difficulth t separation between plutoniu d neptuniumman . Suc a hsepa - achievee ratiosystemb o n tw nca y db s base: n do . Valence adjustment, Pu(IV)/Np(V) then selective extractio Pu(IVf no ) by TOAHNO3 columns . Valence adjustment, Np(IV)/Pu(III) then selective extractiof o n Np(IV TOAHNOy b ) 3 columns.

(i) Pu(IV)/Np(V) separation on TOAHNO3 column In nitric acid media the most stable oxidation states for plutonium and neptunium are Pu(IV) and Np(V), it is thus possible to perform the separation by selective extraction of Pu(IV). Such a separation faces with a difficult Np(Ve th : )y tren disproportionato dt e accordino gt + 2+ ) (1 ^ Np^ + 2 O H + NpO22 NpO4 H 2 + 2 + + If in nitric acid of moderate concentration reaction (1) is shifted to the left, in the presence of an extracting agent having affinities for Np^+, NpO2 bothr fo r 2,*o reactio shiftes e righti ) th n(1 o t d, thu neptuniue th s m is coextracted with the plutonium(IV). To minimize this phenomeno d thue an efficiencth e ns th f o y Pu(IV)/Np(V) separatio e feenth d mus nitri w e adjusteb lo t ca acidito t d y (1 M) compatible with hydrolytic properties of Pu(IV). A dissolution liquor of an NpO2/MgO irradiated target of total volume = 23 1 : Np(V) = 11 g.l'1, Pu(IV) = 0.6 g.l'1 was adjusted to HNC>3 = l M and then injected on a. TOAHNO3/Gas chrom Q column containing 1015 g of stationary phase (0 = wit) hfloa mm w0 rat70 = e l.rr H 5 equa, 2. 61o 0.mm t l After washine gth column l.h" 5 wit2. t h,a plutoniuHNC> s N elute 2 wa 3= 8 d m23 wit a h solution of 1following composition : r^SO^ = l M, HNO3 = 0.2 M.

Forteen gram f 238pso u were recovere cyclee on n di , 237(sj s frea f pW eo 238pUj decontamination factor F.D. Pu(Np) was only moderate =20 but acceptable for a first cycle of Pu/Np separation, decontamination of 238pu from fission products was high.

(u) NpOV)/Pu(m) separation on TOAHNO3 column Separation between neptunium 237 and plutonium 238 can be achieved after reduction to Np(IV) and Pu(III) by selective extraction of Np(IV). Such a separation is difficult to perform with high concentration of plutonium 238 in solution because of the intense radiolysis leading to the destruction of the excess of reductor used and thus to the oxidation of plutonium to Pu(IV). Such a process can be used for final purification of neptuniu contaminate7 m23 d with trace f plutoniuo s m 238. Recentle th y purification of 300 grams of neptunium 237 contaminated with 17.5 ppm of plutoniu e larg8 (duth m 23 eo et differenc e halth f n i elive f theso s e isotope e alphth s a contributios realizewa ) % d0 3 f plutoniu o ns wa 8 m23 using the following procedure.

Feee concentratioTh d M f g.l~neptuniu0 o n6 10 n i * e s fee th wa dn m i HNC>3, the reduction of neptunium to Np(IV) and plutonium to Pu(III) s don a y wa additiowarminb ee d th an f f N2H5NOo no gM 5 1. o t 3 solution to 80°C during two hours. Thus after cooling the solution was ferroun i M adjuste 1 s 0. sulfamat o t d e (Fe(NH2SO3)2) freshly prepared under nitroge dissolutioy nb f irono sulfamin ni c acid e presencTh . f eo ferrous sulfamat e feeth dn i eallow e maintenancth s f plutoniueo s ma Pu(III) presence Th . f smaleo l amoun f iron(III o te feedth n ,i ) inducea s small oxidatio f plutoniuo n o Pu(IVmt d thu an a )decreass e th n i e performanc separatione th f eo .

163 The feed is then injected on a TOAHNÛ3 column containing few percent of octanol 2 in order to decrease a little the distribution coefficien f tetravalento t actinid thud ean s improv e qualite eth th f yo separation. Washin e columTh g n loaded with Np(IV s washewa o )t HNÜy b dM 3 3= remove plutonium and iron ions present in the void volume. Elutioe elutioTh f Np(IVno n s performewa ) d wit a solutioh n: HNO 3= e complexinth ; H2$O o t M e 1 0.74= du 0. g5M propertie f SO^so ~ ions. In the purified neptunium the concentration of plutonium was 0.24 ppm. The decontamination factor for the entire process was F.D.fsjp(pu) = 73 which is a good performance for this low level of 238pu contamination. JH. 1.2. Plutonium 239 irradiated targets £ 11 j, [ 121 This progra s initiate e productiomwa th r fo d f americiuo n d an 3 m24 curiu isotopes4 m24 Pu/Ae Th . l targets containing f fissilinitiallo g e0 y40 material were irradiated with an integrated flux of 1L28 n.kb°l. Chemical treatment started after three year coolinf so g time. The overall proces s operatei s y L.L.Cb d . after dissolutioe th f o n target n nitrii s c acid whic s i realizeh d usin 8 liters8 ge highl Th . y radioactive solutions contai Ci.1-0 n 35 /3,f 1f 242po o 7 emitterg 4 . u 4 d san 8.35 g of 243Am and 7.44 g of 244cm. The firs tprocese steth f po recovere s th lie n si purificatiod yan e th f no plutonium 242, which is realized by extraction chromatography using a TOAHNO column3P columTB prevena o T .r no t cloggin f thesgo e columns by fission products precipitates, the feed is filtrated through a small column containin f silicao d .be Aftega r washin e columth f go n with nitric acid solutio e elutioe plutoniunth th f realizes i no 2 m24 d wit hsulfonitria c solution (TOAHNO3 column a hydroxylammoniu r o ) m nitrate reducing solution (TBP column). The largest difficulty associated with that treatment liess e separatioth In f trivaleno n t actinides (243Am, 244cm) from trivalent lanthanides present in the solution in large quantities (=140 g). e separatioth In n Am/Cm e difficultieTh . s have been solvey b d L.L.C. using a TALSPEAK like system, where TBP extractant was used instead of HDEHP. Thus in this system the selectivity of the separation is only due to the DTP A present in aqueous solution. All the lanthanides possess a higher affinity than Am(III) and Cm(III) for TBP column from LiNOs = 8 M, HNC>3 = 0 + 0.05 M. DTPA = 0.1 M, pH = 3 1.2 solution, the separation factor a= D£^ *.D^m 3+ is large for the heavied lighan b t T rlanthanide r Ln(IH)fo 4 closachievs o i L T . d o et s an ea good separation (Am, Cm) from Ln, it is necessary to perform several cycles. Separation between americium 243^and curium 244 is based for the first e samth ste en po procedure . Figur present1 e e separatioth s n between 1.51 g of curium 244 and 1.75 g of americum 243 using a TBP column and a Talspeak like system. Final purificatio f americiuo n s performei 3 m24 selectivy b d e extractio f Am(VIno HD(DiBM)a n o ) P column usin ga proces s identical with that described below for the purification of americium 241. That process appears very simple and flexible, easy to operate in the difficult conditions found in hot cells. Until now about 100 g of americium 243 and =90 g of curium 244 of high purities have been successfully produced.

164 Concentrotion î"Cm or »«Am ... 2.0- (B.I-M

M. r™" — ,

1.2- 11 — •

• !i~ 0.1- - - f i Tl 1.7' •«) »*Cm 1.5 Si i i r "1i (U- •j.' i i - TM .!~ !__ , 0 j — _ llr u1 i ! i " ~ ^ ^ ' s — ' o^ "* " ' *

-1.57vf 3i ¥,^""»3.251 Eluote volumt

m 1 vi

Figur 1 e Separation 2*3Am/2**C extractioy mb n chromatography on a TBP/Cas Chrom Q column

, LiNOM 5 M 0. 7 3 = * Ap , M = DTPFEE 1 0. DA= SCRUB = LiNO3 g M ELUTION = LiNO3 g M DTPA 0.1 M, pH 1.20 Column TBP (25 %) Gas Chrom Q, m = 500 g, L = 64 cm, 4,2 cm Flow rate= 1.2 l.rrl

m.2. Recover actinidef yo s from alpha aqueous wastes In 197 8a progra s initiatemwa d wite goa th f hproducino l g large amount f o americius 1 use2* o t mprepard e neutron sources (americium 2*l-berylliu r mamericium-lithiumo l industryoi r fo e firs) Th .t sourc f americiuo e e C.E.A1 founth m 2* n i d . consiste alphn di a aqueous liquid wastes: "Masurca" waste resulting from the reprocessing of certain irradiated fuels and from criticaility analyses. Cadmium was added to e solutioth r safetfo n y reason thed s initiaan snit l volum f aboueo t distillatioy b reduces 3 m wa * 403 no 0dm t . Waste coming froe reprocessinth m f fabricatioo g n scrap from fabricatin , PuK>g(U 2 fuels intende r fasdfo t breeder reactorse Th . annual volume of this waste solution amounts to some tens of m^. The americium 2*1 contents of these wastes were respectively 108 mg.1-1mg.1- 8 e uraniu2 th d s 1an A . m conten e "Masurcath f o t " waste was high (12g.l~i t appeari ) s that L.L.C. wasn't suitabl s recoverit r efo y prior to recovery of higher actinides. Thus conventional L.L.E. using TBP organic solution in mixer-settlers was used. The neptunium 237 and plutoniu wer9 m23 e recovere n exchangio y db e chromatography usinn a g anionic resin (IRA *00) prior to the recovery of americium 2*1 by extraction chromatography orden I . decreaso t r concentratioe eth e saltinth f no g out reagent necessar obtaio yt n high distribution coefficien f Am(III)o t - di , nhexyloctoxymethylphosphine oxide (POX.11) Ü31 was selected as extracting agent. A moderate concentration of LiNC>3 = 3.6 M was found sufficient to obtain high KrjAnidlI) (—J.lO^ml.g-*) for a nitric acid concen- nevertheles1 POXe 1 .Th e . disadvantagth M tratios 1 0. ha s f no f extraceo - ting Fe(III) with rather slow kinetics e "Masurca, e th caswhic th f eo n i h " waste gives a sharris o t ep dro n Kr}Am(III)pi e additioTh . f EDTo n o At aqueous solutio equan ni l concentratio Fe(IIIo nt ) (0. help) 2 M o avoit s s dit extraction and to restore all the extractive properties of the POX. 11 for e selectivAm(III)th o t e . eThi du formatio s i s e completh f o n x FeY~ (with Yrf = EDTAy ) whose formation constant log(FeY- )= 25. 1 tl* s i muc3 h higher than that of the americium(III) complex = log (AmY~) = 18.0 115]).

165 Experiments were realized to determine the exchange capacity of POX.l s Am3+v l , simulation f Am3o s Nd^y s +b als *wa o use orden di o t r decreas e irradiatioth e e chemisth f o n t workin a glove-box n i g . Figur, e2 presents studies of saturation of a POX.ll column by Am(III) or by Nd(III) nitrates. In each case the breakthrough is obtained for a stoichiometric ratio POX.11/M(III) close to * indicating that the extracted complex possesses the following formula: M(NC>3)3 (POX.ll)^. That result was confirmed either by the study of POX.ll (30 %)/SiO2 mixture saturation in tese saturatioth t y tubeb a POX.l r f o o ns l dodecane solution. Suca h stoichiometry is very unusual because for trioctylphosphine oxide (TOPO) trivalent actinide r lanthanideo s e extractear s s M(NO3)3.(TOPO>3da e Th . POX.ll possesses a high affinity for U(VI) and Pu(IV), for these ions stoichiometrie e similaar s o thost r e obtained with TOPO* (POX.l 1)2 and Pu(NO3)^ (POX.l 1)2.

v o ; IS 20 2S Volum» (ml)

_u )f-CH2-O.C7H1s C&H^ 1 3/ 0 c = o >,

g- u

Ov, 0 1 8 6 Velum» (ml)

Figxir 2 e Saturatio f POXono l l/SiC>2 column with d 241N Ad man

Column, POX.lg l mm = s0 8 l= (29,H , 7 %)/SiOmm 5 = 0 7 Void volum 1.3e= l (Am**)7 l m (Nd3» m 1 1, ,) Aqueous phase s

1 , [Am'» IHNO, CLiN03M M 1 6 0. 3.8!= 3. 3 ]i = )1 g.l" [Nd^+r o ! g.l = l - Flow rates = 15.1 ml.h-1 (Am^*), 1.25 ml.h-l (Nd^*)

166 Durin e chemicath g l treatmen e "Masurcath f o t " waste rather large POX.ll colum s usewa nd containin kilogram9 g f POX.lo s %)/SiO0 (3 l 2 stationary phase. A feed of 685 liters of total volume was treated, the decontamination factor of the effluent in americium 2*1 was 250 ; s recoverewa ) americiug 7 elutioy db (1 1 nm2* wit liter0 h2 f nitriso c acid (concentratio s ndecontaminatioit yielfacto) % witd 3* 9 an d 8 h = r s v n Cd2+ and Fe3+ was close to 130 1161. The major drawbacks f POX.lencountereo e us le procesth e th n di r sfo recover f americiuo y 1 from2* m alpha liquid waste e rathes th lie n i sr complicated feed adjustment :presenc s salting-oua f M LiNC>o e 6 3. 3t reagenf EDTo d compleo At an t x Fe(III) presen e feedth n y .i t Thawh s i t new molecules abl extraco et t trivaient americium from nitric acid media without nee f d salo possessind an t g high affinit d selectivitan y r fo y trivaient actinides were considered. The first molecule investigated is di-n- hexyldiethylcarbamoyl-methylenephosphonate (DHDECMP) well studied for L.L.E. in the U.S.A. [5]. Distribution coefficient of Am(III) between nitric acid solutions and DHDECMP (30 %)/SiO? stationary phase were measured at low nitric acid concentration KoAm^* = 0.9ml.g~l (HNC>3 = 0.1 M) and KpAm3+ maximum = 26 ml.g-1 is obtained for HNO3 = 6 M. It is thus possible to extract trivaient americium from moderate volumes of 5 to 6 M HNÛ3 (due e relativelth o t valuw lo yf K[)Am3-» o e - maximum o perfort d e an )th m elution with water. Figure 3 presents an experiment related to the separation of americiu 1 from2* m fission products 13*, 137c d lO^Ran s u ancj from inactive contaminants contained in a real waste. During the loading step

»"Arn

0 0 _ (DHDECMP) t 01

o Ê E Figure 3 Purification at americium 241 from an industrial waste o O by extraction chromatography using a DHDECMP/SiOj column

Colum g DHDECM l n= %)/SiO0 m Pm (3 O S = 2 , H «5= , 5mm ) alph'Ç a active waste (HNÛfre) u P N e d î fro3- an mU Aqueous <2> HNOM 5 = 3 5 V(ml) Phases J3> H2O flow rate 2 ml.h" '

167 fission products and inactive contaminants escaped from the column. After e washine th colum th a smal f y o gb n l volume th f , HNC>o eM 5 3= americium 241 is eluted with three void volumes of the column. Americium 241 was recovered with a yield of 98 % and after conversion to AmO2 s puritit » whics foun% verya wa 5 s d8 hi y good performanc r onlfo e y one cycle. Adaptatio f DHDECMo n P proces e purificatioth o t s f largo n e amount f americiuo s grams0 e (5 s realizee th live1 th wa f )m o t 2* s bu d columns were e limitefounb o t d d becaus e washinth f radiolytid eo an g c damag DHDECMPe th f eo . Another molecule able to extract trivalent americium from nitric acid solution was studied °* the diphenyl di-n-butyl carbamoyl méthylène phosphine oxide (#9. High distribution coefficients of Am3+ were obtained from nitric acid solutions with a maximum for HNO3 = 2.5 M. The extraction reaction is :

, 3NO3- + 3^^Am(NO3)3$«03 (2) The usefulnes f thao s t molecul e recoverth r fo ef americiuo y m from nitric acid solution L.L.Cy b s s demonstratewa . experimenn a n i d t where a nitric acid solution (HNO3 = 2 M) containing 13,4 mg.l-i 241^m s loade a smalwa n o d l column containin *Pf %)/SiC>0 o P (3 g l g 2 stationary phase. Four hundreds voids volume of the column were decontaminated with F.D. (Am3+) = 388. Americium 241 can be eluted from the 2 column, after washing by water, with a K2CO3 solution. ÏÏL3» Recover decaf yo y products from aged actiniae stocks m.3.1. Recovery of americium 241 from aged plutonium metal or PuU2 stocks The recover f americiuo y fro1 m24 m alpha active waste a rathe s i s r complicated method due to the large volumes involved and their low concentration in americium,, A more attractive way to produce hundreds of grams amounts of americium 241 is to treat aged stocks of plutonium metal or plutonium dioxide initially rich in 241pu> -rne first step of the process consist dissolutioe th n si e startin th f no g scalg materia 0 n ei 30 n o l nitric medium (containin ion~ F g ) under reflux condition a tantalu n i s m dissolver« After filtration the solution is injected on a TBP column for Pu(IV) extraction, americiu escape1 m24 s fro columne mth . After washin e columth f o g n nitriwitM 5 h c acid e plutoniuth , s mi elute s Pu(IIIda ) wit hhydroxylamina e nitrate solutio thed nan n converted in PuO2 e raffinatTh « e containin e 241^th g s j tnem n adjustee th o t d following conditions: H+ = 0 + 0.05, Al(HI) = 1.8 M, DTPA = 0.1 M and columP TB loade orden a i n n o decontaminatt o dr e americiu fro1 m24 m impurities (Pu, inactive metalli co reduc t ions d e volume an th e)th f o e americiated solution. AftecolumP TB washina r e n th witL1NO f M go h8 3 solution, americium is eluted with 2 M nitric acid. Final purification of americiu s performei 1 m24 d wit hmethoa d describe Figurfirsn e di Th t . e4 step consist e precipitatioth n i s f Am(OH)3o n , whic s redissolvei h d after filtration, in concentrated K,2CC>3 solution. The addition of K2S2Og allows the precipitation of americium(V), potassium double carbonate. After precipitation, the precipitate is taken up with an oxidizing acidic solution containing K2§2Og and Ag ions acting as catalyst in the oxidation of Am(V) to Am(VI) by S2Og—ions. After warmin o 50"t g C americiu s mtotalli y converte o Am(VIt d a spectro( ) - metric measuremen performes i t d prior loadin solutioe gth n intL.L.Ce oth . column). The adjusted Am(VI) feed is then injected on a bis (2,6 dimethyl 4- heptyDphosphoric (HDDiBMP) acid column previously treate n oxia y -b d dizing solution (S2O d Ag(IIg —an ) ions) until persistanc e darth kf eo colo r

168 Nitric eluotf from TBP column

Feed 1 T ———————— AmV2 ~ I 5 g.f ] H M HNO 3 0, ) POt 2.10"2M 1 Prtci pifation1 1 l^MlOHJjJ pVg ] 10"*M KjSjO, 0,1M

»« €> 1 ^ Washin( g (3^ Clution l Fittration HNOiO,3M —— T t , L T N,H4NO) 0,5 M H) PO* «''M 1,51- fi ~j * l HNO) IN J ————— l ^ HNO ————— * i l [Ag|lO- K2SjO, L J D Jîoli D u| iot * n 0,1 M 1 M(OH)j 0 ^\ 1 , ————— , , .,.,.,. .... ——— . /Spectrophoto- B K) CO) 5,5 M Oxidizing solution Imetri c control ü.

K, S, 0,0,25 M HNO,1IH HjPOiJ.IO^M l«>t Air valence P F J l ^ lO-'A, M[Ag]lO M KSS20, 0,2SM \^"' * \^/ ^d) ' 1 1 Precipitation *"^D Heating HjO ^ Ka AmOj ICOj], ^ 1 U^^Am SOj^^*J AmV-»Amin

LQ ,otiot )i cTn1 J Corhonott «ff lutnts / t?) \ f Htoting io 50« Fittratiol C g A n Ln 111 + Am A/ 50 ü 100 mg,!"' '——————————— ' last N H« OH ^ '"_ "- itotion 1 Am(>H)C | \

Figur 4 e Recover f americiuyo fro1 mm24 aged plutonium stocks Final purification process for americium 211

of Ag(II) ions in the effluent. Extraction of Am(VI) is apparent on the column by a greenish band. After washing of the column by an oxidizing solution americium is eluted with a reducing solution (N?H5NO3 = 0.5 M, HN03 = 1 M). Such a process is applied successfully on a 50 to 60 g. of 2*1 Am scale. The yield of the purification lies on the range 85 to 97 % for routine experiment. Americiu purit1 m 2* s measure f 98.yo wa 5% calorimetry db y after transformation into AmC>2. Laboratory experiments were performed in order to determine the stoichiometry of the extracted americium(VI) compound and thus the exchange capacity of the HD(DIBM)P columns. Figure 5 presents the curve obtained on the study of the saturation of a small HD(DIBM)P column with Am(VI), the concentration of 2*1 Am in the effluent is measured in line wit smalha l Geiger-Muller detector breae Th . k throug s obtainehi n a r dfo HD(DIBM)P/Am(VI) rati ostand A , equathu H * r HD(DiBM) f sfo o i t l e Pth formula of the extracted americium(VI) compound is = AmO2A2(HA>2. The same stoichiometr founs r yU(VIwa dfo ) extracted complex. Using this process f higo ,. hg abou 0 purit A1 50 mt 2* yhav e been produced.

m.3.2. Recover f uraniuyo fro* mm23 aged plutoniu stockS m23 s In aged plutonium 238 dioxide, uranium 23* accumulates. Due to the very different half live f o thess e isotope = 87,$ t 7( s year r fo s plutonium 238 and 2.**.10^ years for uranium 23*) the recovery of high nuclear purity uranium 23* necessitates very efficient separation process.

169 V(mt)

Figur e5 Saturatio f HIXDIBMjP/SiOno j column with 2*lAm(Vî)

Column = stationary phase = 0.500 g of HD(DIBM)P (29,5« %)SiO2

Aqueous flNaNO3 , [K, , IHNOM M M 2 S2Og1 2 2 0. 0. s )3 * != ] Phase [_£AgJ = 2.1Q-2M, CAm(VI)J * 3.27 g.l-f Flow rat ml.h-16 e= - 60 keV 2*'Am detection by in-line Geiger Müller gauge

Ten grams of plutonium 238 on the dioxide form were processed for uraniu recover4 m23 y accordin followine th o gt g schemes (i) Dissolutio f PuOno HNOn 2i 3 aci dactinn witio ~ s catalystgha F . (ii) Precipitatio e bul th f plutoniuo kf s o noxalatea 8 m23 , remains in the mother liquor. (iii) Change of medium by Pu(IV), U(VI) coextraction with TBP and back extraction with the addition of alpha bromocapric acid to the organic phas aqueouN l e = toward 3 s O solutionN sH » (iv) Extraction chromatography on HD2EHDTP (20 %), TBP (10 %)/Gas Chro columnmQ .

The extraction Chromatographie separations between 238p d 234^ u y were performed usin gcoluma n containin f stationaro g 0 g2 y phase. During e loadinth ge feed th ste f ,o p extractio f 2o 3*U(VIn s realizei )e th y b d synergetic mixtur ethy2 i d l e: hexy l phosphorodithioic acid (HA)/tributyl phosphate extractee Th . d complex which formul ; UO2A2.TB s i a P possesses an intense yellow color [17] . The di 2 ethyl hexyi phosphorodithioic acid possesses reductive propertie s Pu(IVv s ) thus plutoniu escape8 m23 s from e columth s blua n e Pu(III) fractioe e A topcolum. th th s t )f i o (a nn destroyed by this reaction. After washing of the column with a solution of composition HNO3 = 1 N, NH2S03H = 0.1 M, 23*u(VI) is eluted with a 0,4 M (NH^C^ solution. Forty six milligrams of uranium 23* were recovered and after two cycle f extractioso n Chromatographie separation e alphsth a contributiof no plutoniupurifiee th n i 8 d muraniu23 s onl wa y 4 1.2m23 . 7% Decontamination factors for extraction Chromatographie separations were respectively : cycle 1 F.D. U(Pu) = 517 ; cycle 2 F.D. U(Pu) = 452. For the entire process F.D. U(Pu) = 4.5.107 was obtained and the uranium 234 recovery yield was 75 %.

170 IV CONCLUSIONS Extraction chromatography techniqu s developpewa e e lasn th te t n i d year r actinidefo s s recovery, separation d purificationse an sth o t e Du . large variety of organic extracting ligands available, that technique can be applied successfully to solve difficult actinides separations. The processes developped for the recovery and purification of microgram to kilogram amount f actinideso s isotope e eas sar operato yt e eithe glovn i r e boxer o s t cellsho in . Nevertheless ther s i stile l some difficultie o solvt s e: especially the extraction of trivalent americium from nitric acid solutions for whic f somi h e opportunities exis : tuse f DHDECMo s , theiW r o Pr performance t reallno e y sar ideal . More researc needes hi thin di s area.

REFERENCES

1 BRAUN, T., GHERSINI, G. Extraction Chromatography, Elsevier, Amsterdam (1975) 2 BUIJS , MULLERK , , REULW. , , TOUSSAINT3. , , 3.C. 504eR ,EU , (1973) 3 ESCHRICH , OCHSENFELDH. , . SeparatioW , n Scienc Technod an e - log *(19805 y1 ) 697. 4 WENZEL , MERZU. , , RITTERE. , . SeparatioG , n Scienc Technod ean - logy 15 it (1980)733. 5 MARTELLA, L.L., NAVRATIL, 3.D., SABA, M.T. "Actinide Recovery from Waste and low grade Sources" Harwood Academic Publishers, Chur, London, New-York (1982) 27. e ISAACM 6 , L.D., BAKER, 3.D., KRUPA, 3.F., MEIKRANTZ, D.H., SCHROEDER, N.C. A.C.S. Symposium Series 117, (1980) 395. 7 KNIGHTON, J.B., HAGAN, P.G., NAVRATIL, J.D., THOMPSON, G.H. A.C.S. Symposium Serie (1981)1 16 s. 53 , 8 KOEHLY, G., MADIC, C„ BERGER, R. Proceedings of the Interna- tional Solvent extraction Conference, The Hague. Society of Chemical Industry, London (1971) 768. e entirth r efo 9progra f productiomo f 238p: o B.I.S.Tn e se u . 218, (1976). 10 SONTAG , KOEHLYR. , . B.I.S.TG , (19768 . 21 . 23 ) 11 BOURGES, a., MADIC, C., KOEHLY, G. A.C.S. Symposium Series 117, (1980)33. 12 KOEHLY, G., BOURGES, J., MADIC, C., SONTAG, R., KERTESZ, C. A.C.S. Symposium Series 161, (1981) 19. 13 GUILLAUME . PersonaB , l communication. 14 GUSTAFSON, R.L., MARTELL, A.E. 3. Phys. Chem. 67 (1963) 576. 5 1 MOSKVIN, A.I., KHALTURIN, G.V. GEL'MAN, A.D., Radiokhimiy1 a (1959) 1*1. 16 MADIC, C., KERTESZ, C., SONTAG, R., KOEHLY, G., Separation Science and Technology L5 4 (1980) 745. 17 FITOUSSI, R., Ph D Thesis (1980) CEA-R-5152 (1981)

171 SYNTHETIC INORGANI EXCHANGERN CIO S THEID AN R POTENTIAE LUS NUCLEAE INTH R FUEL CYCLE

A. CLEARFIELD College Station, UniversityM Texad an sA , Texas, United State f Americso a

Abstract This paper describes the structure and principal ion exchange characteristics of acid salts of polyvalent metals and hydrous oxides. The acid salts can be prepared in crystalline (layered) form, as amorphous gels and in intermediate stages of crystallinity n exchangIo . e behavio thes i r n relatee th o t d crystallinity of the exchanger. Separations of importance to nuclear technolog alse yar o described e aciTh .d e saltb n sca classified into five types based upon structure. The hydrous oxides are considered to fall into two main types. Those such as

Zr02, Ti02 and Sn02 consist of small particles of oxide with hydrated surfaces and interparticle water. The nature of the surface is described and related to the ion exchange properties of

the hydrous oxide. Other hydrous oxides suc Sbs ha 2 MnQd 0ce jan ar 2 describe pyrochlore termn th i d f o s d spineean l structures, respectively suggestes i t I . d thapropertiee th t thesf so e exchangers can be controlled to greater specificities through synthetic procedures.

1. Introduction phenomenoe Th exchangn io f no inorganin i e c compoundw no s i s known to be very widespread. Among the major classes of compounds y citma e e w 1. clays 2. zeolites 3. hydrous oxides 4. acid salts of polyvalent metals . heteropol5 y acid salts 6. synthetic apatites . 7 insoluble ferrocyanides 8. insoluble sulfides and sulfates 9. condensed phosphates and titanates In addition a large range of miscellaneous exchangers is known so tha systematizatioa t n base structuran do l principled an s

173 exchange behavior is highly desirable. In this paper we shall confine ourselves to the acid salts of polyvalent cations and to a lesser extent to certain hydrous oxides. Even here a bewildering array of crystalline and amorphous phases are known. Attempts at systématisatio directionw ne d n an r synthesisfo exchangerf so s with greater specificities will be described. The original discovery and subsequent exploitation of acid salts of polyvalent cations grew out of the need for ion exchangers with greater stabilit elevateo t y d temperature effecte th d f so san radiation damage [1,2]. These compounds were initially prepares a d amorphou sn extensiva geld san e literatur n theseo e gels developed [3,^]. In 196H gelatinous zirconium phosphate was crystallized and oH 01 its layered nature demonstrated [5]. This compound, Zr(HPOjj)2 2 beew no n s designateha o-ZrPs a d . Shortly thereafte secona r d

variet zirconiuf o y m phosphate, Zr(HPOjj)2*2H s crystallize20,wa d interlayen a s ha t [8]I r. distanc f 12.2eo A compare 7.6o t r d Afo o-ZrP and is very likely a different layer type. It will be referred to as Y-ZrP. In the intervening time period, a large number of layered inorgani exchangern io c s have been prepares i o tha s t i t] [7 d useful to classify them by structure type. In this we follow the scheme propose Alberty b d i [8] Exchanger) :(1 s with a-layered structure ) Exchanger(2 ; s with Y-type layered structure) (3 ; Exchangers with fibrous structure; (H) Compounds with a threedimensional structure ) Aci(5 d d saltan ;unknowf so n structure. . 2 Exchanger a-Layeree th f so d Type 2.1 Preparation A listin e compound th f som o gf o e s with a-layered structures i s given in Table 1 [8]. They are prepared from their respective gels by refluxin stronn i g g phosphori arsenir co c acide Th . crystallization is gradual and the degree of crystallinity of the products depends upoe concentratioth n f aci no reflue dth used xan d time [5,9]. This is shown in Fig. 1. To achieve an average crystallite size of 1-2ym for o-ZrP requires 1U days of refluxing in 12M H-sPOjj. The process may be greatly speeded up by autoclaving M aci U t 200-250°C1 a d n i » Large crystal e prepareb n ca sy b d dissolvin titaniua g r zirconiuo m d addinan m F gH sal n H^PO^i t , followed by slow evaporation at 60°C [10-12]. For most purposes small crystallites refluxinf o r h ,e obtaine 2 ar y suc 1- gma s ha n i d be used.

174 Tabl . e1 Som e Importan ravalent e salt d t f ad tso t metals with layered structure of the a-type.

Compound Formula Inter-layer Ion-exchange distanc ) capacit(A e y meq/g

Titanium phosf»hate T1(HP04)2-H20 7.56 7.76

Zirconium phosiphate Zr(HP04)2-H20 "7.56 6.64

Hafnium phosptlate Hf(HP04)2-H20 7.56 4.1.7

Germanlum(IV) phosphate Ge(HP04)2»H20 7.6 7.08

T1n(IV) phosptlate Sn(HP04)2-H20 7.76 6.08

Lead(IV) phosf»hate Pb(HP04)2-H20 7.8 4.79

Titanium arsertäte T1(HAs04)2-H20 7.77 5.78

Zirconium arsi•nate Zr(HAs04)2-H20 7.78 5.14

T1n(IV) arsen e iit Sn(HAs0 8 47. )2*H20 4.80

Get

i i i t i t i i

0.5=48 A.

i i i i i i i i

A 0-8=48

i i i i i '" "i ' t " ' ' i

2.5:48 A

\\fi\i\T

J^_JLjL__i__ . _ i - - i i i i r • T - • i

4.5=48 1 |

. X -II A JL i i i i i i i i

604 I"I ' 1 002 ...12:48 A .A1 Kir A1 1 1 I I 1 1 1 1 0 1 0 2 0 3 0 4 ANGLE 20 X-ray powder pattern a-zirconiuf so m phosphate differenf so t degreef so crystallinity. Numbers above each pattern indicate concentration of H,PO, d tim an f refluxeo . (Fro mwit, 9 Refh. permission.) . 1 Fig.

175 Amorphous gele preparear s rapiy b d d additio solubla f o n e salt of the cation to the acid. Even if the acid is present in excess, the rati phosphatf o arsenatr o e M(IVo t e less i ) s thao [U]ntw . Stirrin r severafo g l hour heatinr so requires i g obtaio t d na stoichiometric product non-stoichiometrie Th . c e solidb n ca s

represente M(IV)(OH)s a d z(HXOZ . S Jt )0 2-z/2 d »YHan s A 20 , wherP - eX exchangn io e Th e . behavio2 S othed an r r propertie f thesso e acid salts depend upocompositioe th n degred nan f crystallinityo e . This wil discussee lb morn i d e detail followin discussioa g e th f no a-structure. 2.2 Structure of the a-layered compounds e crystaTh l structur botf eo h o-zirconiura phosphate [13»1*d an 0 arsenate [15] have been determined. The layers consist of metal atoms which ver plana y n nearli e e (±0.25li y A froplanee th m d )an are bridge phosphaty b d e (arsenate) groups. Thes situatee ear d alternately above and below the metal atom plane. Three oxygen atom f eacso h phosphate grou bondee ar p o thret d e different zirconium atoms which form a distorted equilateral triangle. The coordination of the metal atoms is therefore six-fold. The adjacent layer foro arrangee x sidet ar ssi ms a d o cavities d s a s show Fign oxyge e i n. .th 2 Eac f nho atom bondet metasa no o t dl atom bear protosa d formnan cavite s th par f yo t walls e P-OTh .H groups hydroge watee n th bon ro t dmolecule s which e residth n i e center of the cavities [16]. There is just one mole of cavities per formula weight and the free volume into the cavities is that of an unhydrated K+.

9Ï*

Fig. 2. Idealized structure of a-ZrP showing one of the zeolite like cavities created by the layer arrangement. (From Albert!, C. in "Study Week on Membranes," Fassino, R., Ed., Fontificiae Académie Scientarum Scripta Varia, Rome, 1976, 629, with permission.)

176 pH

I 234567 meq OH"/ g a -ZrP

Fig. 3. Potentiometric titration curves for alkali metal ions on a-ZrP (12:336)s curvC obtaines e wa eTh a les . n so d crystalline exchanger. (From Ref. 17, with permission.)

2.3 Ion Exchange Behavior of Crystals exchangn io e Th e behavio a-Zrf o r P crystal conditiones i s y b d e tighth t smale bindine layerth th l d f sentrancewayo an g s inte th o interlayer cavities. This is illustrated by the ion exchange titration curves of Fig. 3 [7,17]. Li+, Na* and K* are exchanged in acid solutio o fort n m half-exchanged phasee typth e f so

ZrMH(POn)2'XH20. The ions take up positions in the interlayer spaces which requires diffusio f unhydrateno d ions intcrystae th o l lattice. Thus, the selectivity favors the more weakly hydrated cation viz. K+ > Na+ > Li*. Rb* and Cs* do not exchange in acid solution since ,unhydratee th eve n i n d conditiono ,to thee ar y large to diffuse in between the layers. However on addition of thoughs basei t i , t tha~ removeOH t s protons frolattice th m e causin layere th g opeo st n somewha permid tan t diffusioe th f no 1 larger ions alkalinn I . e mediu 3/a m * phase ZrMQ c(POjj)'XHcH ^ 20 is formed rather than the half-exchanged type [17]. Exchange of ions into the crystalline layered group(IV) phosphates result discontinuoun i s s change e interlayeth n i s r distances. The interlayer distance depends upon the type of phase, i.e., half, three-quarter or fully exchanged, the size of the entering amoun e catioth watef d o t an n r take [7,18]p u n . Thuse ,th half-exchanged sodium phase has 5 waters of hydration and an interlayer distanc f 11.8eo A whil fulle th e y exchanged phase,

177 Zr(NaPOJ+)2*3H20 has an interlayer spacing of 9.8A [19]. The hydrated sodium ion is too large to enter between the a-ZrP layers and therefore must she waterds it som hydratiof f seo o diffuso t n e inte layersoth . Rehydratio wit* Na he expansioth f e o nth f no layers then occurs for the pentahydrate. Diffusion begins at the edges and works toward the center so that at intermediate stages of exchange both the unexchanged or proton form and the exchanged phase are present in the same crystal [18]. This mechanism results in irreversible exchange behavio e forwarth n i rd versue sth backward directions [7,18]. e differehceTh exchangn io n i s e behavio f crystallino r e group(IV) phosphates which have an a-layered structure arise from the different size f theiso r unit cell dimensions e smalleTh . r these dimensions, the smaller the dimensions of the openings into the cavities greatee th sterie d th ran , c hindranc o cationst e . Furthermore, as the product of the cell dimensions within the decreases) b layer x a ( sdensite ,th fixef o y d charges (i.e. phosphate groups) increases .creating a greater barrier to ion diffusion. Therefore, higher pH values are required for exchange to occur, i.e spreao .t layere th d se sal apartth t n formI . f so the exchange forcee th r s holdin layere th g s togethee rar essentially coulombic. likelThuss i t ,i y that these forces become greater as the density of fixed charges increases. The stronger the forces holding the layers together, the more rigid they become more anth d e exchangere likelth e ar y exhibio st t ion-sieve effects [8], Froe foregointh m evidens i t i g t that cation exchange th n o e a-layered acids wilt occu no lhighln i r y acid solutions. Under milder conditions where exchange takes place exchange th , e behavior becomes complicated because of the formation of a second phase. Thi usualls i s half-exchangea y d cation phase largea whic s hha r interlayer spacing thaoriginae th n l a-phase. a thirThu f si d cation is present, it may interact with the half-exchanged phase to give complicated phase relationships. A number of studies have been carrieelucidato t t ou d e these phase relationshipa s sa necessary prelude to Chromatographie separations [20,21]. Other approaches to utilizing the crystalline phases in Chromatographie procedures involve expanding the layers prior to exchanging cations. These procedures will be discussed subsequently. 2.4 Ion Exchange Behavior of Gels Amorphous gels of zirconium phosphate behave very differently from the crystals. In the early literature very little attention

178 was paid to preparative methods and to proper characterization of the gels. This led to irreproducible behavior. The reasons for welw no l thi e understooar s d [M,22] durinf I . g preparatiol mo e th n rati f phosphato zirconiuo t e less i m s than two, precipitation still produce occurs th hydroxyphosphatea t s i t;bu , n sucI nn lo Q e tn 3 Zr(OH) (HPOjj)2_x/2« exchangS ^- e behavioe th f o r hydroxyl groups must be considered as well as those of the monohydrogen phosphate groups. Even whe ratie th n f phosphato o t e zirconium exceeds two in the reactant mix, the precipitate which first forms may have a ratio lower than two [22]. However on aging e precipitatth mothee th n i re liquor s phosphatit , e content increases and in the limit P/Zr - 2. Ratios higher than this indicate sorbed phosphate [M]. Greater resistance to hydrolysis and more reproducible behavior result if the gel is refluxed in about 0.5M H^POii for 10 hours or more [9]. When dried thoroughly over P refluxe2e 0cth d gels havsame th ee compositioe th s na crystals t the,bu y swel waten i l r takinadditionan a p u g mole4 l s of HpO. This property of the gels allows them to exchange rather large cationinterlayee th s sa r distance e greatlar s y increasey b d the swelling. A typical Cs+ titration curve for a gel that has been refluxed in 0.5M H^POjj for 48h is shown in Fig. *». In contrast to the crystals where the pH remains relatively constant at ca. 7.5-8 with ion uptake, each increment of Cs+ exchanged results in an increas break. o n TherpH e n r endpointi ear so curve th d ean n i s the process is reversible [9]. This is similar to results obtained with weak acid organic resi exchangern io n s except thaeffece th t t is much more marked t indicateI . s thacatioe th t uniformls i n y distributed in the gel to form a solid solution rather than clustering in adjacent cavities to form a new phase as in the crystals behavel ge weaa e s ka sTh field exchange sense th f en o i r Eisenman'+ Na > s* K theor > * selectivite yRb th > [23 d + ]an Cs s i y > Li* [24]. As the cation content increases, the selectivities change gradually in the way predicted by Eisenman's theory, that is, to the strong field situation with Li* > Na* > K+ > Rb* > Cs+. e interlayeTh r distance attain maximusa relativelt a m catiow lo y n uptake and then decreases with further loading [9]. Thus the cations are brought closer and closer to the negative sites of the layers with consequent increase in field strength. 5 Semi-crystallin2. e Exchangers By refluxing a gel in phosphoric acid of different concentrations and for different lengths of time, it is possible to build in any desired amount of crystallinity [5,9]. The particle

179 Fig. 4. Potentiometric titration curve for Cs uptake (Titrant 0.1N CsCl+CsOH amorphoun )o s zirconium phosphate (0.5:48). Solid lin left ea phosphats i t e releas e solutionth o et . (From Clearfleld Oskarsson, ,A. Exchangn Io . ,Membranesd A ean . J_, (1974) 205, with permission.)

refluxee th siz f eo d product increases with increasn i e crystallinity e titratioth d ,an n curves chang a continuou n e-i s fashio l typ no thage t e fro e t th mexhibite crystale th y b d s [4,9,25]. crystallinite th Thi shows s i sA Fign ni . 5 . y increases the solid solution range decreases for each of the phases -formed. For example, with a gel that had been refluxed in 4.5M H^PO^ for 48 hours, Na* is initially taken up with no phase change. At a loading of 0.25 meq/g the half-exchanged phase appears and increases in amount relative to the sodium containing ot-phase until at 3 meq/g load it is the only phase present. The composition range for the half-exchanged phase is approximately 3~3-45 meq/g. This may sound like a contradiction in terms, but by the half- exchanged phase for a semi-crystalline preparation we mean, a solid whose X-ray pattern resembles that of the true half exchanged phase, ZrM^HCPO-XHO, but with broadened reflections.

180 11

. 5

012345678 Meq (OHf/g . Fig5 Titratio. d intermediatn an curve w a-Zrr lo fo sf Po e crystallinity: 0.8:48 -O , 2.5:48 -A, 4.5:48 -ft, 12:48 -|. (From wit, Ref9 .h permission.)

6 Catio2. n Separations Aguida s carryino t e t catioou g n separations several comprehensive determinations of distribution coefficients for amorphous zirconium phosphate have been published [26,27]. Similiar values for amorphous zirconium tungstate and molybdate were included [27]. The trouble with this data is that the exchangers were poorly characterized so that the results serve only as a rough guide. The effect of crystallinity of the exchanger on. measured distribution coefficient divalenf so t ions measureswa y b d Tuhtar [28] results .Hi dramatiA showe sar . Fign 6 i n. c decrease

in Kd as a function of increasing crystallinity is observed due largelexclusioe th o t y f highlno y hydrated cationsmale th ly b s passageways betwee layerse th n e samTh .e tren exhibites i d y b d tetravalend trian - t cations [28]. A rather complete summary of possible cation separations by group(IV) phosphate gelrecentls ha s y been published [29] thin .I s pape e shalw r l confine ourselve o thosst e separations relateo t d nuclear processes. s bee^'Cha ns separated from nuclear reprocessing'solutions [30,31]. Distribution coefficients and some

181 lOOOO Co11 O Zn" O Mn" A Ni" A Co"

1000

100

0.3:48 45:48 9:48 12:96 CRYSTALLINITY

Fig. 6. Variation of distribution coefficient (at almost zero load) with increasing crystallinit a-ZrPf yo .

Chromatographie data for recovery of Cs+ and Sr2* from reprocessing solutions has also been given [32,33]. Separation of Ba2"1" from other divalent cation s[35* [34Cs bees ]d ha n ]an reported . Amorphous zirconium phosphate has received a great deal of attention for separation of radioactive ions because of its resistance to ionizing radiation, oxidizing media and strong acid solutions [36] sorptioe .Th tracef no r concentration Am(III)f so , Cm(III), Bk(IV), Cf(III), Ce(III), Eu(III) and U(VI) on ZrP with P/Zr - 1.3H was investigated at 75°C [37]. It was found that separation factors for ions with the same charge are low, whereas thos r ionefo s with different oxidation states were s largewa t I . suggested that Bk(IV) coul separatee b d d frotrivalene th m t ions. A good separation of americium from curium had also been reported [383. A number of procedures are based upon the fact that better separations are effected on zirconium phosphate if the cations differ in oxidation state. For example, Am(III) is separated from Cm(III firsy )b t oxidizin Am(V)o t e Cm(IIIt i gTh . s )i preferentially retaine colume th n no d [38-40]. This general procedure has also been applied to neptunium- Plutonium separations [4M], Np(V) - Pu(IV) - Pr(III) [42] and

182 Ce(IV )Pr(III- )Cm(III- ) separation sbees [43]ha n t founI . d that separations which take advantage of different oxidation states mafurthee b y rnon-aqueouf o enhance e us y b d s solvents [44]e Th .

lowee sharth rpr valenfo declin d K tn i estat e contrasts wite th h increas highee maximua th o t er r mfo valen talcohoe statth s ea l content goes to 50$. These differences are attributed to the relative stabilitie e chloro-complexeth f so o iontw ss e formeth y db in the changing solvent composition. In fact it has led to a metho determininf o d stabilite th g y constant f thesso d relateean d uranium complexes as well as improved separations [45, 46]. Distribution coefficients for Th, Pa, U, Np and Pu were also measured and procedures for their separation outlined [4?]. Separation of uranyl ion from several di- and trivalent cations has also been described 2.7. Nuclear Waste Processin Wated an g r Purification. originae Th l w inorganisearcne r exchangerhn fo io c s swa instigate desire th r materialy fo eb d s stabl n intensi e e radiation fields and at high temperatures. Early efforts in this direction have already been mentioned [1-3,35]. Zirconium phosphate has indeed been found to be stable to radiation [49,50], but exhibits increased hydrolysi elevatet sa d temperatures. Resistanco et hydrolysis is increased by use of semicrystalline zirconium phosphat shows a e Ahrlany nb Carlesod an d n [51]. mixea Eve , dnso hydrous zirconia-zirconium phosphate column was used to prevent phosphate bleeding during purificatio f diluto n e solutions containing transition metal ions. A method for separating 1'°'C 37 s from fission products has been used by the French Atomic Energy Commission which utilizes a mixed ammoniud an P Zr m be f phosphotungstato d e [52] have W . e already mentioned the separation of uranium from plutonium on amorphous zirconium phosphate with P/Zr - 1.34. The effect of added fission products was also assessed. Other interesting applications of zirconium phosphate include the remova 10f 95°Rlo d Z uran from purex solvent [53d ]an separatio parenf no t -daughter radioisotope 11- ^d mln^C s ,suc s ha ^Ce- 1J+V and210 Pb - 210Bi [54]. n ExchangIo 8 e2. with Expanded Layer Zirconium Phosphates Froearliee th m r discussio n crystallinno e evidena-Zrs i t Pi t that its usefulness for separations is severely limited. However, it turns out that it is relatively easy to increase the inter lay er distance of ot-ZrP by a number of processes. In the expanded layer forms the barrier to diffusion of large cations is drastically

183 reduced. Therefore, exchange processes can take place without the addef o e d us base half-exchangee Th . d sodium form,

ZrNaH(P04 )inter-laye2n '5Ha s 20ha r distanc 11.8f eo A [19]t .I behave monofunctionaa s sa l exchanger wit capacita h 2.5f o y 3 meq/g [55]. Kornyei and Szirtes have reported [56] the separation of

strontium from radioactive wastewater on ZrNaH(POjj)2«5H20, If sufficient acid is added to remove the sodium ion from

ZrNaH(POj,)2'5H20 ,obtainet a-Zrno s rathePi t bu d highea r r hydrate containing 5-6 moles of H20 and named 0-ZrP [19,57]. This exchange n interlayea s ha r r spacin f 10.4o g o thaA s acceptt i t s large cations more readily than ct-ZrP. This is shown in Fig. 7 where potentiometric titrations for alkaline earth cations on a-ZrP and 8-ZrP are compared [58]. As yet no cation separations have been carrie witt ou dh 9-ZrP thit ,bu s woul fruitfua de b see o t ml area for investigation. PERCEN CONVERSIOF TO N 5 7 0 5 25 100

2.0 -

0 5. 0 4. 0 3. 0 2. I.O UPTAKE meq M2* Fig. 7. Potentiometric titration curves for alkaline earth cations on 6-ZrP (solid lines a-Zrd )an P (dashed lines). (From Ref wit, 58 .h permission.)

Anothef expandino y wa re layer th g amina-Zrf y so b s ePi intercalation [59]. For example, in the butylamine intercalate the interlayer spacin s 18»6gi d thi an s largA i s e enougt complepu o t h x ions in between the layers [60,61]. In this case the amine acts to keep the pH high so that no added base is required to load ions into the exchanger. Up to U-5 meq/g of divalent ions may be exchanged and multiple ions can be loaded. Thus it would seem that this could be the basis for many cation separation techniques.

184 . 3 Exchanger Y-Layeree th f so d Type Onlmembero tw y f thiso s serie knowne sar Y-Tid ; an Y-Zr P] [6 P [62]. These compound merelt no e y sar mor e highly hydrated formf o s o-compounds but have different, as yet unknown, types of layers [63,64]. Becaus interlayee eth r distance thesf so compoundo etw s are large, 12.2A and 11.6A, respectively, they are able to exchange large cations and complexes [65,66]. However due to the closeness fixee oth f d charges, hydrolysis becomes sever higt ea h levelf so exchange. A highly crystalline form has recently been prepared [67 ]lese b whic sy susceptiblhma hydrolysiso et selectivite Th . y sequence at low to moderate loadings as obtained from titrâtion * [65]Li > . I + vie Na f thi o w> s* sequenceCs > * Rb ,> * K dat s i a it is surprising that an extremely high affinity for Cs* by Y-ZrP has been reported [68], high enough to recommend its use in reactor water cleanup. . 4 Acid Salts With Fibrous Structure

Member thif so s group include cerium(IV) phosphate (CeP) [69] f thorium phosphate (ThPf) [70] and titanium phosphate and arsenate [8,71]. These fibrous compounds have been use preparo t d e inorganic ion-exchange papers r thi,o n layers, suitabl fasr fo et separation f cationso s [72] fibere . Th preparee sar addiny b d g

dropwise 0.05M Ce(SOij)2-4H 0.5n 20i M HjSO wela o ^lt stirred solution of 6M H^PO^ previously heated to 95°C. Flexible sheets e obtaineb n ca dispersiny b d fibroue th g s precipitat larga n i ee volume of water, then filtering the slurry in a large Buchner

funnel e formulTh . CePf ao probabls i f y Ce(HPOjj)s 2it *H t 20,bu structure is unknown. . 5 Acid Salts with Three-Dimensional Structure This class includes compounds witgenerae th h l formula y an - A d an s A , P AM- 2X (IV)(X0 ; Ge , 1|Th ) 3, [(M(IVZr , Ti )- univalent cation]. We have prepared these compounds by thermal decomposition of exchanged forms of ct-ZrP [7,19] and by hydrothermal means [73,7*0. HZ^CPO preparee b y ^ ma treatiny b d g LiZr2(POjj)o wit thermar ho aci] [8 dl decompositio NH^Zr^CPO^)f no , [74]. The structure of these triphosphates is shown in Fig. 8 r bridge Z r througZ builf s o [75]i o t p dt u t I .phosphat 3 h e groups. This arrangement produces M(IV) octahedra linked by cornertetrahedre th o st whole th a e formin ga thre e dimensional network containing cavitie typeso tw s f .whico e har There are four cavities per formula, one of which is occupied by Na+. Silicate can replace part or all of the PO^" and this requires additiona + ion lNa progressivelo t s remainine y th fil p u l g cavities [76,77]. The silicate ions enlarge the cavities sufficiently to allow relatively free movement of the univalent

185 Fig. 8 Projectio. half nunie o th ft cel NaZr.(PO,)f lo plane,c a onte th ,o showing the cavity built up by the Zr.CPO,)- groups. Numbers next to the atoms indicate height (in hundredths) above the zero level plane. (From Ref. 76, with permission.)

cations phosphate wheth nf o abou 3 e t2/ group replacede sar e Th . Na+ ions can now be exchanged by Li4" or H*„ Similiar Li^-H* exchange reaction e possiblsar e with HZrfPOjjK. Sinc cavite eth y d passagewaan y size exchangd san e capacit modifiee b n ca y d ovea r broaanio^ substitutionXO d dne an rangth , choicy e b M r f eso fo shoult i , possible Mdb preparo t e e exchanger thin si s system which are highly selective and stable to high temperature, radiation fieldbroaa d dan s. ThirangpH sf o e ide extendes i a o t d other systemnexe th t n i sections . 6. Hydrous Oxides and other "Amorphous" Exchangers

A number of hydrous oxides including Ti02» Zr02, Mn02 and antimonic acid have bee nuclean ni examinee us r r processesfo d . The exchange characteristics of these compounds has been summarized and reviewed in a recent book [73. Hydrous titania readily sorbs uraniu mbees [78ha nd considere]an extractior fo d uraniuf no m from se'a water [79,80]. Hydrous oxides were alse o th use n i d purificatio d isolationan neptuniumf no , plutonium [81,82] americiura and berkelium [83].

186 A B . Fig9 Schemati. c representatio hydrouf no s oxide particles . ZrO.'nH-A . O wit hintercrystalline th hydroxy n i n lio e water neutralizine th g nHn i 0 wit hH positively charged particle surface SnO. ;B 2" 2 ° 3 the intercrystalline water neutralizing the negatively charged surface.

A simple structual mode hydrouf lo s Ti0e oxideth , f Zr0o s , 2

Sn0 type has recently been given [84].2 The solid precipitated by ammoni2 a consist particlef so . 25-30sca diametern i Ae th n I . interior of these particles the structure is that of a crystalline anhydrous oxide [85]. These small particle joinee sar d togetheo t r form agglomerates containing large cracks and fissures. The full metal coordinatio preserves i n surface th t partiay a deb l coordinatio e remainin HpOf Th no . g water molecules forn a m interparticle solution, the acidity of which depends upon the degree of protonation of the surface oxygens. By aging or refluxing it is possible to obtain more crystalline forms of these oxides detaileA . d descriptio w amorphouho f no d crystallinsan e hydrous zirconi formee aar d from solution tetramerie th f so c zirconyl ion has been given by Clearfield [86]. The formula of these hydrous oxide bess i s t represented [84]s a 1| (2a ( b) [MOa(OH)b(H20)R^ assumptioe a+bth ] n ~O "e . nS 'th tha+ l al t metal atoms are fully coordinated, the oxygen-metal ratio, R, will be greater than 2. For the fluorite structure (Zr02) a 25A cubic particle will e extr hav2.15- Th R ea. oxygen surface atomth t sa e must be supplied by water molecules. The remaining water molecules, S, form the interparticle solution. Whether S is acidic

(Sn02) or basic (Zr02) depends on the degree of protonation of the oxygen particle th t sa e surface. Thi illustrates i s d schematically in Fig. 9. In the basic limit, the surface oxygens are supplied by metad terminaan l 0 bridgin2 lH acidie th g n OH~ci limi; e th t oxygens are bare 02~ ions making the particle charge negative with compensating HoO* in the interparticle solution.

187 . \? « 0 1 9 8 7 6 5 4 3 2 I pH Fig. 10. Ion exchange capacity versus pH for hydrous tin and zirconium oxides.

Based on the foregoing picture of hydrous oxides, hydrous tin oxide should readily exchange cation acin i s d solution because H^O* is presene interparticlth n i t e solution morA . e basic oxide such as that of zirconium would require higher pH values to exchange cations because the OH" in the interparticle water must be neutralized first before surface protons will detach froe th m surface oxygens. The opposite is then true for anion exchange. This is illustrated in Fig. 10 [35]. The exchange capacity for cations will increase wit e limihth increasinn ti dependt bu H p gs upototae th n l removable protonsurfacee th e n selectivitso Th . y particulaa r fo wiln onlt io rlno y depend upon these factort sbu also the complex formation constant of the surface-ion complex. n contras I foregoine th o t g picture crystalline antimonic acid exchanges only cation d exhibitsan anomaloue th s s selectivity sequence [87] Na+ > K+ > Rb* > Cs+ > Li*. This is attributed to the fact that antimonic acid has a pyrochlore structure with an 2 (SbgOg) "" framework with almost 3H20 distributed over the ApO* interpenetrating framework watee [84Th r] . molecules woule b d b site8 d san * witA d o protone I6 tw hth e n i sth n expectei e b o t d sites. Thus, the formula SbjO^'S^HgO is structurally represented as (H^O^SbjOg'O.gHj analogn Oi y witAe 2Bth h2 Og pyrochlore representation. There is no interparticle water as it is all in the crystal lattice. The theoretical exchange capacity, 5.08 meq matche, g experimentae th s „ Furthermore ~ g q lme one1 5. ,s antimonic acid behaves as a strong acid exchanger with a relatively constant exchange capacity ove largra rangeH ep . Recently a layered form of HSbOg'XHgO was prepared by ion exchange froilmenite th m e for KSbOf o m ^ [88]. Similiar reactions were carried out to prepare HNbO^ and HTaO^. When the starting phases were the pyrochlore thallium salts, heating them in 3N HNO,

188 Fig. 11. Potentiometric titration of hydr.ous MnO_ by the static method. Titrant 0.1M (MCI + MOH) with M - Li* -Q . Na* - A and K* - Q . converted them to the corresponding acids HNbO^'XHgO and HTa03«XH20 with retentio pyrochlorf no e structure [88], Howeve lithiuf i r m niobat r tantalato e e were use s startina d g phase acie th sd products ha perovskita d e structure [89]. We have recently examined hydrous manganese dioxide prepared electrolytically [90] exchange ,Th e capacity depends upoe nth temperature of preparation but for most samples is about U.6 meq/g. The exchanger behaves in some respects as a hydrous oxide, its capacity changing with pH; but it also exhibits ion sieve behavior. This is shown in the titration curves in Fig. 11. The exchange capacity for K* is about 1/3 the theoretical capacity and much lower tha . nHoweveK* thar r eitheo acifo t n i * r drNa solution, the distribution coefficients indicat selectivite th e y sequenc* eK >Na Li**> e .interpreW t this dat indicato at e thaexchangee th t r contains cavities with different sizes largese Th . t cavities, perhaps surfac e thosth n eo e prefe largese th r t unhydrated ion. ,K* A second grou f intermediato p e sized cavitie n accomodatca s n i * eNa acid solutio d finallnan basin i y c solutio become* nLi mose sth t preferred cation. However when the fully exchanged Li* form of hydrous manganese dioxide was heated at 500°C, it formed a weakly crystalline A-Mn02 and on washing out the Li* with acid showed a remarkably high selectivit r Li*fo y . This A-Mn0 relates e i 2 th o t d spinel LiMn201( [91] and Thackeray, et al. have shown that Li* can be replaced by H* in this spinel [92]. All of this suggests possible synthesiroutee th w inorganio ne t s f so exchangern io c s

189 with high selectivitie r individuasfo l iond greasan t temperature and radiation stability e havW . e mentioned that pyrochlores, spinels, illmenites, perovskites and triphosphates can all be converted to acid forms and that certain hydrous oxides can take on these crystalline structures synthesiziny B . g man thesf o y e phases hydrothermally, or from gels, in the presence of a particular ion it may be possible to alter the structure or size of the cavities or tunnels to be highly specific for ions which easily fit into the structure. A case in point is zirconium phosphosilicate. It is prepare treatiny b d zirconiuga m silicat witl ege h H^PO^ [93]° This exchanger has been examined at length for Sorption and separatio lanthanidef no sseparatior fo [94 d ]an neptuniuf no m from berkelium [96]. Decontamination factor wer0 20 300ef sd o 0an obtaineseparatioe th r frou P fo d m f no ^ 5Z r^+ 10 5d N°Ruban , respectively [96] large .Th e chang distribution i e n coefficients which this exchanger exhibits with pH change suggests its amorphous hydrous nature. However we have succeeded in converting such gels into crystalline triphosphate-like phases under mild conditions. The ion exchange behavior of these exchangers is under active investigation.

7. Acknowledgement wore author'e Th kth carrien i st ou researcd h laborators wa y supporte Nationae th y db l Science Foundation under grants numbered CHE81-14613 (Chemical Division) and DMR80-25184 (Division of Materials Research r whic)fo h grateful acknowledgemen mades i t .

8. References [1.] RUSSELL, E.Ro, ADAMSON, A.W., SHUBERT BOYD, ,J. , G.E., "Decontaminatio producf no adsorptioy tb tailor-madn no e inorganic adsorbents", U.S.A.E.G. Rep. CN-508 (1943). [2.] KRAUS, K.A., PHILLIPS, H.O., J. Amer. Chem. Soc. 78_ (1956) 644. [3.] AMPHLETT, C.B., Inorgani Exchangersn Io c , Elsevier, Amsterdam (1964). [4.] AHRLAND , ALBERTSSON,S. JOHANSSON, ,Jo NIHLGARD, ,L. , ,Be NILSSON Act, ,L. a Chem. Scand ^ (1964.1 ) 1346, 1357. [5.] CLEARFIELD, A., STÏNES, J.A., J. Inorg. Nucl. Chem. 26_ (1964) 117. [6.] CLEARFIELD BLESSING, ,A. , R.H., STYNES, J.A. Inorg. ,J . Nucl. Chem. 30 (1968) 2249.

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194 FUNDAMENTAL PROPERTIES OF ACID SALTS OF TETRAVALENT METALS AND SOME CONSIDERATIONS ON THE PERSPECTIVE THEIF SO R APPLICATIO NUCLEAN I R TECHNOLOGY

ALBERT. G ! Departmen f Chemistryo t , Perugia University, Perugia, Italy

Abstract Beginning from 1964, acid salt f tetravaleno s t metals started to be obtained as crystalline materials and research was essentially focused for many years in their syntheses and structures w thae chemistrNo .th t f theso y e inorganic ion-exchangers is well enough established, interest has turne o theit d r use n varioui s s technological fields. Among them, their employment in the nuclear fuel cycle as well as in the recovery, separation and purification of some radionuclide s veri s y attractive. After a brief survey in the syntheses and classification e variouth f o s acid salt d theian s r derivative d aftean s a r short review of the ion-exchange properties of zirconium phosphate and its salt forms with a and y lamellar structures some properties suc s hydrolysia h d stabilitan s y towards temperature and ionizing radiations, that are of fundamental importance for their use in nuclear technology^are reported and discussed.

1Introductio. n

Acid salt f tetravaleno s t metals have been a know r fo n long time only as amorphous materials Qi] . Because of their high ion-exchange capacitr theifo d r an ygoo d stabilito t y acid and oxidizing solutions as well as to temperature and ionizing radiations, the potential uses of these materials were subject to intense investigations in the decade 19-54 - 1964, especially in nuclear centers. The success was however very limited, essentially because of the great tendency of

195 the amorphous acid salts to hydrolyze in alkaline or even in neutral warm solutions. Beginning from 1964, acid salt f tetravaleno s t metals started e obtaineb o t s crystallina d e material d researcan s s wa h essentially focuse r manfo d y year n theii s r synthese• d an s structures, as well as in the determination of their fundamental properties. To-day, after two decady of research carried out on these materials in various laboratories, a large number of crystalline acid salts with lamellar, fibrou d threan s e dimensional structures has been obtained and their fundamental properties determined. A vast literature has accumulated and it is currently well known that acid salts of tetravalent metals are members of a very wide class of inorganic compounds ["2-5J . w thae chemistrNo th t f theso y e material s weli s l enough established, interes s agaiha t n turne o theit d r usen i s various technological fields. Among them, their employment e nucleaith n r fuel e cyclerecoveryth n i , , separatiod an n purification of some radionuclides as well as in the treatment of effluents from nuclear powder plants, is still very attractive. Furthermor e crystallinth e e acid salts, contrary e amorphouth o t s ones, have definite chemical compositions and structures. The study of their ion-exchange properties is therefore considerably facilitated by a knowledge of their crystalline structure, phase transitions during ion-exchange processes, thermal behaviour and so on. Useful informatio e obtaineb n d ca even an d n some predictions on their potential applications in nuclear technology can be made,, if the presently known fundamental properties of these materials are taken into careful consideration. For this reason, it is useful to divide the crystalline acid salts into various subclasses according to their compositio d crystallinan n e structur o givt a d briee an e f account

196 on the known fundamental properties of each subclass with particular attentio o thost n e usefua bette r fo lr understanding f ion-exchango e behaviour.

2. Classification of acid salts of tetravalent metals.

Acid salt f tetravaleno s t e consideremetalb n ca s s a d being constitute y macroanionb d s carrying fixed negative charges whic e neutralizear h y protonb d r otheo s r counterions. A classification can therefore be made on the basis of the composition and structure of their macroanions. o dateT o differentw , t kind f macroaniono s e knownar s :

and .IV n elemena h whils T i , X teCe , Pb , n S , Ge , Hf , Zr , i T = wher M e e fiftth h f o group (usuall d As)an P y. Table 1 reports such a type of classification. Note that the X0 structure of the M ^ 4)p macroanion may be either a fibrous or a lamellar one. In turn, two different lamellar structures, usuall yy lamella called an a dr structuree ar s know t presena n t

Table 1

Classification of acid salts of tetravalent metals according to their composition and crystalline structur e macroanionth f o e .

Amorphous i2n- linear (or fibrous structure) a-type bidimensional (or lamellar) structure V-type

amorphous at least threedimensional structure 2 different modification

M(IV) * Zr, Ti, Ge, Sn; Ce; Th. X - P; As; Sb.

197 o givT a complete e picture classificatioth f o e e acith d f o n salts of tetravalent metals it must also be remembered that F IV ~1 lamellar compounds havin e genera; th g ) l O formulaRX ( M e r iv n r iv l L J n n organia wher, 0 XRXs ( i ) O R ce M M d an (R-OX ) O L_ o ^l n l ^ «o o _j n radical, have recently been obtained JJ3,7_]., These compound e knowar ss organia n c derivative f acio s d salts. If the organic R group contains an acid group such as -COOH and -SO^H, the organic derivative may be considered o _ as an organic-inorganic ion-exchanger I 8,9J . Because of their potential uses in nuclear technology, our attention will be essentially focused on lamellar zirconium e phosphatethreedimensionath n o d an s l acid salts with skeleton structure. It seems that the presently known fibrous acid salts are not very stable to hydrolysis and to high doses of r -i ionizing radiations hol while the organic derivatives are expecte o exhibit d t less resistenc o drastit e c working conditions than pure inorganic exchangers.

. Aci3 d salts with g-layered structure.

Some important acid salts with a-layered structure ar e listed in Table 2. Among them, the a-zirconium phosphate is

~~ Tabl2 e Some important acid salt f tetravaleno s t metals with layered structur a-typee th f o e.

Interlayer distance Ion exchange capacity Compound Formula (A) meq/g

Titanium phosphate fTi{P04)_ H2°H 0 - 7.56 7.76

Zirconium phosphate _Zr(P04)2_H2.H20 7.56 6.64

Hafnium phosphate Hf(P04)^H2.H20 7.56 4.17

r Germanium! IV) phosphate 6e(P04)2]H2.H20 7.6 7.08

Tin(IV) phosphate Sn(P04)2"JH2-H20 7.76 6.08

Lead (IV) phosphate Pb(P04)JJH2-H20 7.8 4.79

Titanium arsenate Ti( As04>2"jH2'H20 7.8 5.78

Zirconium arsenate "zr(A304)?]H2-H20 7.78 5.14

H Tin(iv) arsenate Sn(As04)J 2"H20 7.8 4.80

198 7.56 A

7.56 A

• Zr O P O O

Fig Schemati1. . c structur layereO of e -H JzrlP o- H d ij] O (protons and water are not chown in the figure). certainly the fullest investigated and the most stable to hydrolysis. Our attention will therefore be focused on this compound. Its layered structure was first elucidated by Clearfiel A schemati . l Mi d c representatio e arrangementh f o n t of three layers of a- Zr(PO ) planar macroanions is shown in Fig. 1. The distance between adjacent negative charges is o 5.3 A. The layered crystals may be regarded as packed macroanions with intercalated counterions. Consequently e interth , - layer distance will e arrangementdepenth n .o d , volumd an e solvatio e inter-lamellath f o n r counterions. The packing of the layers creates cavities in the interlayer region (on r eacfo e h zirconium atom) where ther s rooi e m enoug r accommodatinfo h e wateon g r molecul r largo e e divalent

Or cation. sa B suc s a h Such cavitie e interconnectear s y narrob d w windows whose dimensions depend on the interlayer distance.

199 —, u D Ir(PO ) JH -H 0 (d = 7.56 A) there is a strong steric hindranc e diffusioth o t e f counteriono n s with ionic diameter o larger than 2.6 A. Thus, at acid pH values, only a small s abli ) e a C , u C , l T numbe, g f A cation o r, a N s, (e.gi L . to exchange the protons of a- Zr(PO ) JH "H 0. Potassium and strontium give slow but appreciable ion-exchange rates while, in the case of large monovalent and divalent cations

such as NH4> Rb , Cs and Ba or highly hydrated divalent or trivalent cacions such as + 3 Mg 2+ , Co etc., the protons in the inter-lamellar region canno e appreciablb t y exchanged even after several days of contact with aqueous solutions of these cations at room temperature. Thus, a- Zr(PO ) H -H 0 seems to

U* .aeJ ba pooe r exchanger. However, zirconium phosphat t sucno h s i e a rigid exchanger as zeolites; as the interlayer distance is increased there is a correspondent increase in the dimensions e windowth f o s connectin e cavitieth g d cationan s f largo s e e exchangedb sizn ca e . Various possibilities are currently known for obtaining considerable enlargemen e originath f o t l interlayer distance. For example, a naked or partially .hydrated Na ion diffuses very easily in the interlayer region replacing one proton for eac hOnc. | cavite3~ , insidfz y e cavitieth e t tendi s o t s réhydrat d watean es forcei r o entet d r too. Since there is no room in the original cavities for water accommodation, the interlayer distance increases until an equilibrium between hydration and expansion of the interlayer distance is reached. At room temperature this process can be summarize s followsa d :

H °H 0 + Na ———»• HNa.SH 0 + HO 2 2 aq. 23

) I (7.56 ) A (11-9 where barred specie se exchange refeth o t r r phase while number n bracketi s s give correspondinth e g interlayer distances .s largOwinit o et g interlayer distancee th ,

200 monosodium form is now able to exchange large cationic s wela s highla lspecie a B y, hydrate ss C suc s a h d divalent or trivalent cations JJ2-14,2,3J. + 2 + For example, the ion-exchange processes with Cs , Ba and s followsa e ar C:r 3+

HNa-5H 20 + Cs aq. ————*• HCs + Naaq. (11.) A 9 (8.) 0A

———>" HNa-5Ha B • 5 0. HBa + 0 ^a N .4 + H 0 q a 2 5 0. . aq 20 r

(11.9 A) (10.0 A)

HNa-5H O + 0.5 Cr3+ ———>• H_ _Cr_ '4H_0 + 0.5H 0+ +Na+ . aq 3 2 5 0. b . O e. 0 ) A (1 9 1. (11.6) A

It was found in our laboratory p5,16j that large cations can easil e exchangeb y originae d 0 th eve H n - i n- Zr(P a H l ) O f smali e presene lexternaar th amount n a i N t f l s-o solution. The sodium ion acts as an ion-exchange catalyst in the sense that there is a considerable increase of the ion-exchange rate while, after the ion-exchange process, all the added Na was found to be present in the aqueous solution. The mechanism of this catalytic behaviour can be understood by taking into account thae diffusioth t ne crystale beginedgeth th f o t sa s s and proceeds towards the inner part. As an example, Fig. 2 2+ shows schematicall a B exchange e presencth yth n i ef addeo e d Initiall. a N y onl a N yione abl o ar exchangst e e protonth e s of the exchanger by enlarging the interlayer distance of the O externa e crystal th par f o o 11.tt l . 8A 2 + This expansion permit a e diffusioB th whicse th hf o nexchange s the Na ions within the enlarged cavities. The exchanged Na ions go towards the inner part replacing other protons and enlarging another fraction of the crystals and so on. At tl e processth enf o de proton ,th whee l exchangear al sn a N d

201 ion must come back to the aqueous solution since the selectivity 2+ + s muci e exchangeh th a highe B f o . r a fo rrN thar fo n There exists another possibility for the exchange of large cationic species in crystalline zirconium phosphate

WïT- H,o î.« ______t

! «jw

I ! f A |5SJ222HZ& ^.V/AV.-V ^ ZZZZZZS&'J>gZBgg> pizaz1___z „„,,,,,vzzzzzzzzzz*, s2* {r ~ j^SKSSSSSSSSSS^ , . . j j jjj. j >. J.JJ

< 8 » Li _îllC..« M H»-tH,0» ÏÏ.JH. * O"' ie.J»,llii 0 ,..!

Fig. 2-Catalytic affect of Na* en the exchange of large cation»

on a-CzrlPOj,] H,-HtO (schematic)

Whee monosodiuth n m for s regeneratei m d inte hydrogeth o n f— -î form with dilute mineral acids, the original a-jZr(PO ) H -H 0 L 4 ^-J ^ ^ is no longer obtained but one has instead a pentahycira-ce form (usually called ^-zirconium phosphate) with an inter- o layer distance of 10.4 A. The process is the following:

HNa-5H 0 + HO ——————*• H °5H 0 + Na . aq 2 2 3 2

(11. ) A 9 (10.) 4A Also, the replacement of the Na ion with some alcohols in acid media (e.g. ethanol acidified with HC1) gives intercalates with a high interlayer distance. When these intercalates are put in contact with water, the alcohol is replaced by water and a- Zr(PO ) H •5H 0 is obtained e zirconiuTh m phosphate pentahydrat a ver s i ye selective exchanger for large cations and deserves further attention. When air-dried four molecules of water are lost with an irre-

versible conversion intoGHZr(P04)J H2 '. ThuH0 s a-|Zr(P. H ) O

202 5H 0 must be kept wet. We are presently studying its thermal stabilit n watei yd preliminaran r y results see o shot m w that a-zirconium phosphate possesse a sgoo d stabilit o 150°Ct p u y. In conclusion, many possibilities are known to-day for obviating the steric hindrance to the diffusion of large cationic species in the interlayer region of a-zirconium phosphate. This exchange n thereforca r e useb e d e exchangfor e th mos. th t f o e par f monovalento t , divalen d trivalenan t t cations. e followingth e problew Th no s e i chemica:m th wha s i d t an l thermal stability of zirconium phosphate for applications at high temperature s higIt ?h thermal s welstabiliti lr ai n i y known £3,19,20] a temperatur t .A e rat f 2°C/mino e e thermath , l behaviours is the following: 22o°- r c P 2 ———— 4

450°-650°Q 800°-1000°CC 1 >

o (6.10 A) (cubic pyrophosphate)

Th e ehydratio th los f o s n wate s irreversibli r d rehydraan e - tion does not occur even when Zr(PO ) H is dipped in cold water e phasTh . e transitio t 220°a n s insteai C d reversible. t The condensation of =P—OH groups to lamellar pyrophosphate depends obviously on the experimental conditions of dehydration [21j (e.g. relative humidity, rat f temperaturo e e increase etc.) and on the degree of crystallinity. For crystalline samples the condensation does not occur until about 450°C s salt-formit whil r fo e s condensation cannoe t th occu d an r thermal stability is even higher than 500°C. In conclusion, a-zirconium phosphate may be considered a very stable exchanger in anhydrous conditions; but what happens at high temperature in the presence of water?

203 Sinc o dat n e eavailabl ar a e literaturth n i e e havw e e carried out some preliminary investigations on the stability of a- |Zr( PO ) l H °H 0 and its monosodium form. L 4 2J z 2 It was found that a- Zr(PO ) H -H 0 is stable in water at least up to 300°C. Only a reversible dehydration at 200°-300°C s foundwa .

Hydrolysis is negligible and it seems to be limited to the externa microcrystale l th par f o t s. e monosodiu th e cas Ith f no e m fore followinth m g behaviour was found : j— -i 5O°— "ioo°c r~ ~~\ a- Zr(P4O ) HNa-5H 0 ———————— >• a- ZrCPO ) HNa-H 0 ^ J £ • « L o H 4 _ L ^ 4J 2 (11.9 A) (7.9 A)

The decomposition into sodium dizirconium -chreephosphate is irreversible; therefore a-zirconium .phosphate cannot be used in an aqueous medium at temperature higher than 200°C if Na ions are present. This temperature is however high enough r manfo y practical uses. Another f lamellainconvenienco e us e r th aci r defo salts i s that thee usuallar y y obtaine s smala d l platelet) M 1 ~ ( s Thus very compact Chromatographie column e usuallar s y obtained The flows are therefore slow while some particles tend to be released inte externath o l solution e problee overb Th . n -ca m come by using larger particles. Incur laboratory it is now possibl o prepart e e crystal f zirconiuo s m phosphate having dimensions of some millimeters No inconvenience Tor the preparation of the columns is enountered with large crystals; however, it must be taken in consideration that the rate of exchange decreases when the size of the crystals is increased.

204 . Aci4 d salts with y-layered structure.

Other than wite tt-structurth h e already discussed, lamellar acid salt f tetravaleno s t metals have been obtained wita h different structure usually known as y-structure [3,23-25]. Until now single crystals large enough for X-ray structural determination have beet obtainedno n o presens , t information on the structure of X-acid salts is very limited. Investigations performed by Yamanaka and Tanaka [23j as well as studies carried out in our laboratory J2a indicate that.the y-structure, similar to a-acid salts, is built up by the packing of planar macroanions M (*0 ) whose negatives charges are neutra- lize y protonb d r otheo s r counterions. Howeve e structurth r e of the y-macroanion is different from that of the a-one. The planar density of fixed charges is higher in y than in a-n?.c poanions (about 1.37 times). Counterions, wate d othean r r polar molecule e arrange b e inter-lamella n th ca sn i d r region.

Owing to its large interlayer distance, steric hindrance e diffusionth o t largf oo e cation s muci s h less than i n a-acid salts n particularI . = 12. d ) ( 2A 0 , H y-[zr(P-2 H )J O |__|_ 4 2J 2 2 seems to be very selective for Cs and its use in nuclear technolog s thereforha y e been predicted [26].

5. Acid salts of tetravalent metals with fibrous structure.

The first fibrous acid a tetravalensal f o t t metal, cerium phosphate s obtaine t laboratorwa , ou n i d n 196i y 8 [27]. Subsequently, fibrous thorium phosphat s alswa e o obtained.[28J and zirconiu d titaniuan m m phosphates have also recently been prepared by hydrothermal methods {~29J . Fibrous inorganic ion exchangers are interesting because they can be used to prepare ion-exchange papers and membranes. A photo- H PO^( e grapsampla C J2 f * o H0 hf showino e fibrous it g s structur s showi e n Figi n . 3 Fibrous acid salts have not been widely investigated and no structures studies are presently available.

205 Fig. 3 - Photograph, taken at optical microscope,of a fragment of Ce(HPC> ) -H O. 42 2

Therma d chemicaan l l stabilit f fibrouo y s cerium phosphate e loweseemb o t rs than tha f a-zirconiuo t m phosphate. Also the resistance to radiation, seems to be lower than that of a-zirconium phosphate po] . However fibrous cerium phosphate exhibits some particular selectivity (i.e. towards lead) that coul e utilizee separatiob d th r fo d f radionuclideo n s £/,30]• Furthermore, fibrous acid salts, contrar e lamellath o t y r ones, e suitablar e material e preparatioth r fo s f Chromatographio n e e preparatioth columnr fo r f o svero n y compact ion-exchange filters.

6 . Acid salts of tetravalent metals with skeleton structures.

X0 M Compounds of the general formula M ( 4^o have been known for a long time [31,32] . They have usually been prepared by solid state reaction t higa s h temperatur 1100°-1400°( e . C)

206 Their structures may be described in terms of XO tetrahedra 0 octahedran M d a whic e linkear h y cornerb d o fort s a m 6 threedimensional network. Although these compounds may be considered as salt forms of acid salts M (XO ) |H, the replacement of M counterions with proton s veri s y difficult. However s recentlwa t i , y found e Li-forth tha f s i prepareti m y decompositiob d f o n a- [zr(Pzr(PO ) 1 Li at 600°-800°C

y solib r o d state synthesi t relativela s w temperaturlo y e j_34J the replacemen i L wit e hth protonf o t n easilca s e carrieb y d ouy singlb t e contact with acid solutions e dizirconiuTh . m threephosphat n hydrogei e n fors beeha m n foun o behavt da s a e + + + + narrow ionic sieve. Only H , Li , Na and Ag can diffuse easily withi e exchangeth n r while large e polymonovalenth l - al d an t valent t exchangedcationno e ar s . The acid salts of tetravalent metals with skeleton structure + + + g A d an n therefor a ca N e recover, th i e useL r b e fo f do y or for the purification of cationic species, even in very concentrated solutions, from traces of Li , Na and Ag .

Som. 7 e concluding remarks.

Table 3 reports some important characteristics of some acid salts of tetravalent metals for use in nuclear technology. A great deal of experimental work is still needed to see if some of these materials are indeed useful fro some practical applications in nuclear technology. Certainly crystalline acid salts exhibit in many cases a complete different behaviour from the corresponding amorphous materials.

207 Tabl3 e

Important characteristic f somo s e acid salt f tetravaleno s te metalus r fo s in nuclear technology.

Lamellar Lamellar Fibrous Threedimensional Characteristics y-lZr(P L 42J2O )JH*22 H 0

Ion-exchange capacity very good Very good Very good Low

Resistance to ionizing radiations M «B ? Bad Very good (estimated)

Resistance to oxidizing solutions .. Very good Very good " < " > .. .. j Resistanc o reducint e g solutions „ " Bad ( .. ( .. j Resistance to acid solutions « 1 « 1 Good Moderate Resistanc o alkalint e e -solutions Bad Bad Very bad Sufficient Thermal stabilitr ai n i y Very good up to Good Good Very good up to 600°C 45O°C Veryo t goo p u d Thermal stability in water 30O°C;bad over T » Very good - mo s it 210» r fo C nosodium form.

Ion-exchange rate Sufficiently good Good Sufficiently Sufficiently good good + 2+ 2+ High for Cs High for Pb , Highly selective for High for Cs* Sr 2+ + 2 and NH£ ions Ba +, Cu Li , Na+ and Agt Selectivity Ba * and trivalent + + . anK d Large monovalent cation traNa f -i s cation d polyvaan s - e presentar s ce . lent cations are not exchanged .

Thus many negative results previously obtained with amorphous acid salts shoul e criticallb d y examinef i e n ordese i d o t r some inconveniences may be overcome by usina the new crystalline acid salt f tetravaleno s t metals y casean n ,I . the investigation e chemistrth n o s f acio y d salte ar s continuing in several laboratories and new materials and new properties are continously being found. Referring only to new acid salts recently obtained in our laboratory, we can cite the preparation of pellicular a-zirconium phosphate, [35] the synthesis of a-zirconium phosphate-phosphite {36,37]

with compositio (Zr(PO.)• n (HFO__ d severaan ) „ „ l H _ . L CC 4 O.66. U j 3oo 1. 66 1 038J . We are now studying the ion-exchange properties of these new materials n particula.I s founwa t di r that zirconium phosphate-phosphite, for its structure (Fig. 4),is able to accommodate in the interlayer region large cationic and polar species.

208 (b)

x \l Photphit« Group

O Phosphlt* Group

Fig. 4. a) Idealized structure of a-zirconium phosphate-pho-

sphit f eo compositio n a-j~Zr(P0>glHPO> 4 6 0

H0.66 ) Phosphatb phosphitd an e e side groupf on o e n o s a layer.

Acknowledgements

The work carried out in the author's laboratory was supporte y Finalizeb d d Projec f F.C.S.Co t f C.N.Ro . d an . by M.P.I. e authoTh . s indebte i rs co-worker hi o t d . U s Costantino and M. Casciola for many discussions and the use of certain results in advance of full publications.

References

] [AMPHLETTi. , C.B., "Inorgani n Exchangers",Io c - Elsevier, Amsterdam (1964). [2.] ALBERTI , G., Ace. Chem. Res., 1J_ (1978) 163.

209 [j3/] CLEARFIELD, A., Chapter 1 of "Inorganic Ion Exchange Materials" (Clearfield, A., Ed.) CRC Press, Boca Raton, Florida (1982). [4.] ALBERTI Réf, G. ,Chap, 3 . . 2 . Qp.] COSTANTINO, U., Réf. 3, Chap. 3. (6.J ALBERTI, G., ALLULLI , S., COSTANTINO, U., TOMASSINI , N., J. Inorg. Nucl. Chem., 40 (1978) 1113. [?.] DINES, M.B. I GIACOMOD , , P.M., Inorg. Chem. 0 (19812 . . )92 [8.] ALBERTI, G., COSTANTINO, U., .LUCIANI GIOVAGNOTTI , M.L., J, Chromatog. 180 (1980) 45. fg.] DINES, M.B. I GIACOMOD , , P.M., Inorg. Chem 0 (19812! . . )92 [10.] ZSINKA , SZIRTESL. , MINK, . , KALMAN ,L . Inorg J. J , , K. ,. Nucl. Chem (19746 3_ . ) 1147. [11.3 CLEARFIELD, A., STYNES, J.A., J. Inorg. Nucl. Chem. 26 (1974) 117. [12.] ALBERTI, G., COSTANTINO, U., GUPTA, J.P., J. Inorg. Nucl. Chem. 3J5 (1974) 2103. [13.] ALBERTI , BERTRAMG. , CASCIOLA, R. I, COSTANTINO, M. , , U. , GUPTA, J.P. . InorgJ , . Nucl. Che 8 (19783 m ) 843. (J4.] ALBERTI, G., BERNASCONI, M.G., COSTANTINO, U., GILL, J.S., J. Chromatog 2 (197713 . ) 477. Qisj ALBERTI , COSTANTINOG. , GUPTA, U. , , J.P. . InorgJ , . Nucl. Chem. 356 (1974) 2109. [16„] ALBERTI , BERTR'AMG. , COSTANTINO, R. I. . InorgJ , U. ,. Nucl. Chem. _38 (1976) 1729. [17.] ALBERTI, G., COSTANTINO, U., GILL, J.S., J. Inorg. Nucl. Chem. 38 (1976) 1733. [18.] COSTANTINO J.C.S, U. , . Dalton (1979) 402. [19.] LA GINESTRA, A., MASSUCCI, M.A., FERRAGINA, C., TOMASSINI, N., Thermal Analysis, Proc. 4th I.C.T.A. Budapest 1974, Vol. I, Heyden, London (1975) 631. C20.] HORSLEY, S.E., NOWELL, D.V., Thermal Analysis, Proc. 3th I.C.T.A., Davos, Vol , Birkhause2 . r Verlag, Basel (1971) 611.

210 .1 COSTANTINO, U., LA GINESTRA, A., Thermochim. Acta 58 (1982) 179. ["22.] ALBERTI , COSTANTINOG. , , GIULIETT U. ,. Inorg J , R. . I, Nucl. Chem 2 (1980£ . ) 1062. [23.U YAMANAKA, S., TANAKA, M., J. Inorg. Nucl. Chem. 41_ (1979) 45. [24/] ALLULLI, S., FERRAGINA, C., LA GINESTRA, A., MASSUCCI, M.A., TOMASSINI . InorgJ , N. ,. Nucl. Chem. _3_9 (1977)1043.

[25.] ALBERTI, G., COSTANTINO, U., LUCIANI GIOVAGNOTTI , M.L., J. Inorg. Nucl. Chem (1979_ 41 . ) 643. [26.] KOMARMENI ROY, S. Natur,, , R. 9 (198229 e ) 707. [27.] ALBERTI, G., COSTANTINO, U. , DI GREGORIO, F., GALLI , P., TORRACCA, E., J. Inorg. Nucl. Chem. 30 (1968) 295. [28.J ALBERTI COSTANTINO, G. , Chromatog. J , U. , (19700 !5 . ) 482. [[29.] ALBERTI, G., COSTANTINO, U., LUCIANI GIOVAGNOTTI, M. L., e publishedb o t . [30J ALBERTI , MASSUCCIG. , , M.A., TORRACCA Chromatog. J , E. , . _30 (1967) 579. J HAGMAN 1 [3 , L.O., KIERKEGAARD Act, P. ,a Chem. Scan2 2_ d. (1968) 1822. [32.] SLJUKIC MATKOVIC, M. , PRODIC, B. , SCAVNICAR, B. , , Croat. Chem. Acta _37 (1965) 115. [33.] ALLULLI MASSUCCI, S. , , M.A., TOMASSINI Italia, ,N. n Paten 48852A/79. no t 6 apri,2 l (1979). [34.] ALBERTI, G., COSTANTINO, U., MASSUCCI, M. A., TOMASSINI, N., Italian Patent no. 49546A/83, 20 december (1983). [35.] ALBERTI COSTANTINO, G. , Italia, U. , n Paten 484387/A. no t , 17 may (1982). [36.] ALBERTI, G., BARTOLI, F., COSTANTINO, U., DI GREGORIO, F., Italian Patent no. 22365/A83, 1 august (1983). C.37.] ALBERTI COSTANTINO, G. , , GIULIETTIU. , Gazz, R. , . Chim. Ital. 113 (1983) 547. [38.] ALBERTI, G., COSTANTINO, U., KORNYEI, J., to be published.

211 RECOVERY OF CESIUM-137 FROM RADIOACTIVE WASTE SOLUTIONS COMPLEW WITNE HA X INORGANI EXCHANGEN CIO R

. ZHAOXIANGS . ZHIGANGT , . ZUNH , , L. ZHENGHAO, W. HAOMIN, L. TAIHUA, L. BOLI Beijing Normal University, Beijing, China

Abstract Recovery of 137Cs from radioactive waste solution has been studied by using titanium phosphate - ammonium molybdophosphate. The complex ion-exchanger shows favorable performanc adsorptioe th r fo ed nan elution of Cs. The product 137Cs has more than 99% radioactive purity and high decontamination factors for 103 Ru, 95 Zr, 22 Na, 90 Sr, Am but less for Rb and Ce. A scheme for recovery process s froC mo fradioactiv 137 e waste solutio s suggesteni d froe th m experimental results with this new complex ion-exchanger.

Introduction

Utilizion of high-level liquid waste from spent fuel reprocessing containing useful isotopes is a matter of great importance for their potential nuclear applications. Two isotopes that have received considerable attention are cesium-137 and strontium-90. Possible technique for recovery of cesium and strontium from radioactive waste solution(IAW) fall into three categories: precipitation, solvent extractio exchangen io d am generan I . precipitatioe lth d solvenan n t extraction processes are rejected because of their adverse effect on waste management due to the added chemical or because they are not effective in the acid range easily attainable in IAW. Furthermore, radiation decompositions easily occur in the extractant. Since inorganic exchangers and adsorbents usually have high chemical sélectivités and high radiation resistance, they have been considere r recoverinfo d g cesium-137 and strontium-90 in

Several exchangers have also been developed for removing cesium from highly acidic solution; these exchangers include the heteropoly acid and acid salts of tungsten and molybdenum, as well as the zirconium, titanium, and tin salts of condensed phosphates. (3-7^ ' As is well known, ' these inorgani n exchangerio c s have some shortcomings r exampl,fo e th e acid salts of tungsten and molybdenum are amorphous. The zirconium and

213 titaniu f condenseo m d phosphates easily exchange blocth p u k e columnf ,i presene trivalentth n i t sfeede statth .n i e This report presents our preliminary laboratory study on a new complex inorganic cation exchanger titanium phosphate-ammonium molybdophosphate (Tip-AMP), determining its cesium capacities and relative sélectivités and capabilities. The results are shown below. w complene e xTh exchanger Tip-AMP possesses unique advantage d shosan o wn shortcomingsc Experimental

Preparation of the ion exchanger

The ion-exchanger titaniun phosphate was prepared according to the method of C. Beaudet ' '. Good granules (50-100 mesh) of Tip were obtaine r laboratoryou n i d , the exchangee nth s loaderwa glasn i d s column or beaker, after being combine ddynamiy b wit P hr statiAM co c methoda ; yellow complex ion-exchanger Tip-AM s obtained Pwa s thoroughl wa t I . y washed with distilled wate removo t r e free acid tileffluene s th l wa H p t neutral, and finally was air-dried for three days.

Determination of breakthrough curves and evaluation of breakthrough capacity

1 gm exchanger was loaded in a glass column (20 cm x 0.5cm) provided with glass wool padding at the bottom, bed volume being 1 ml/go e loadeTh d colum washes nwa d wit 0 columh1 n volume nitrif so c acif o d appropriate molarity e synthetiTh . c feed solution containing known amount of carrier and tracer Cs-137 was passed through the exchanger at a fixed flow rate (2.0-3.0 ml/hr) at room temperature. 1 ml fraction were collected with an automatic collector till C/Co reached approx.l. The C/Co valu s plotteewa d agains e feevolume d breakthrougth th tan d f o e h capacity calculated where C/Co reached 0.0BT)% exchang e 1 1( Th . e capacity is 0.47 meq/g and breakthrough volume is 85 BV« The results are shown in figure 1 and composition of synthetic feed solution in Table I.

Effec f acidite cesiuo t th n mo y capacit f compleo y x exchanger Tip-AMP The exchange capacity was investigated according to the above static method. Table II shows the results.

214 Am-241 7 13 C- s 100 %'

60 o 40 o

123 20 40 60 80 90 100 120 BV Fig. l Breakthrough curves of Am241 and Csl37 from Tip-AMP column

Table I. Composition of synthetic feed solution

Ion Concentration

2.5 C8 0.75 0.075 0.5 Ba 0.25 3.75 2.0 2.0 0.25 l H

Table II. The exchange capacity of various acidity

Acidity HNO- 5 0.5 1.0 2.0 4.0

Exchange capacity 0.50 0.47 0.46 0.45 meq/g

215 Effect of sodium ion concentration on the cesium capacity of complex exchanger Tip-AMP

Both sodiu d cesiuan m m alkale belonth o it g metals. It is greatly meaningful to observe the exchange competition between sodiu d cesiuman m , henc exchange th e e capacit s evaluateywa d with dynamic method at different sodium ion concentration. We found that the sodium ion concentration in feed solution slightly influenced the cesium capacity of Tip-AMP as shown in figure II.

10096 •

Vi o 4» G8 ft

02 1 •a a < u 20JÉ

Tg.(Na/Cs) Fig.2 Effec e adsorf Na/Co th t n o bs percent of Cs-137

Table III. Effect of cycle on the Cesiun capacity of Tip-AMP

Cycle Capacity meq/g

1 0„47 2 0.49 3 0.50 4 0.49 5 0«47 6 0„48 7 Oo45

216 cesiue Effecth n mo f cycl o t capacit e eus Tip-AMf o y P

Exchange capacity for cesium was determined under the above condition on the complex exchanger Tip-AMP. When the cesium breakthrough reached 1% in the effluent, cesium elution was started using 5.5 N NH.NO., + 0.5 N HNO. at 60°C. Then the exchange column was regenerated with 10 column volumes of l N HNOß. Seven cycles could be used at least with this complex exchanger, and results are shown in Table III.

Recovery of cesium-137 and determination of decontamination facto Ce-141f o r , Ce-144, Ru-106, Ru-103, Zr-9 Nb-9d 5an 5

Experiments were als decontaminatioe oth carrien o t ou d n factor of Ce-141, Ce-144, Ru-106, Ru-103, Zr-95 and Nb-95 from feed solution, to whic hfresa h fission produc n indicators addea wa ts a d e feeTh d. solution passed throug exchangee th h fixea t a rd flow rate 2.0-3.0 ml/hr at room temperature. Bed volume was 70. The loaded column was washed with 5 column volumes of 0.5 N HNO, and 15 column volumes of 0.5 N 23° 2°°' 4 S Ce-141, Ce-144, Ru-106, Ru-103, HN0N H removinr 5 C f + Zr-95 and Nb-95 and the purpose of decontamination was achieved. Cesium elutio s carriepassiny nb wa t 0 columou d1 g n volumef o s 5.5 N NH.NO. + 0.5 N HNO, at 60°C. The total cesium eluted was 43 3 examined on their activity at 4096 channel analyser. The cesium effluent concentration indicate a lesd s tha % los n1 n loadin i s colume th g o t n

essentially 100% breakthrough. Washing liquid of 0.5 N HN03 and 0.5 N H2C2°4 + °*5N HN03 were collecte examined theian dr fo dr activity by 4096 channel analyser. The results are shown in Table IV and figures 3,4, and 5. Table IV. Distribution of contaminating nuclldee in the process

Nuclidea ' feed effluent HNO H2C204 HN03 0.5N HH4NO,+ decontamina- cpin cpa 0.5N 0.5H 0.5N HN05 tion factor Ce-141 230990 228557 516 - - _ Ce-144 86110 65600 216 - - ... Ru-103 129738 128336 960 - 1226 106 Zr-95 364273 292488 714 313483 403 2300 Nb-95 569940 107765 1022 1 Na-22 S26670 - 9396 - 710 1164 Sr-85+89 566995 - 2112 a. 670 850 Am-241 1164150 - 4198 - 1536 758 Rb-66 760856 — 32680 — 507274 1.5 .

217 H s i T a, H u ê ITMTv 135 B o 36 - O C-3 Z H H ^ O 1* ? 90 H » * S 24 » g • •H > 0 « •H

K"V £~ 45 0S e 0 ^> O 1 » £ 12 °T p-S H o0)

L.'l lüü 30ü 500 700 800 kev o _ v ke 0 80 0 70 0 50 0 30 0 10 Fig. 4 Gama Spectrometrf o y Fig.3« Gama spectrometr feef yo d effluent

IT» 144 ON

1-1 108 M

72 «i es o •rt •Ö 36

600 800 kev Fig Gam. .§ a Spectrometrf yo washing oxalic acid

Recover f Cs-13o y 7 froW usinIA m g Tip-AMP

highes i s C ry papee puritb Th tha% rf 99 no y chromatograph 137y and los Am-24f so less e 1produci Th s tha. t 1% nCs-13 7 containe, Sr , Ce d Na'an witb R d h concentratio 0.6%f no , 0.1%, 0.28 6.7d an % % respectively. Therefore, this product migh e separatb t e with regar cesiuo t d d an m rubidium. The results are shown in figure 6,7,8.

218 a ÇN Sr-90 Ru-106 Ce-144 u H 1 ^ 4 «a m o o a s H o K -i — 0 t H » • 4» •r« M > "$ 2• - £ 4 . U ^fc *•! A K^s o H •+ •H •r4 1 |C\ •Ö 1 « H o p - Cs-137 o) 1 (Ö * cü IQ o •H "l T) ,** JL°. 1 , J r- « , n . "I v ke 300 090 0 60 8 10 cm Fig. 6. Gama apectrometry of Fig Pape. .7 r chromatographf o y Recovered 137Cg feed

° 6 Cs-137 <»o• n r-l M >, 4 4»

mU o Sr-90- ) P -H Ru.-106 •cOd Ce-144 « i__< U__i 4 6 8 10 cm Fig. 8.Paper chromatographf o y recovered 137Ca

Result Discussiond san s

1. The results reveal that complex ion exchanger Tip-AMP is much better than titanium phosphate (Tip) or ammonium molybdophosphate (AMP) alone. First ion-exchange ,th e column block-up whe uses nr i onlfo dp yTi recovery of Cs-137 from IAW is overcome and the problem of making granulated AMP is solved too. Secondly, the ion-exchange behavior is improved. Although Tip has a high rate of adsorbing Cs, Cs adsorbed is readily eluted; however, it almost loses the ability of adsorbing Cs at high acidity. AMP has a high selectivity capacity for Cs.but Cs adsorbed on it is hard to elute. The new complex exchanger Tip-AMP keeps only the advantages and thus overcomes the demerit.

219 The complex ion-exchanger Tip-AMP has a higher selectivity capacit s reachinC r fo y g 0.47 meq/g when C/C mors i o0.01= e d an , readily eluted; only 10 column volumes can remove above 99% Cs from the loaded column. e producTh s containC t s other radioactive nuclides, total amount being less tha % excep1 n . Rb t The radioactive purity of recovered Cs-137 by paper chromatography and A096 channel analyser is higher than 99%.

The complex ion-exchanger does not adsorb Am-241s so the loss of Am-241 is less than 1% in the course of ion exchange. This is very significant.

2. Tip-AMP exchanger is a complex compound and not a stoichiometric compound. This has been confirmed by X ray diffraction analysis.

3. The results indicate that this new complex inorganic ion exchange gooa s d rha ion-exchange performanc IAWn i .s C r fo e On the basis of these experiments, a scheme for Cs-137 recovering process is suggested. It was shown in Figure 9.

IN HN0 3 0.5N HNO, 0.5N H C 0 0.5N NIi NO + feed solu- 2 2 4 4 tion 0„5N HNO, flow rate 2-3BV/hr flow rate 2-?BV/hr flow rate 2-3BV/hr fluence 5 BV fluence 10 BV fluence 10 BV temperature 15-20°C temperature 15-20°<2 tempera-cure 60 C

1 1 1

i ( t Tip-AMP exchange column !—— —

1 j 1

elute Û.5N UNO Oo5N H2C204 fluent Cs-137 washes waste solution Sr.Ce.Ru Zr-95 Mb-95 No Cs-137

Figure 9« Recovery process of Ca-137 from 1AW

220 References

1. V. Vesley et. al. Talanta 19. 219; 19. 1245 (1972) Clearfield,Ed. A . 2 . Inorgani Exchangn Io c e Materials (1982C )CR Press. Kouri. V n3 .et.al . Atomic Energy Review 12(2) 215(1974) 4. L. Baetsle; D. Vam Deyck; D. Hugs and A. Guery B. L. G. 2 67 (1964) 5. L. Baetsle; D. Hugs B. L. G. 487 (1973) 6. Yvonne M. Jones BNWL - 270 (1964) 7. D.K. Davis; J.A. Partridge; O.K. Koski BNWL - 2063 (1977) 8. C. Beaudet et. al. Nuclear Application 5(2) 64 (1968)

221 TREATMENT OF LIQUID WASTES CONTAINING ACTINIDE FISSIOD SAN N PRODUCTS USING SODIUM TITANATE AS AN ION EXCHANGER

Y. YING, L. MEIQIONG, F. XIANNUA Institute of Atomic Energy, Beijing, China

Abstract e ion-exchangTh e selectivit d capacitan y f sodiuo y m titanate have been studie e preparatio th wels a ds la d characterizationan f thio n s inorganic compound. Sodium titanat a hig s h ha eselectivit r fo y polyvalent ions (Sr, Ce, Fe, Pu, Am, etc.) but not for monovalent ions (Cs) in neutral and basic radioactive waste solutions. Sodium titanate e separatioth e user b fo dn d consolidatioca an n f theso n e actinided an s of fission products int a osuitabl e waste form.

Introduction New inorganic ion-exchangers of high adsorption selectivity have been developed owing to their ideal irradiation, thermal and chemical stability. They might be used in treating and dis- posing of high-level radioactive wastes in nuclear energy field. The techniques of solidification by means of Synroc and monazite high-levee th x tofi l waste high-lever so l o(-waste havl sal e made considerable headway. Inorganic ion-exchangers can be used as adsorbent and additives in solidification of liquid wastes. Natural inorganic materials which have exchange capacity could be use back-fils a d multiplf lo e barrie rdispose systeth r -fo m al of nuclear wastes. Among several new ion-exchange materials of hydrous oxide type, sodium titanate is of most interested. In this work, sodium titanate (ST) was used as an adsorbent for decontaminating and separating the liquid wastes which contains actinide nuclide fissiod san n produc d satisfactortan y result were obtained.

Experiment Preparatio d analysinan f sodiuso m titanate Lab-Scal bees ha nT eS prepare follow:s a d ' 1. Tetra-isopropyl-titanate Titanate tetrachloride was allowed to react with iso-

223 propyl alcoho anhydroun i l s benzene solution metallie th d ,an c sodium was added later to this reaction mixture with stirring. The reaction mixture was heated and refluxed for 3 hours. The product was centrifuged and isolated by vacuum fractionation. Sodiu. 2 m titanate Tetra-isopropyl # titanatNaO10 n i He addes th ewa o t d methanol solution with stirring, then hydrolyse d filteredan d . The product was dried at room temperature. Sodium in ST's product was determined by atomic adsorption spectrophotometry d titaniu,an s determinemwa y cupferrob d n reagent. The components of ST were determined by X-ray diffraction tecnique. Radioactive measurement: e^c 3 and ^-activities were measured by silicon surface semi- conductors plastic scintillation detector Ge(Lid an , ) detector respectively. Result and discussion

- Analysis product The determined valuT producS n ei t agreed witfeee th hd on Na/Ti ratio(in Mole), but in the case of Na/Ti=l/20 in feed the T producresulS r lowes tfo twa shows ra Tabln i n . eI

TABLEI DETERMINATION OP Na/Ti RATIO (IN MOLE) ON ST SAMPLES

NO s Na/Ti RATIO Na/Ti RATIO FEEDN (I ) (ASSAY)

1 1/1 1/1.04 2 1/2 1/1.8 3 1/?0 1/15.2

The sample of ST (Na/Ti)=1/1 which were sintered under different temperature were determine X-ray db y diffraction technique. These data were roughly similar. The sample seemed to be purer when it had been treated at temperature up to 900°C. The diffraction data of sodium titanate are shown in Tablel . These

data are very similar to that of A2B2Xr such as (Ca, Pe, Na)2

(Sb, Ti)2(0,OH)7 type of compound.

224 TABL E1 DAT X-RAF O A Y DIFFRACTIOT S N NO

RADIATION CONDITION Cu KA( Ni ) Al FOIL s EXPOSURHr 5 8. EA m CONDITIO5 1 V K 5 5 N

ST SINTERED ST SINTERED TYPICAL 700° C 900" C A,B,X7 Q D Q D D 11.4 3.900 13.5 3.302 3.285 14.2 3.103 14.8 3.018 3.105 16.9 2.959 16.8 2.667 2.964 21.8 2.076 22.6 2.006 2.007 25.1 1.817 24.6 1.852 1.817 27.1 1.692 28.6 1.610 1.716 28.6 1.642 30.7 1.510 1.550 30.9 1.501 31.7 1.467 1.481 32.3 1.443 32.3 1.443 1.440 34.4 1.365 34.3 1.368 1.337 40.0 1.199 39.9 1.202 1.178

In addition, both Na and Ti were uniformly distributed in each micro regions of whole sample when observed under electron- microscope. - Different way hydrolysis The ion-exchange performance of ST product depends on different condition of hydrolysis. The product hydrolyzed in pure water fome dhara d granule it't sbu s ion-exchange performance goot no d s i enough e produc.Th t hydrolyze water-acetonn i d e solution formed loose granule it't sbu s abilit ion-exchangr yfo e 6 tim5- e s greatewa r tha n purni e water. The ion-exchange performanc differenf eo t Na/Ti ratio(in Kole) was shown in TableJE. Sodium titanate of Na/Ti=l/l(in Mole) seemed better than Na/Ti=l/2. - Hydraulics resistance The liquid flow experimen s carrietwa usint ou d gcoluma n diameter(l.D.n i m heightn C i 1 m C f o .1 1 d Disti y )witbe b T h-S lled water passed through ST column with a flow rate of 6-7 bed volume(B.V.) per hour. After 440 hours(3000 B.V. effluent) opper- ation no resistance change was observed. This result indicated that ST exchanger has a good performance in column operation.

225 TABLK E ION EXCHANGE PERFORMANC WITT S HP E O DIFFERENT Na/Ti RATIO (IN MOLE )

NO: Na/Ti RATIO ST . FEED DISTRIBUTION RATIO ) (M WEIGH ) VOLUM(g T E (Kd)

1 1/2 0.15 10 4.17X102- 6.6x10* a IL 2 1/1 0.15 A.LX U o , 1 10 U J- « Ö I /™o Y i. TJ T \ Qv"\r\J* 3 1.5/1 0.15 10 2 U A97 J. QY T r\J.©Ö* J ÄX( ^ o if 1 n

4 (HTii0,H) 0.15 10 7 15

V, 100 N—• s I—I

I 8° Sr

•^

20-

10' 10 10" 10* TIMS EFFECS TH ? VARIOU 0 T1 PIG. S COÏJTACT TIME ADSORPTIOS O N RATIO » Flow rat ion-exchangf eo e Like other inorganic ion-exchanger, the flow rate of ST exchanger is slow. In batch adsorption experiment, the effect of various contacm A t timd n adsorptioan o e e C n, Sr rati r ofo are shown in Fig.l. The effects of various flow rate on adsorption capacitvolume th d effluenf eo yan break-throug $ 1 t ta e ar h give TableEn ni experimentr ou n .I floe th ,w T columS rat f o en was 0.36Cm/min. (4-5B.V./hr.) Distributio- n coefficient, capacit decontaminatiod an y n factor Batch adsorption experiment were performed by contacting

weighted 0.1 5T sampl S gra NaN0f f o o me l witm fee 0 h1 d solution. 3

226 TABLEE THE EFFECT OF FLOW RAT CAPACITN EO Y

FLOW VOLUM EF EFFLUENO T CAPACITY AT I°/t> BREAKTHROUGH B.V./hr Cm/min ml B.V. meq/g 1.5 0.12 25.5 42.5 2.12 4.5 0.36 24 40 2.01 7.5 0.6 20 33.3 1.93 11 0.88 15 25 1.84 16 1.3 12 20 1.5

TABLI E DISTRIBUTION COEFFICIENT (Kd) FOR ST WITH SEVERAL CATION

IONS FEED COMPOSITION TRACE NUCLIDE Kd Cs + 0.6MNaNOj-0.02M Cs '"Cs 24 z 3 Sr * 0.6KNaN05-0.02M S r *.e?Sr 1.8X10 Ce3* 0.6MNaNOj-0.02M Ce '*'Ce 1.1^10* Fe^ 0.6MNaNOj-0.02M Fe ^Fe 3.4'(10Z Co" 0.6MNaNOj-O.01W Co "GO 1.2*103 Pu''"* 0 OMNaNO. 9. x lcf£//2 - , u P l Pu 8.8X103 Am'* 0.6WNaNOs-1.3xlo"c//m A l '«'Am 1.1X10^ Cm* 0.6MNaNOi-8.6xio"o//l Cm *^m 8.6X103

Each cation to be measured was to feed solution and traced with their radionuclidea respectively acidite th s adjusted ywa ,an o t d resule Th P. tH2 show Tabln ni e1 indicat higa s eh ha tha T S t distribution coefficient for polyvalent cations than for monovalent cation. The equilibrium distribution coefficient, Kd, is difined as: ci W (3) Wher : originaeCo l concentratio liquin ni d phase Ci : remained of solution weigh: T WS f o t volumt V f solutioeo n The column experiment with a flow rate of 4-5 B.V. per hour are given in Table]! . The break-through curves of Sr, Ce, Co and Am on ST column are shown in Fig. 2.

227 o o 80 Co

20 40 60 80 100 VOLUÎu' 7 EFFLÜSir0 î r (B.V.)

BREAtCfHROUGE TH FIG 2 . H CURVEF O S A„ mCo . Ce , Sr

TABLE 21 CAPACITIES AND DECONTAMINATION FACTOR OP ST FOR SEVERAL CATIO COLUMN I N N OPERATION

NOs NUCLIDE FEED CAPACITY D. F. (BEFORE .COMPOSITION (meq/g) \aß> BREAKTHROUGH)

1 Sr 0.6WNaN03-0.02r MS 3.9 10 2 Ce 0.6MNaNOj-0.02M Ce 3o8 10

3 Am 0.6MNaN03-0.02M Nd 4.2 10 4 Co O.öMNaNO 3-0.01*1 Co 1.9 10

5 Pu 0.6MNaN0 .9*10~e//2 3- l D.F.>-10^in 300 B.V. C"Pu) EFFLUENT.

Effec- salf o t t content

The influence of various quantity of Na, Fes Ca in the feed solution on distribution coefficients of Cs, Sr, and Am for ST are shown in Fig. 3« It indicates that Kd of Cs on ST decreases ag the concentration of Na increases. There is no marked change for adsorption of Sr, Ce and Am. While the concentrations of Ca or Fe in the feed solution are more than 2 grams per litre, the

228 Kd

Ara

10'

10* J

10'

Ha:i05 (M)

FIG. 3 THE EFFECT OF Ha* ON ADSORPTION RATIO

iorj

10'

10 10 10 10 (K)

FIG. 4 THE EFFEC-r OF Fe" OÏI DISTRIBUTION RATIO FOr S R

229 JCd

10'

10 10" 10" Ça (M)

FIG. 5 THE EFFECT OF Ca Oil DISTRIBUTION RATIO FOr S R

distribution coefficient of Sr on ST remains greater than 10. and ther significans i e lests i changst i thaf 5 gramei n0. r pe s litre,(See F.g. 4-5) Effec- complexinf to g agent content With the organic complexing agent in feed, most polyvalent cation formed complex with it. In 10ml feed, containing 0.6M

NaNOjand 0.02M Sr, 0.15 grams of ST, and various quantity of NTA, DTPA or EDTA were added. The curves of Fig. 6-8 showed that these three complexing agent all effected adsorption ratio of cations« In simulated high-level liquid wastes(components listed as Table2.), ther littls ewa e adsorptio 0.1f ni 5 T gramS f so addeds wa whet quantite Bu « th n T increas S 3 gramsf o y0. o T et ,S showed certain effec adsorptionn o t concentratioe , th eve f i n n of EDTA was much higher than Sr (See TableM). -= Adsorption performance The adsorption of ST for cation are more suitable in neu- trabasid lan c solution tha acidin ni c solution- ex r . fo Tak r eS ample, when acidic feed solutio passe) 2 H columnT dS (P n ^ 1 ,a break-through would appear in the effluent at 70-80 B.V. However

230 100 s 80

20'

0.001 0.002 0.003 0.004 O.OC5 IirA(M) ABSORPTION O A EFFECE NT TH P O NT6 RATIPIG. O

100 100

-80 g 80 <

60 60

40- 40-

20 20-

0.001 0.002 0.003 O.OC4 0.000 5 0.01 0.02 0.03 DTP) (K A SDTA (M) FIG. 8 THE EFFECT 0? EDTA OH ADSOSPTIOM RATIO EFFECE TH F DTP l O TADSOHPTIC7 Oi A FIG. H RATIO FOR Sr

231 TABLE 31 THE EFFECT OF EDTA(M) ON ST ADSORPTION (IN SIMULATED HIGH LEVEL LIQUID WASTES)

Nos EDTA (M) ST WEIGHT (g) feed volume adsorption (ml) (%} 1 0 0.3 10 95 2 0.0005 0.3 10 94.1 3 0.01 0.3 10 90.4 4 0.05 0.5 10 89.7 5 0.1 0.3 10 47.2 6 0.15 0.3 10 9c7 7 0.2 0.3 10 7.8 a 0 0.3 10 3

TABLE 3[ COMPOSITION OF SIMULATED HIGH LEVEL LIQUID WASTES

COMPOSITION COMPOUND WEIGHT CONCKNTHATION (g/1) (M)

Na NaN03 50 Oo598

Fe Fe(NOi)j9HzO 18 Oo04 Cr Cr(NOj)j . 2 3 0.013

Ni Ni(N03)3 1.2 O.COfi5

Sr Sr(NO,)z 1.5 0.007

Zr ZrO(N03)i2HiO 9 Oo059 Cs Cs(NOj) 0.14 0.007

Ba Ba(N03)z 1.7 0.0065 0.048 Ce Ce(N03)2 6H20 21

then neutral fee7-8H (P passeds ) wa d throughbreak% 8 0. - ,a through appeared after 1180 B.V. of effluent. Desorptio- regeneratiod nan n desorptioe Th d regenerationan T columS f no n have been studied fee.A d solution which containing 0.6M NaNO^-O.Old MN and Am as tracers passed through column until the effluent reached 80^ break-through. Nitric acid (0.2M) was used for desorption. After 10 B.V. acid passed through ST column, the

232 TABLE K E COMPARISIOTH P ADSORPTIONO N PERFORMANC BEFORT S V D E0 EAN AFTER IRRADIATION

DOSE DISTRIBUTION. RATIO( IN BATCH) EXCHANGE CAPACITY (rad) ÜCd) (IN COLUMN) Na/Ti=l/l Na/Ti=l/? (meq/g)

0 4.3X102 6.5*10 2.78 5*107 5.6x10* 7.0x10 2.73 8 5*lo 8.4x10* 7.1X10. 2.75

desorption rati highes owa r than 70# .columT S The e s nth nwa

regeneratio NaN0M 3 y nb 3(addin g H 10)NaQP o .t H Afte cyclo tw r e of adsorption-desorption, the adsorption performance of ST col- t shoy changeno an w d umdi n . Irradiatio- n resistance Irradiation resistance tests for ST have been accomplished, and the result are presented in Table IZ . It shows that both the exchange capacity and distribution ratio of ST did'nt change after up to 5*108rad of irradiation.

Experiment of treatment of actual low-level liquid wastes Feed solution

The low-level liquid wastes were taken frowaste th m e stor- tane Institutn ag i k Atomif eo c Energy specifie .Th whicf o c h was l-2*lo"^(/l, calcute gross a d s-activityß . Analytical data showed thanuclidee th t s they contained werd an e s mainlC , ySr Co. For the ease of measurement, Sr was added to liquid wastes and the PH value of the liquid wastes was adjected to about 7. Test facility

The colum s 5-6"mnwa internan i m l diamete d 11.5Cran m high containing two grams of ST adsorbent. The liquid wastes passed throug T columhS peristaltiy nb c pump wit floha w rat f 3.5eo - B.V./hr5 4. decontaminatioe .Th B.V0 18 n.n i facto efflr S f -ro uent was greater then 104 and remained no less than 103 after 1180 B.V. passing through. The f-spectra of low-level liquid wastes befor afted ean r decontaminatio showe nar Fign ni . 9-10.

233 10 EHERGY(Nev) FIG. 9 THE V-SPSCTRA CF LOW-LEVEL LIQAID WASTES BEFORE DECOIITAriIiïATICH 0? ST COLUMN

CH O 1C'

I Ü

'Co 10

0 ————————————————————————————————————————————————1 — EJïERCY(Hev) K-SPECTRE TH PIC 0 F LOW-IE/E1 O A. L LIQUID WASTES AFTER DECOIITAMIiJATIU T COLUW 3 F O J J The adsorption of Cs was interfered by some ions presented in liquid wastes to be treated. After the liquid wastes were treated by flocculation, the decontamination factor(DoP'e) of Cs raised. The D.P. of Co in low-level liquid wastes was very poor. However, when the liquid wastes passed through an anionic resin column, more than 95^ of cobalt was removed, checking the data of Ü.F. s appeari t Tabln i i s, o V eC ftha t cobalt exist s anionia s c complex stat n liquii e d wastes«

234 TABLE S DETERMINATIO LOW-LEVEF NO L LIQUID WASTES AFTER THREE- COLUMN (AVERAGE EFFLUENT OF 1180B.V.) PURIFICATION

LOW-LEVEL LIQUID WASTES DNTRATED PRETREATED WITH FLOCCULATION

D.F. OF ST COLUMN Sr 1.8X10? 1.1*10 EFFLUENT C3 2.1X10 1.87X10* Co 1.1 1.1

ZEOLITF D.FO . E Sr 4.0 3.5 Cs COLUMN EFFLUENT 1.53X10 1.2X10 Co 1 1

D.F ANIOF .O N RESIo NC COLUMN EFFLUENT C° 9'7 l'°

furthee Th r decomtaminatio existino C d an anionin gs i C f no c complex state was accomplished by using a natural zeolite column and a macroporous anionic resin column. A three—column purification system——ST column, zeolite colum anionid nan c resin column -— was assembled for experiment. Table X is the analytical results of effluents of low-level liquid wastes passing through three column system s determinewa t .I 409y b d 6 multi channel )f-ray spectrometer. After 3-column purification, gross fl-activity in effluent dropped down to 1X1(5 ci/1, almost reaching the permissible level of direct discharge, 7xlo"'ci/l.

Conclusion woulT S cose cheapee f Th b do t r than other inorganic ion- exchanger and commercial ST would be much cheaper than ST in Lab-scale. After adsorption, ST could be sintered at 1000°C translucent glass-like ceramic form, which reduced volumf eo waste and facilitated waste treatment, In treating actual liquid wastes, the ST-aeolite-anionic resin column system is simpler and easier to operate as compared with existing evaporation method. Moreover, it can be installed as a small mobile device suitable for liquid wastes treating under local condition.

235 REFERENCES ai.t SANDI e e 1 Lynch Th ,. Dosc . G " W A . . hsolidifica,H R - tion process broa—A d range aqueous wastes solidification method" IAEA-SM-207/71 36 5 2 Schwoebel R. L. et al., " An inorganic ion exchange process r partitioninfo g high-level commercial wastes" NR-CONF-001 30? titanatf o e us decontamination i ee 3 Th Dosc. G . hR f no defence wastes solidification SADD-78-0710 (1978) 4 Johnstone J.K. Headly T.J. Hlava P.P. and Stohl F.V. Characterizatio" titanata f no e based cerami higr cfo h level nuclear waste solidification (Scientific basir sfo nuclear wastes managemen (1979^ 1 ) t ) 211-217 5 Schulz W.W. Koenst J.W. and Tallant D.R. "Application of inorganic sorbent Actinian i s e separation process" (Actinide Separation ) (1980) 17. 6 Forber Westermar. gS al.t Arne. e tT . ,R k "Fixatiof no medium level waste titanatn i s zeolites:progresd ean s toward a system for transfer of nuclear reactor activities from spent organic to inorganic ion exchanges" ( Science Basis Nuclear Wastes Management ) 2 (1979) 867.

236 ION-EXCHANGE PROPERTIE ZEOLITEF SO THEID SAN R APPLICATION TO PROCESSING OF HIGH-LEVEL LIQUID WASTE

. KANNOT . MIMURH , A Research Institute of Mineral Dressing and Metallurgy, Tohoku University, Sendai, Japan

Abstract Zeolites which have framework, structures are excellent inorganic ion exchangers having high stabilitie o radioactivt s e irradiation. Thei n exchangio r e propertie d thermaan s l transformation have been studied for the purpose of their application to the separation of cesium and strontium from the high-level liquid waste (HLLW). Mordenitee zeolitesth f o e ,, on selectively exchanged cesium ion with high selectivity coefficient more than 1000, thus cesium ion can e preferentiallb y exchanged intmordenite th o e even in the acid solution around pH 1, and separate from other metal ion HLLWn i e sselec Th . - tivit r cesiufo yn decreaseio me order th n i s, mordenit » ezeolit zeolit> eY zeolit> X e . eA The cesium exchanged mordenite is transformed to cesium aluminosilicate (CsAlSi5O12) by calcination above 1200°C. On the other hand, strontium is selectively exchanged int s selectivito it zeolit d an A es i y reversed to that of cesium ion,zeolite A > zeolite > zeolitX mordenite> eY . Strontium formf so zeolite A, X and Y are recrystallized to triclinic strontium aluminosilicate (SrAl2Si208) by calcination above 1100°C. The leaching rates of cesium and strontium ions from these stable minerals are less than 10~9 g cm"2 day"1, which are lower by the three order f magnitudo s e than those from borosilicate glass products. These excellent properties of zeolites for the separation and fixation of cesium and strontium can be applied to the partitioning - proces so e processin th e th n i s o t f HLL o gd an W lidification. 1. Introduction Zeolites are crystalline aluminosilicates with framework structures and have excellent cation exchange properties. Zeolit soms ha ee cavities occupied by cations such as Na, K, Ca and water molecules. Thes- eex catione b n ca s changed with other cations, and-the selectivities of zeolite for cations seem to be different with their structures, Si/Al atomic ratios and the position of ion exchange sites.

237 Som f zeoliteo e d formeha s d naturalln i y deep geologic formatio y hydrothermab n l reaction, and also many zeolites are synthesized artifi- ciall y varioub y s methods. Som f theseo e synthetic zeolites have been used as molecular sieves and catalyzer petroliun i s m industr d othersan y . Most e naturallth f o y occuring zeolite Japan i se ar n mordenite, clinoptilolit r theio e r mixtures. e atomiIth n c energy industry e treatth , - mend managemenan t f high-leveo t l liquid waste (HLLW) are the most serious problem. In fission products, the most soluble and long lived radioactive isotopes are 137Cs and 90Sr, therefore the fixatio f theso n e isotope o insolublt s - ma e teriale mosth t s i valuabls o securitt e y against the management of HLLW. e havW e studiecatioe th n no d exchange properties of zeolites and their application to the fixation of 137Cs and 90Sr, and some interesting results were obtained. 2c Zeolites Synthetic X usezeolit n thid i d an s eA stud y are obtained from Toyo Soda Co. Ltd«, and synthetic zeolite Y and synthetic mordenite (Zeolon 900Na, abbreviated as SM) are made by Norton Co. Ltd. Natural mordenite (from Sendai area, Miyagi prefecture, Japan, abbreviated clinoptiloan ) NM s a d- lites (from Futatsui, Akita prefectur d froan e m Itaya, Yamagata prefecture, Japan, abbreviated ) werasCP e also used. n exchangIo . 3 e propertie f zeoliteso r varioufo s s cations Effects of pH on the distribution of Cs, Sr and some of transition elements such as Cr, Fe, Co, Ce, Eu and Tb into Na- and H-zeolites from simple cation solutions have been investigated. Distribution curves of these cations for NaA and NaX are shown in FIG. 1, and these distribution curves of cations except for cesium have e distributiomaximTh . 7 a - aroun 6 n H ratiop d s f theso e cation maximt a s aboue ar a t 10 1- 10* s. e distributioTh e alsar o nY curveNa r fo s X zeolitesNa d simila- an De A „o thosNa t r f o e crease f distributioso d higan nh w ratiolo t a s pH ranges seem to be mainly attributable to competitive exchange with hydrogen ion and hydrolysis of cations respectively. On the other hand, distribution rati f cesiuo n decreaseio m s monotonously with decreasin. pH g In the cases of Na- and H-forms of mordenites, selectivels cesiui n io m y exchanged into these zeolite H rangp s w showa e lo eveFIGn i nt a n , .2 therefor e separateb ey cesiuma n d io m from other multivalent cations under this condition.

238 NaX Tb '"V\Eu ra - NaA Cs Tb

10* 10' Sr

103 Fe

10'

,P . 10 12345678910 10 l 2 34 56789 10 pHeq. FIG. l Cation exchange properties of various cations X Na d an intA oNa

10'

10 12345678910 1 23450 6

FIG. 2 Cation exchange properties of various cations into NaSM and~HSM

239 4. Selectivity of zeolite for cesium ion Zeolite has an ability to exchange various cations s selectivitit t ,bu s influencei y e th y b d structure and the position of ion exchange sites. In the case of exchange of cesium with a large ionic radius, sodium ions located in large cages or large channels can be easily exchanged with cesium sodiut bu , m ions locate narron i d w channels or frameworks are hardly exchanged with cesium by ion sieve properties of zeolite. n exchangIo e r exchangisothermfo A Na ef o s reaction with casiu e showe ar mFIGn Th i n . .3 sigmoid curve seems to show the presence of two kind f sito s e exchangeable with cesium. Plotts of corrected selectivity coefficients against to the equivalent fractions of Cs in NaA for Na - Cs exchange are shown in FIG. 4. In this figure, upper line shows the corrected selectivity.coefficients at the sites in large a-cages (site whicn i , h 1) cesiu mors i m e selective than sodium.

1.0

0.8

0.6

0.«

0.2

0 1. 8 0. 6 0. 4 0. 2 0.

FIG- 3 Ion exchange isotherm of NaA for Cs exchange

1.5

1.0

0.5

-1.5

Site 2 -2.0

-2.5 0 02 0.4 0.6 0.8 1.0 Zcs FIG. 4 Ion exchange selectivity of Na-Cs exchange

240 Lower line shows the corrected selectivity coefficients at the sites in the plane of 6-membered ring (site 2), in which sodium is more selective than cesium. In NaA, 4 sodium atoms in large a-cage of unit cell are easily exchanged with cesium, while other 8 sodium atoms in the planes of 6-membered ring are hardly exchanged with cesium. From these results, the cations in a-cage and in the plan f 6-membereeo consideree db rinn ca g s a d hydrated cations and bared cations, respectively. Ion exchange isother f NaSmo r exchang fo M e with cesiu e show ar mFIGn i n . 5 .Thi s simple curve seem o shot s w thae sitth te exchangeable with cesiu onls i m y one exchangn .Io e selec.- tivity for Na - Cs exchange in NaSM are shown a simpli nd FIGan e , straigh6 . - t ob lins wa e tained.

0 1. 8 0 6 0 4 0. 0.2

FIG. 5 Ion exchange isotherm of NaSM for Cs exchange

3.0

2.5 -

20 SM

1.0

0.5

I______I______I______I 0 1. 8 0 6 0. 4 0. 2 0. 0 n exchangIo FIG6 .e selectivit Na-Cf o y s exchange

241 From this result s knowi t i n, thae th t selectivity coefficien r f NaScesiuo tfo - M ex m change is more than 1,000 at initial exchange with cesium which is very large than that of other zeolites. The order of the selectivities for cesium exchange is NaSM » NaY > Nax > NaA. 5. Thermal transformatio- Sr d an f Na- o n- Cs , zeolites s welIi t l known that zeolite e aluminoar s - silicates having three dimentional structures, and include the various cations in their framework structures. After zeolite structure ar e contracte y heatingb d , cations containen i d zeolites may become to be insoluble to water, since these cation e locatee aluminoar sth n i d - silicate frameworks of zeolites, zeolites contained cesium and strontium may be also transforme o cesiut d d strontiuan m m aluminosilicates y heatinb t higa g h temperature. Thermal transformations of Na-, Cs- and Sr- zeolites are investigated by means of differ- ential thermal analysis (DTA), thermogravimetric analysis (TGA) and X-ray powder diffraction, in order to know the possibility of fixation of 137Cs and 90Sr. Phase transformation of Na-, Cs- and Sr- forms of zeolites at high temperature are summarized in TABLE I. In the case of Na-A zeolite, the mixtur f carnegieito e d nephelinan e s wa e obtaine t 900°Ca d d the,an n change a simpl o t d e nepheline phase. Cs-A zeolite is transformed to the mixture of nepheline and pollucite, one of the cesium aluminosilicate, above 1200°C. Strontium A zeolitfor f o m s transformei e o t d hexagonal strontium aluminosilicat t 1000°a e C and then change o triclinit d c strontium aluminosilicate above 1200°C. Sodium form of X zeolite changed to amorphous phase at 850°C, and then recrystallized to nepheline above 1000°C. Cesium X zeolitfor f o m s finalli e y transforme o pollucitt d e above 1000°C, througe th h amorphou sn thi i phas sd casan e o nephelinn e e phase was detected. The thermal transformation f Sr-o X zeolite s similai s o that rf Sr- o t A zeolite. Sodium form of Y zeolite was completely destroyed at 800°C and does not formed any crystalline phase. On the other hand, CsY formed pollucite phase above 1100°CY Sr d ,an changed to strontium aluminosilicate (SrAl2Si208) at 1200°C. Some different results between X- and Y-zeolites may be due to the difference in Si/Al atomic ratios. Mordenit moss i e t stable zeolite, hence there is no change in its structure until about 900°C. Cesium exchanged synthetic mordenite is transformed to a cesium aluminosilicate

242 TABL I ETransformatio zeolitef no higt a s h temperature

Cation Na Cs Sr Zeolite

900° >1000° 900° >1000° 800° 900-1000° >110Q° A Cam. Neph. + •*• Neph. Cam. -»+ • Am. -*• Hex. -»• Trie. Neph. Pol. 850° >1000 800-900° >4000° 800-900° 1000° >1100° X Am. -»• NephPol• -» . . Am Am. -*• Hex. -*• Trie.. >600= 1000° >1100= 900= 1000s 1100* 1200s Y Am. Am. -*• Pol. Am' . -*•?-»• Am.-»- Trie. >900° 1100° 1200° 1000-1200° >1300° SM Am. Am. -»• R. Tric.+C.+Q. C - -* . >1000° >1000° >1000° NM Am. Am. Am. >1000° >1000° • >100.0° CP Am. Am. Am. Am.: Amorphous, Carn.: Carnegieite(NaAlSiO

Neph.: Nepheline(NaAlSi04), Pol.: PolluciteCCsAlSi Og), Hex.: Hexagonal(SrAl„Si 0 ), Trie.: Triclinic(SrAl,Si 0 ) o / / o / Z n 0 0 Q R.: CsAlSi 0 , C.rCristobalite, Q.: a-Qualtz

(CsAlSis012) at 1200°C, and strontium exchanged synthetic mordenite changemixture th o f t so e triclinic strontium aluminosilicate (SrAl2Si208), cristobalit d ot-qualtan e t 100a z 0 - 1200°C d ,an then changes finall cristobalito t y e over 1300°C. othee th rn O hand d Sr-form, an Na- - f ,Cs so natural mordenit d clinoptilolitan e e chango t e amorphous phase over 1000°Co crystallinn d ,an e phas detecteds i e . 6. reachabilities of cesium and strontium from calcined zeolite The leachability of 137Cs and other radioactive nuclides from the solidified product of HLLW is one of the most important parameters in the safety assessmen managemene th r fo t f HLLWo t , because the leaching of radioactive nuclide from soli de firswastth e s nuclidti eth ste f o pe migration with groun de repository wateth n i r . In order to realize the ceramic solidification with lower leachability, the authors have investigated the solidification procedure by calcination of zeolites exchanged with cesium and

243 strontium as mentioned above. Therefore/ the leachability of cesium and strontium from these calcined zeolite s investigatedwa s . 6.1 Effect of calcining temperature on the leachability of cesium Alkaline ions such as cesium are the most soluble element HLLWe th n ,i stherefore e leachth , - ability of 137Cs is often measured as one of the parameters for security assessments. FIGURE 7 shows the variation of the leachability of cesium from calcined product into pure water with calcining temperature. Leachability decreased with increasing in calcining temperature, and tha X decreaset Cs fro calcinee d th man A d Cs d rapidly between 900° and 1000°C by the crystallization of pollucite (CsAlSiOit ) , and nepheline (NaAlSiOit) while the amount of leached cesium frocalcinee th m d natural zeolites graduall- de y creased with increasing in calcining temperature. e leachabilitTh y froe calcineth m d natural zeolites was a little higher than that from calcied A and X zeolites, since the mordenite and clinoptilolite are more stable at high temperature than A and X zeolites e leachabilitTh . f strontiuo y cons i m - siderably lower than tha f cesiumo t .

800

o CsA • CsX N 600 • CsNM D CsCP(Itaya) • CsCP(Futatsui) U 0 400

o £

900 1000 11001200 Temp. CC) FIG 7 .Variatio f leachabilitno s intC of o ydistille d water with calcining temperature (Calcining time) h 3 ;

2 Leachin6. g rate The manners of representation for leaching results were contrived accordin objece th o t g and method of leaching test. The leaching rate presented by ORNL is available and expressed as follows[1].

244 Leaching rat= e (g cm~2 day"1)

where X: Amount of a constituent present in the sample/ g : QuantitY y leached frosample th m n ei unit time, g/day A: Surface area of the sample, cm2 samplee Weigh: W th f to ,g The evaluation of surface area necessary for the calculation of leaching rate were made by means of BET method using nitrogen gas. The measured values of leaching rate are shown in TABLE II. The leaching rates into pure water were about 10~ 810~- cesiur 9 fo 10~d an m 910~- r 10fo strontium, respectively e valuTh .f leachin eo g rate of cesium into distilled water was very small, and that from the calcined y zeolite was smallest e synthetiith n c zeolite e formatioth y b s f o n pollucite. The amount of cesium leached from cesium zeolites calcined at 1200°C and that of strontium leached from calcinezeoliteX d an dA s were under the detection limit of atomic absorption analysis.

TABLE II Leaching rate of Cs and Sr for calcined zeolites , -I , 2 - . ) y da m (c g

Sr Zeolite Cs Distilled Sea Distilled Sea water water water water A 2.5xlO~8 1.2xlO"7 <2.6xlO~10 1.8xlO~S X 1.7xlO~8 6.0xlO~8 <2.6xlO-10 l.SxlO"8 7 Y 1.3xlO~8 i.ixicT 2.8xlO~9 9.7xlO~9 SM 7.4xlO~8 8.5xlO~8 5.0xlO~9 1.6xlO~8 9 8 NM <2.3xlO~ l.lxlO" _ _ CP <3.8xlO~9 1.3xlO~8 (Itaya) CP 9 8 _ _ (Futatsui) <3.2xlO~ 1.5xlO~ werM S e d calcinean ZeolitY . , t h a X d 1100°3 , A er Cfo Zeolite CP and NM were calcined at 1200°C for 3 h.

. 7 Volatilit f cesiuo y m from zeoliten i s calcination Volatilization of Cs-zeolites at high temperature is investigated for the purpose of the safety assessment. As shown in FIG. 8, volatilization curves of cesium are very different with zeolitee amounth d f volatilizeo tan s d cesium

245 0.24

O 110KO 0 0 12090 00 13060 0 70 Temp. ("Ci FIG 8 .Variatio volatilizatiof no n rate with calcining temperature

is increased with increasing temperature. It is show FIGn i n . 9 tha t cesiu volatilizes i m d only e firsath t t stag f heatino e t 1200°Ca g d ,an X zeolitexcepCs r fo et volatilizatio f césiuno m is depressed after abou 0 min-heating4 t t seemI . s that this depressio e destractio owins th i n o t g n f zeolito e structure t higsa h temperature. Vola- tilized specie f cesiuso s determinemwa Css a d 20 by mass spectrometry. Therefore/volatilitf o y cesium is decreased in inert gas flow such as argon e volatilizeTh . d amount calcination i s f o n Cs-zeolites X werexcepCs er verfo t y little (Table III).

0.2*

0.20 •

01« - 8. 0.1Î -

n 0.08

0.04 <

0 16 0 12 0 8 240 Tim«(min.)

FIG. 9 Variation of volatilization rate with calcining time

246 TABLE irr Volatilization of Cs from zeolites (Calcining condition: 1200°C) h ,3

Zeolite Air Ar CsA 4.1xlO~2 2.9x10* 1 2 CsX 2.0x10 8.8x10 CsY l.OxlO'1 8.4xlO~2 CsSM 7.3xlO~2 l.lxlO~2 CsNM 3.4xlO~2 4.4xlO~3 CsCP 2.4xlO~2 5.2xlO~3 Volatilized Cs_____ x 100 Volatilization rate '(%= ) Totan zeoliti s lC e

8. Conclusion Zeolite e verar sy excellent materialr fo s adsorption and fixation of radioactive nuclides in HLL mentiones a W d above. Mordenite is most selective to cesium ion e zeolitesith n d cesiu,an m exchanged synthetic mordenite is transformed to cesium aluminosilicate (CsAlSisOia t higa ) h temperature, whic s veri h y stable and had difficult leachability. Therefore, mordenite is very suitable for solidification of 137Cs. Zeolite A exchanged with Sr is transformed to strontium aluminosilicate (SrAl2Si20y 8)b calcination. Since zeolites have very excellent properties e treatmenfoth r f radioactivto e nucliden i s high-level liquid waste s necessari t i , o promatt y e the investigation on the application of zeolites e nucleafoth r r fuel cycl n atomii e c energ- de y velopment. REFERENCE [1] IAEA Technical Report Series, No.82 (1968) 101.

247 STUDIES ON INORGANIC ION EXCHANGERS- FUNDAMENTAL PROPERTIE APPLICATIOD SAN N NI NUCLEAR ENERGY PROGRAMMES

R.M. IYER, A.R. GUPTA, B. VENKATARAMANI Chemistry Division, Bhabha Atomic Research Centre, Bombay, India

Abstract Investigation n o inorganis n exchangerio c s have been carried out with a view to understand the fundamental properties and to assess their suitability in specific applications in the Nuclear Fuel Cycle. Method of preparation had a marked influence on the structure and properties of this class of materials, as in the case of crystalline thorium arsenate (TA) and hydrous oxides, especially hydrous titanium oxide (HTiO). The new crystalline modifications of TA, in contrast to the monohydrate, exchanged larger n additioi ion * - ,K sLi , liko t * n Na e due to the large interlayer spacings. Hysteresis in H*-Li* exchange was observed. The selectivity sequence r alkalifo s , alkaline eartd transitioan h n metal ions on hydrous thorium oxide (HThO) and magnetite (MAG) were simila o t thosr e observer fo d other hydrous oxides. While hysteresi n i anios n exchange involving uni-univalent anion n o s HThswa O observed, cation exchange reactions d involvinan a C g C* weru e e non-reversiblefounb o t d . Preliminary investigations on the usefulness of hydrous e oxiderecoverth n i sf o uraniuy m a frose m water, and purification of reactor coolant water are reported in this paper. Different hydrous oxides having varying acidities were screened for their uranium sorption capacities n O HTiOe th mos., t promising adsorbent amone hydrouth g s oxidese th , influence of anions on the sorption of uranium. the possible mechanism of sorption of uranium from carborate solution d undean s r simulate a watese d r conditions were studied. Investigatione th n o s sorption of cationic corrosion products on hydrous zirconium oxide (HZrO), HThO and MAG upto 363 K and of anion n o mixes d oxides. like Bi(111)-Th(IV) mixed oxides, from alkaline solutions were carrie t witou da h vie o test w t their utilit n i reactoy r coolant water purification systems.

Adsorbents when used in Nuclear Energy applications are liable to see radiation fields, which can affect and/or influence their sorption properties. With this view, in-vivo •y-radiation e th effec n o t sorption properties of some representative hydrous oxide s beeha s n e resultth studie d an s d wile b l discussed.

249 Results of our recent studies on the properties and mechanism of reactions of silver exchanged molecular sieve for the removal of CH Ï are also discussed.

1.0. INTRODUCTION Investigation n o inorganis n exchangerio c s like thorium arsenate (TA) [1.23. a variety of hydrous oxides and mixed oxides [3-14], ferrocyanide molybdate, ferrocyanide tungstate CIS d an sorbent] s like molecular sieves have been carried out in Chemical Group. 8ARC. wit a viewh : 1. to understand the fundamental aspects of ion exchange and/or sorption properties of these materialsd an ,

e whethese o 2t .r these materials could find possible application in Nuclear Energy Programmes, like purification of reactor coolant water, recovery of uranium from sea water, treatment of radioactive wastes etc., Salient features of the studies will be summarised in this paper. Results of preliminary experiments on the effect of irradiation on inorganic ion-exchangers are also included.

2.0. BASIC STUDIES ON INORGANIC ION EXCHANGERS 2.1. Influenc f o Methoe f o Preparatiod e th n o n Properties of Inorganic Ion Exchangers In general e stabilityth . , n structurio d an e exchange and/or sorption properties of inorganic ion exchangers have been foun e conditiono depent dth n o d s f preparatioo n r experiencCIS]Ou . s indicateha e d that s i applicabl t e i cas th f thoriuo en i e m arsenate (TA) and hydrous oxides also. In the case of TA, slight variation in the preparation conditions w resultecrystallinne n i d e forms of TA viz Th(HAs04> 2.5 H20 and ThfHAsO^) . 4H 0. [1,2], compared to the one reported in tne lilerature, viz, Th(HAsO^) .H 0. [17]. The interlayer distance of the new crystalline forms were larger namely, for TA.2.5H 0 it was 8.59 A and TA.4H 0 it was 8.23 A - as compared to the material reported in literature, TA.HgO (7.05A) [173. The new crystalline forms were capable of exchanging larger ions like Na , n additioi K* o Li*t n , while TA.HO could tak p onlu e y 2 LJL* [17]. Anion and cation sorption properties of hydrous oxides were found to depend on the nature of the alkali used for precipitation [3.4,7,8,10.11]. Thus, hydrous oxides prepared using NaOH had better cation sorption and the ones prepared using NH OH had better anion

250 100 -

oc.ÖD o

ce. 15 I- Z uU tt Lü (L

0 -

0-4 0 -3.0 -20 -10

Fig. t Effect of method of preparation and carbonat e sorptioe th ion n f o so Uraniun n o m hydrous titanium oxide (HTiO) [13] O.HTiOJD) • .HTiO(H}; ;.HTiO(D-TA )A,HTiO

sorption properties. In the specific case of hydrous titanium oxide [13], it was established that when NaOH was used for precipitation, excess Na* was incorporated in the oxide matrix and when urea was used, the s incompletehydrolysiwa a i consequenceT s f A o s. , the two preparations had different sorption characteristics (Fig.1). However, when excess Na* was removed by dilute acid treatment and complete hydrolysis ensured by alkali treatment. the sorption chareacteristics of the two materials were very similar. Our experience. thus. indicates that for better understandin e propertieth f o gf o inorganis n io c exchangers, conditions of preparations should be carefully chosen and impurities present in the materials shoul e b detected d removean d y suitablb d e treatment.

2.2. Propertie f Crystallino s e Thorium Arsenate [1.2] As mentioned earlier e thoriuth . m arsenate (TA) prepared ane dth composition studied ha d , Th(HAsO .2.5H.O, ,) , and TMHAsO ) .*H 0. These materiaals wereliss stable chemically and thermally than

the Th(HAsO.)..H_.•0* reporte•'•vp*v*«.9Vdi i^.nn literaturJ.At*V.L«llUre [17]. Although both*tne materials exchanged witd h an Li* " ,Na K*. TA.2.5H 0 was found to be chemically and thermally more stable than TA.4H.O e exchangerTh . s underwent

251 4.0

2 1 0 1

Fig.2 Alkali metal ion/H exchange as a n o TA.2.5functio n i H aqueoup d 0 f H o nan s methanolic media [13 o,metal ion uptake in aqueous medium;•,metal ion uptake in methanolic medium ;Q .arsenate release n aqueoui d s medium;«.arsenate released n methanolii c medium;A,Li uptak n i aqueoue s medium (reverse exchange>;ASarsenate released in aqueous medium (reverse exchange)

extensive hydrolysis during exchange e maximuTh .* Li m ion uptake and the corresponding arsenate released, respectively, are: 1.75 and 1.20 mmol/g for TA„4H 0 at pH 9.0 and 3.42 and 0.28 mmol/g for TA.S.SH^O ai pH It.8. In methanolic medium, it was possible to incorporate significant amounts of Na and K , ions A T 2.5.in , 0 Hwithou t extensive releas f arsenato e e ions. Hysteresi H - exchang i L s n i sobservewa e r fo d TA.2.5H 0 (Fig„2). The metal ion exchanged materials had different crystal structure as compared to H*-form. The number of water molecules associated with the metal ion exchanged form followed the Same sequence as observed in the various cationic form f synthetio c organiclion exchangers [18]: Li*>Na*> K* . Th t e freeno wate s ,wa r as indicated by the absence of HO bending R modeI n i s spectra, and presumably existed partially hydrolysed structures.

252 2.3. Propertie f Hydrouo s s oxides As hydrous oxides behav s catioa e n exchangern i s alkaline medium and as anion exchangers in acid medium the catio d anioan n n exchange behaviou f somo r e hydrous oxides have been investigated.

2.3.1. Alkali metal ions Sorptio f o alkaln i metal ions from alkaline solutions either decreases with increasing ionic radii e cationth f o s (for example, Li* > NH^ * >n o *Cs Na*2 > * .K hydrous thorium oxide [3])d an wit. h increasing hydrate e d cationth radi f o is (for example,C> K - s Na*> Li* n magnetito , e [5])(Tabl . ThiI) e s indicates that on hydrous thorium oxide, cations are partially or fully dehydrate n o sorptiond , whil n o emagnetit e hydration of the exchanging ions is more or less unaffected.

TABL : SorptioI E f alkalo n i metal ion n hydrouo s s thorium oxide (HThOd an ) magnetite(MAG) from 0.1M MeOH solutions(Me: Li*.Na* Cs*, H *)[3.5K* .N , ]

Meta n io l Crystal ionic Amount sorbed (meq/g) radii G MA HThO

Li* 0.68 1.15 0.54

Na* 0.97 1.03 0.68*

K* 1.33 1.04 0.78

Cs* 1.67 0.96 0.71

^ L 1 * U nU n .L • *r j 0.46 * From 0.13M solution.

2.3.2. Alkaline earth metal ions Sorptio f o alkalinn e earth metala ions i s complicated process because contributioth f e o e th o t n sorption from hydrolysis of the cations. As a result of this, sorption show an abrupt increase over a narrow pH range [12]. 2.3.3. Transition metal ions Sorptio f o transition n metal ions from neutral solutions on hydrous thorium and zirconium oxides, magnetite, La(111)-Fe(111) mixed oxides, followed the sequence: [3,4,5,9,1V], Cu 2> Zn* 2> Ni* 2> Co* 2> Mn* 2* The sequence parallel e Irwing-Williath s m series observed in the case of stability of metal complexes.

253 H"h»IO~3N 1.0 1.0 0.8 «^ 0.8 e « o. ^ 4 • U 0.6 Z 0.4 5 0.2 0.2 0.0 0.0 0.0 0.4 0.8 1.0 0.0 0.2 0.4 0.6 0.8 1.0

(Cl) C/CC e/eee(c(Co< ) (0) (b)

Fig,3l»} NO --Cl Exchange isothermson HThO[7]_ __ _ A. N03"-» Cl"; A. Cl"-» N03" ) (b NH.*--C * aExchang e isothermn o s HThOm C—a 2 + ;*.C— 2a + -» N—H

This indicates that surface hydroxyl e groupth f o s hydrous oxides coul e b considered 9 » ligandd d an s sorption of cations viewed as surface complex formation. 2.3.4. Hysteresis in ion exchange isotherms Hysteresi n i anios n exchange involving simple anion t a constan sCl". ~ lik, " O .N e O tCNSS d "an solution acidity and ionic strength on hydrous thorium oxide (HThO) has been observed for the first tim® £73 (Fig.3a). Subsequent to our reporting, hysteresis in anion exchang n commerciallo e y available hydrous oxides s beeha n reported C19 1, although n generali , s i t i , claimed that anion exchange reactions on hydrous oxides are reversible £20.21,22]e th e reaso Th r fo .n hysteresi n anioi s t e nclearye On exchang t .no s i e could, however, tentatively sugges e differenceth t n i s the structure of the exchanging sites (for example, 5Me-OH2*N03 ~, IMe-OH 2*Cl", where =Me-O a surfac s i H e hydroxyl group), in the forward and reverse exchange as cause of the observed phenomenon. Further investigations are needed in this area to clearly understan e anioth d n exchange behaviou n hydrouo r s oxides. Certain cation exchange reaction n HTho s O have been found to be non-reversible C9], for example, between alkali meta a lC .pair iond an sf transitioo s n metal ions, like Cu -Zn , Cu -Ni . Cu -Co . The non-reversibilit o selectivt y e couldu e b ed bindinf o g ions as in the case of Ca * or due to specific sorption. which include electrolyte sorption, as in the ease of Cu *. However, NH *-Ca * exchange exhibited hysteresis (Fig. 3b). A combination of buffering 2 action of NH4* and precipitation of Ca * in solution, has been proposed to explain the observed behaviour.

254 2.4. Effect of t-radiation on the sorption of Metal Ions on Hydrous Oxides e impetuTh s e studgaineth f inorgani o n yi d n io c exchangers is due to the radiation stability of these e b expecte materialo t s di tha st I t[23 thes 1- e materials, when exposed to radiation in dry conditions, t woulno appreciabld y change their physicad an l chemical characteristics. However, when exposeo t d radiatio n i contacn t with aqueous solutionsn i s a , actual applications e inorganith , n exchangerio c s could undergo some change d hencan s e behave differentially aa resuls f formatioo t d interactioan n f o nradiolyti c 0 etc productH . ~ e [2*1f wate, o s t OH rNo ., H suc s a h many studies have been reported withe th regar o t d influence of in-situ radiation on the sorption properties of inorganic ion exchangers. Some preliminary investigations have been carried out with HTiO and magnetite (MAG) (representing hydrous oxide f o varyins g acidities) (Fig.4)e th o fint t , ou d influenc f o ^-radiatioe n fiel n o theid r sorption properties. These hydrous oxides also find possible application in radioactive waste disposal (HTiO) and in reactor coolant water systems (MAG). Sorptioi L f o n was studied, because it is a non-specific ion and its sorption would indicat e changee oxideth eth e n i du s to -f-radiation .

9 0.6 x I I I I I 7 o o-MA HTi- • O, G " 0.6 o> CC. UJ °- 0.4 O UJ O) oIT 0.2

0.0 5y

Fig. 4 Sorptiof o n n magnetitd o Lian ) G MA e( hydrous titanium oxide(HTiO) from LiCl*LiOH( -0.0 IM) solutions

The salient e investigationresultth f o s e ar s listed in Table II. Within the limits of the experimental error, there o changn appear e b n i eo t s o hydroutw e th s thn oxideo e uptaki L s f eveo e n after in-situ t-dose of >17 Mrad. The significant observatio e studth s thawa n n i ypresenci nt f o e radiation, the pH of the supernatant solution in contact with HTiO was higher than the unirradiated samples, in contrast to the pH of the supernatant

255 TABL : EffecI II -radiatioE f e o sorptiot th l n L o n f o n on magnetite (MAG) and hydrous titanium oxide (HTiO).

Time No irradiartion With irradiation

h pH * Li Amtf o . pH * Li f Amto . Initial Equil. sorbed Initial Equil. sorbed meq/g meq/g (±0.02) (±0.02>

MAGNETITE

4 11.27 10.92 0.05 11.27 10.90 0.07

24 11. 27 10.95 0.07 11. 27 10. 77 0.06

48 11.27 10.72 0.05 11.27 10. 38 0.06

67 11.28 10.76 0.08 11.28 10.4 0.10

HTiO

4 11.57 9.57 0.58 11.57 10.11 0. 56

24 11.57 8.15 0.58 11.57 9.00 0. 57

48 11.61 7.76 0.60 11.61 8.50 0.57

68 11.61 ao .o 0.60 11.61 8.27 0.60 * Time of equilibration or irradiation.. Conditions* Oxide 0.4g; Soin mixtura S . MA 25mlf LiOH+LiC o r e fo ; ( l0.01M ) for HTiO 0.01M LiOH. Strength of f-source 0.26M rad/h.

solution in contact with MAG, which was lower. There o n apparen s wa t e physicachangth n i e l natur f MAGo e . HTiO change s colouit do t yellor n o exposurw o t e radiation, presumably due to the interaction of H 0 (a produc f radiolysiso t )f HTiO o wit i T .h Further wors i needek n i orded o understant r d better the pH changes observed in this study.

3.0. APPLIED STUDIES WITH RESPEC O NUCLEAT T R ENERGY PROGRAMMES 3.1. Purification of Reactor Coolant Water One of the envisaged application of inorganic ion exchangers was their use as possible purification material e reactoth r fo rs coolant water, because th f o e expected thermal stabilit f thio y s clas f o materials s [253. Magnetite (MAG n attractiv)a appearee b o t ed material as it is already present in the reactor and its use could not add a new component to the reactor coolant system. Sorption studies with MAG would also lead to a better understanding of the activity transport phenomeno n reactoi n r coolant systems.

256 TABLE III: Sorption of corrosion product cations on hydrous oxides from a mixture of ions at 363 K under flow conditions [4]

Influent Amoun f traceo t n sorbedio d , pg/g Hydrous _-_____--__.__------_------____------__---_____-.._..___._ oxide Total amt. Amt. of o C n M e F f ionr o C s trace n io d presen n preseni t n i t mixtur e mixturth e e (ug) lug)

HZrO 6400 1 600 816 560 256 224

HThO 4000 1000 350 340 70 50

MAG 6400 1 SOO 1184 688 0 13 Conditions: Hydrous oxide lg; Soln. pH -6; Concn. of each ion 1 ug/ml; flow rate 50 ml/h.

3.1.1. Sorptio f corrosioo n n product cations Sorption of corrosion product cations (Fe *, Cr *. Co , Mn ) in trace concentrations on hydrous zirconium oxide (HZrO), HThO and MAG under dynamic conditions upto 363 K was investigated [43. The performanc e hydrouth f o se oxide s givei s n Tabli n e III. At higher pH, the sorption of Co * and Mn on these hydrous oxides increased, presumably due to presence of hydrolysed species.

3.1.2. Studies on mixed oxides With a view to develop an adsorbent capable of sorbing anions from alkaline solutions, different hydrous mixed oxides were prepare y incorporatinb d a g f higheo meta n io rl valency inte bulth ok oxide matrix. Among the several combination of mixed oxides tested, like Bi(III)-ThdV) , Bi ( 111 )-Zr ( IV ) . Bi( 111 )-Ti ( IV) . TidV)-Sb(V) etc. Bi( III)-Th(IV) , mixed oxide was found e mostb o t promising, sorbin " ionCl g s from alkaline solutions [63. Attempts were mado t improve e halidth en io e sorptio f Bi(111)-Th(IVo n ) mixed oxid y incorporatinb e g Ag* ion in the matrix [8]. By loading Ag ion on Bi(III)-Th(IV) mixed oxide, by equilibrating with AgNO solution o benefician , l effec s observedwa t . However, sorption of halide ion increased with increasing Ag* contene mixeth dn i toxide , whee oxid th s nprepare wa e d by coprecipitating Ag* along with Bi(III) and Th(IV) (Table IV). These material y finma e s dth n possibli e us e removal of halide ions from reactor coolant water and from spent fuel storage tank water. 3.2. Remova f Organio l c Radioiodine usin* Ag g Molecular Sieve

Anothe rr currenou aspec f o t interesn i s i t evaluating the basic properties of silver exchanged

257 TABLE IV: Sorption of halj.de ions on hydrous Bi ( 111 ) -Th ( IV ) mixed oxides (Bi/Tn . 4) [8] Sorption capacity , meq/g Test Solution Bi-Th mixed oxide Bi-Th-Ag coprecipita ted oxide Pure Ag loaded Ag* presene mixeth dn i t oxide, meq/g oxide oxide 9 2 . 0 0.50. 10 7 1.38

M NaC1 . l 0 1 .07 0.91 0. 16 0.46 0.70 1.38

M NaB1 . r 0 1 .04 0.90 0. 12 0.72 0.84 1.76

0 . 1M KI 0.94 1 .05 0. 17 0.50 0.79 1.46

O.OSM KI 1 .0 1 .01 0. 16 0.45 0.77 1.46

O.OSM KÏ » 0.98 1 .05 0. 15 0.47 0.77 1.51 M NaO10 H

+ I K M 05 . 0 0.96 0.89 0. 19 0.47 0.77 1 .54 10 M NaOH

O.OS» Î K M 0.92 0.90 - „ 10 M NaOH

0.05M K I * 0.42 0.49 - - 10* NaOH (a) Mixed oxide precipitated using NH OH; (b) Ag loading on the mixed oxide q Ag*/0.0me f 8oxide o g precipitatio) tc ; n effected using NaOH. Conditions : Oxide 0.5 g; solution 25 ml; equil.time 24 h.

TABLE V: Quantity of CH I adsortoed/consumed [26] (x 10" * 0.5 ml)

Temperature AB C 0 Hydrated Dehydrated Hydrated Dehydrated m 13X 13X Ag* 13X X 13 * Ag

298 0.84 23.6 13 .8 1 3 . 5 ( 16.5)* (36 .2)*

323 0.56 21.8 1 1 .2 30 . 1

375 0.42 16.3 2 5 .3 28 . 2

423 1 2 13.1 7 . 26 . 2 Molecular sieve ; useg I Silved r g contenA 13X n i ,t 1 .5 mmol ( )* with concommittent t dose o-f 0.4M rad h .

molecular sieves used for the removal of volatile organic radioactive iodine species from off gas streams of nuclear installations. The development of an e samefficienth et a tim d ean t less expensive adsorbent than silver loaded matrices demand a knowledgs e th f o e basic chemical processese th involve n i d chemisorption-decompositio e rolth f o e f o d an I H C f o n silver.

258 Our studies [26] covered identit d yieldan y f o s chemical products I e reactioformeove th H C rn i df o n plai d silvean n r exchanged (16X X molecula13 ) r sievs a e a function of temperature, humidity and ^-radiation field. Mechanistic aspects of the reaction were inferred from supportive evidence . from X-ray photoelectron spectröscopy e maiTh n. result e givear s n below:

) Hydrate(i d (182) plai X permit13 n s only negligible physiabsorption of CH I which is uninfluenced by concommittent f-radiation field (Tabl. V) e , (ii) Dehydrated (at 575 K for 6 hours) plain 13X shows significant adsorptio I nwhic H capacitC h r fo y decreases with increasing temperatur e capacitth d an ey s unaffectei y ^-irradiatiob d no N reactio(Tabl . V) e n products are observera at temperatures below 500 K and the adsorbed CH I could be desorbed at 600 K. Repeat e samth ef uso matrie r uptfo x5 cycle o s (adsorption and regeneration) show o n smeasurabl es it chang n i e behaviour. Ingres f moisturo s e during adsorption cycle drastically reduces the adsorption capacity. (iii o )t hydrate I ExposurH C d f o (18Ze ) silver exchanged (16Z) 13X produces two products namely. CH OH , th'H OC eH C anyieldd f o whics h increase with increasing temperature e Th adsorptio. n capacity increases with temperatur d -y-radiatioan e n dose (Table V). It appears that more CH I is consumed than is required for a stoichiometric reaction between CH I and silver which is particularly noticeable at higher temperatures (Table V, Column C). The reaction products CH OH and CH OCH ^do not react with bound iodine if present in Ag* 13X to regenerate again volatile organic iodides under the influence of temperature and/or f-radiation field. (iv) In the case of dehydrated Ag* 13X. although as expecte o oxygenaten d d reaction product e producedar s , smale lar e amounthydrocarbonth H C f o d san H C s released. the yields of which increase with temperature. Beyond 575 K free iodine is also released froe matrixth m . A few significant conclusions drawn from this study are

) mor(i e CHs i consume3I X thads 13 i n * oveAg r require a stoichiometri r fo d c reaction between d silvean I 3 r CH indicating participatio* Na f o n of molecular sieve in the fixation of iodine.

I interactio3 (iiCH ) n wits i h X hydrate13 * Ag d throug e formatioth h f methyo n l carboniun io m and

o dehydratetw f o (iii e dus ) plain molecular sieve beds in series followed by a small bed of hydrate * Ag moleculad r sieve coule b d considere n attractiva s a d e alternative th o t e use of a single Ag* molecular sieve adsorber e removath r f radioactivfo o l . I H C e

259 TABL : SorptioVI E f uraniuo n n hydrouo m s oxides from different uranyl salt solutions [11]

Uptake of uraniumn . mmol U/f o g oxide Hydrous _--____ oxide Sal* • t Nitrate Chloride Sulphate Tricarbonate used

HTiO)

HTiO( Am) 0 .66 0 . 52 0. 56 0. 34

HTiO(So) 0 . 81 0 .68 0.60 . 0 26

HCeOI Am) 0 .31 0 . 16 0. 42 0. 1?

HCeO(So) 0.43 0 .29 Q .33 0. 25

HZrO(Am) 0.20 Q . 12 0. 29 0. 13

HZrO) (So 0.33 0 . 1 <> 0 .36 0. 17

HThOI Am) 0 .05 0 .02 0.09 0.01

HThO(So) 0 . 15 Q .09 0. 08 0. 05 ) Oxid(H e homogeneously precipitated. (Am) Oxide precipitate d. (SousinOH H )N g Oxide precipitated using NaOH. Conditions: UO * - 0.01M: pH 3.5; for tricarbonate soin. -9; oxide 0.3g: soin. 25 ml; contact time 24 h.

3.3. Recover f Uraniuo y m from Sea-water

In recent years ther s beeha ea growin n g interest in developing sorbente recoverth r f fo uraniuso y m from sea-water. With this view in mind, different hydrous oxides, HTiO, hydrous cerium oxide (HCeO), HZrO, and HThO were screened in order to find a suitable sorbent for uranium [103. In general sorption of uranium decreased with decreasing acidic nature of the oxide,. Moreover, on acidic oxides like HTiO. HCeO, uranium was n leso sortoed s an s acidia urany dn io c l oxides like HZrO, HThO. electrolyte sorption, involving uranyn io l and anion s predominanwa , t [103. Uranium sorption also depended on the salt used for the sorption experiment [113 (Table VI). Sorptio s higwa n h whe a nitratn s wa e used and was low when carbonate was used. Among the hydrous oxides investigated, HTiO and HCeO were found to be more suitable for the sorption of uranium from carbonate solutions [113. More- detailed investigations were carried out with HTiO using uraniu n millimolai m r concentrations [13]. HTiO prepared usin t exhibigno NaOd di tH good sorption characteristic e s materialpresenth wa * n Na i ts a .s e sorptioTh n characteristics improve n removino d e th g Na* impurity (Fig.1). HTiO prepared using d ureha a large surface area and had only partially hydrolysed Ti , and exhibited good sorption property as such. In large scale preparation and operation, this particular material could prove useful, e unlikth e NaOH-precipitated material. becaus e partiallth e y

260 (ANION)G LO ,M -3.0 -2.0 -1.0 100 - T- 100

Fig.5 Sorption of Uranium on HTiO[13] as a function of : •.pH ;o. NO ~ concentration; " A,SconcentrationO . Equilibriu H valuep m r fo s anion e givear sn brackets.4.Precipitatioi n f o n uraniu a functio . s a mpH f o n

hydrolyse * site i T ds coul s da acidi alst ac oc sites, n i e surfacadditioth o t e n hydroxyl groups d binan , d more uranyl ions.

pH of the solution had a marked influence on the sorption of uranyl ion (Fig.5) than the added anions . " O SorptioS , " f o uraniuO nN . " likm Cl efro m carbonate solution e s limitee aciditth wa sth f y o b dy surface groups, becaus e sorptioth e n involved breaking e uranium-carbonatth f o e complex presen n i solutiot n [13] (Fig.1) e sorptioTh . f uraniuo n m from simulated sea-water was restricted by the competive sorption of Ca present in solution

REFERENCES

[1] SARPAL. S.K., GUPTA, A.R., J. Inorg. Nucl. Chem.. il (1981> 1347. [2] SARPAL. S.K.. GUPTA. A.R.. J.Inorg. Nucl. Chem.. 4198( 3 12043) . ] [3 VENKATARAMANI . VENKATESWARLUB. . , K.S., SHANKAR, J.. Proc. Indian Acad. Sei. A (197887 . ) 409. ] [4 VENKATARAMANI , VENKATESWARLUB. , , K.S.. SHANKAR, J.. BEATSLE. L.H.. Proc. Indian Acad. Sei.. 87A (1978) 415. ] VENKATARAMAN[5 I , VENKATESWARLUB. . K.S.. SHANKAR, J., J.Collois. Inrterf. Sei.. 67 (1978) 187.

261 [63 VENKATARAMANI, 8., VENKATESWARLU. K.S.. SHANKAR, J., 8EATSLE, L.H., 3,Colloid Interf. Sei.. Ifi (1930. 1 ) [7] VENKATARAMANI, B.. VENKATESWARLU. K.S., J. Inorg. Nucl. Chem., 42 (1980) 909. .[8] VENKATARAMANI , VENKATESWARLU8. , , K.S., Proc. Indian Acad. Sei., (Chem. Sei.), H (1980) 241. [93 VENKATARAMANI, 8., VENKATESWARLU. K.S., J. Inorg. Nucl. Chem., .il (1981) 2549. [103 MAHAL, H.S., VENKATARAMANI , VENKATESWARLU8. , . KoS., J. Inorg. Nucl. Chem., 43 (1981) 3335. M 13 MAHAL, H.S., VENKATARAMANI, 8., VENKATESWARLU, K.S., Proc. Indian Acad. Sei., (Chem. Sei,), 91 (1982) 321. [12] MAHAL, H.S.. VENKATARAMANI. 8.. Unpublished work. [13] VENKATARAMANI, 8., 6UPTA, A.R., Paper presented at the Int. Meeting on the Recovery of Uranium from Sea Water. Tokyo, Japan (Oet*. 17-19, 1983)), Paper No. 32. [14] MAHAL, H.S., VENKATESWARLU, K.S., India. J n Chem., 16A (1978) 712. [153 BHARGAVA. S.S., VENKATESWARLU, K.S., Indian Chem.. 14A (1976Î 800. [163 CLEARFÏELO , NANCOLLASA. . . S.H., BLESSING, R, . "Ion Exchange and solvent Extraction" (MARINSKY, J.A., MARCUS . Eds.Y , ) Vol.3, Marcel Oekkerw Ne , York Y (1973N . . 1 ) [173 ALBERTI, G., MASSUCCI, M.A.. J.Inorg.Nucl.Chem.. 32 (1970) 1719. [183 NANOAN . GUPTA , D . , A.R., India . 3 Chem.n , 12 , (1974) 808. [193 DYER . MALIKA. , , S.A. . Inorg3 , . Nuel. Chem.3 4 , (1981) 2975. [203 NANCOLLAS. G.H., PATERSON , J.Inorg.Nucl.ChemR. . . 29 (1967) 565. [213 RUVARAC. A. Lj., TRSTANJ.. M.I.. J. Inorg. Nuel. Chem., 34 ( 1972) 3893. [223 MISAK. N.Z., MIKHAÏL, E.M.. J.Inorg. Nucl. Chem., 43 (1981) 1903. [233 E6ROV, E.V.. NOVIKOV, P.O., "Action of Ionizing Radiation ExchangIo n o ne Materials", Israel Progr. Sei., Transi., Jerusalem (1967). [24] SELLERS, R.M., "The Radiation Chemistry of Colloid A Review" - s . Central Elect. Sen, Board, Berkeley Nucl. Labs., Report, RO/B/N 3707 (1976). [253 VENKATARAMANI , VENKATESWARLU8, . , K.S., "Water Chemistry Studies V!: Role of Inorganic Ion Exchangers in the Purification of Reactor Coolant Water", Bhabha Atomic Research Centre Rep. BARC/ 1-123 (1971). [263 BELAPURKAR, A.D., ANNAJI RAO. K., GUPTA. N.M., IYER, R.M., Surf. Technol. 1 2 ,(1984 ) 263.

262 ION-EXCHANGE SELECTIVITIES ON ANTIMONIC ACIDS AND METAL ANTIMONATES

M.ABE Department of Chemistry, . Tokyo Institut f Technologyeo , Ookayama, Meguro-ku, Tokyo, Japan

Abstract Antimonic acids and metal antimonates as the inorganic ion-exchangers exhibit extremely high selectivity for a certain element or group of elements for comparison with sulfonated polystyrene ion-exchange resin. Various antimonic acid materials have been obtained with different compositions and ion-exchange properties, depending on the metho theif do r preparation n o aginge s Th swela s a .l e speciedivideb n sca d into three groups -crystalline, amorphous and glassy. The affinity sequence for alkali metal ions shows Li < Na< K Na > K > Rb> Cs. These selectivities are discussed in the terms of steric effect and entropy changes of the ion-exchange reactions.

1. INTRODUCTION Synthetic inorganic ion exchangers have both thermal resistanc radiatiod ean n stabilit havd an ye been applien i d reactor chemistry such as reprocessing of irradiated nuclear fuel. Much attentio s beeha n n paid e recentlth o t y selective adsorption properties which are not found with the commercial organic resins. Durin e laso th decadesgtw t w ne inorgani, n io c exchangers have been synthesized and studied extensively for their ion exchange properties by various authors [1-7].

263 Some of the inorganic ion exchangers exhibit excellent selectivities with respect to certain element or group of elements selectivite Th inorganie . th n o yexchangern io c s increases usually with decreasing concentration of the elemene separatedb o t . Howevere e th changth , n i e selectivit mucs i y h depende n chemicao d l e specieth f o s elements, co-io ne crystallinitpresentth n o e d th ,an f o y adsorbent chemicae th n I .l processin f nucleao g r reactor cycle, the separation of the radionuclides may be needed fro larga m e e amounbulth kf to components. Overaln io l exchange reaction from microamoun o macroamountt e th f o s elements should be studied for this porpose. Furthermore, high stabilit acidie th n cyi solutio higt na h concentration e marequireb y d otherwis n higi e h resistanc o higt e h temperature and radiation field. Various antimonic acid materials have been obtained with different chemical compositions, characteristicsd an , cation exchange properties. The species can be divided into three groups- crystalline( C-Sb amorphous, ) A ( A-Sb, ) A and glassy( G-SbA ) [ 8-11]. Compounds similar to C-SbA have been reporte polyantimonis da hydrated an ) c A acidP ( d differeny pentoxidb ) P HA t ( eauthor s [12, 13]. Quadrivalent metal antimonates as ion exchangers have been [14] o studienumbeA It .d metaf an ro d e firslAb y tb antimonat s beeha en reporte y b varioud s authors. Especially, antimonate f titaniumo s ( d tinTi-Sban ) (A Sn-SbA ) exhibit an excellent selectivity for lithium ions [15, 16]. This paper describes compariso e selectivitth f no f o y cations on various antimonic acid and metal antimonates.

2. Experimental Preparation. 1 . 2 antimonif so c acid cation exchangers. The antimonic acid cation exchangers of different species were prepare s describea d d previousl. ] 0 1 yF Crystallin. 1 . 1 . 2 e antimonic(VL m ) acid5 7 (A C-Sb . ) A antimonf o y pentachlorid preliminarils ewa y hydrolyze5 7 y db mL of cold water, and was then hydrolyzed by 5 L of water. precipitate Th e obtaine e motheth s kep n dwa ri t solution at 40 °C over 20 days, and then washed with cold demineralized water until free from chloride ions using a centrifuge (about 10000 rpm). After dryine th g precipitate, the product was ground and sieved in the appropriate fraction of 100-200 mesh size. The collected sample was rewashed with cold demineralized water in order to remov y adherenean t fine particl obtaio t d a cleanean r supernatant solutio subsequene th n ni t experiments. 2. 1. 2. Amorphous antimonic acid ( A-SbA ). Similar procedure thoso st r C-Sb efo A were carried out, except aging overnigh n coli t d mother solutione Th . subsequent procedures are the same as above. 2. 1. 3. Glassy antimonic acid ( G—SbA ). The washed amorphous antimonic acid with cold demineralized wate s dissolvewatet rwa ho d coolera an n i dd rapidly s e a solutioth e quicklwate Th n s .i rwa yn evaporate possibls a d y usinb e g fan. After dryinge th , subsequent procedure same s aboveth a e e sar .

264 Preparatio. 2 . 2 metaf no l antimonate. e Th SnSbA exchange s wa preparer s a described d previously[ 15] . 2. 2. 1. Preparation of tin antimonate( SnSbA ). A liquid antimony(V) chlorid s prehydrolyzeewa d witn a h equal amount of demineralized water in order to prevent fractional precipitation of antimonic acid by further procedures. The 3.9M antimony(V) chloride solution obtained was mixed with 4.5M tin(IV) chloride solution at 80 e mixeTh d . solutio°C s immediatelwa n y hydrolyzen i d 50-fold volum demineralizef o e d wate e whit80°Ct rTh a .e precipitat mothee th keps en wa ri t solution overnight, then s filterewa water-washed dan y usindb a centrifuge(aboug t 10000 rpm) until the pH value of the supernatant solution was higher than 1.5. After drying at 60°C, the product was ground and sieved to 100-200 mesh size. The collected samples were rewashed with water in order to remove adherent fine particle of SnSbA, and finally dried to 60°C. 2. 2. 2. Preparation of titanium antimonate( TiSbA ) and zirconium antimonate( ZrSbA ) Similar procedures to those for SnSbA were carried out. The solutions of titanium(IV) chloride and zirconium oxide chloride were used instea f tin(lVdo ) chloride. 2. 3. Distribution coefficients(Kd). The equilibriu d valuemK metaf so l ionr microamountsfo s v,ere determined with a batch technique. The exchanger was intmerse e metath n li d salt solution containing hydrochloric acir nitrio d c acit a ddifferen t concentrationn i s thermostatted bath. The initial concentrations of metal ions were adjusted mainly to 10~4 M. The concentration of the e metasolutioth le exchangen s i th ionwa n n si d an r deduced e froconcentratioth m n relative initiath o t el concentratio solutione th n concentratioe i nTh . f metano l ions were determined by flame photometry or atomic absorption spectrometry. The Kd values were calculated as follows; _ Amount metaf so l ion exchangen i ssolutiof o L m „ r n Amounts of metal ions in solution g of exchanger d ThvakueK e f metao s l r ioncomparisofo s n were normalized to a 1M nitric acid solution by extrapolating the Kd values determined. 2. 4. Overall ion exchange reactions In the forward reactions, the exchanger( 0.10 or 0.20g) in H+ form was immersed in 10.0 or 20.0 mL of a mixed solutio varyinf no gmetae ratith lf o nitrate / nitric acid in a sealed glass tube with intermittent shaking at different temperatures. The ionic strength in the mixed solutio s 1 adjustewa afte0. n o rt d attainmene th f o t equilibrium. In the reverse reaction, the ion exchanger in various metal ion forms, corresponding in weight to 0.10 or 0.20 g of the exchanger in H+ form was immersed in 10.0 or 20.0 mL of the mixed solutions with ionic strength of 0.1 after equilibriation. The concentration of the metal ions in the exchanger and in the solution was deduced from the concentration relativ e initiath o et l concentratioe th n i n solution.

265 2. 5. Preparation of the exchanger in metal ion form. A O.Pî metal nitrate solutio s chargenwa d continuously at the top of the column of the exchanger in the hydrogen ion form until the change in the concentration of the metal ions was negligible between the effluent and feed. 2. 6. Reagents. The antimony(V) chloride (Yotsuhata Chemical Co . d tin(IVLtd.an ) ) chloride (Wako Chemica . Ltd.Co l ) were used without further purification. The other reagents used were all of analytical grade. 3. Theoretical Aspects The ion exchange reactions of metal ions/H+ exchange system on the exchanger can be represented by the following expression;

n H Mn+ n H ( 1 )

wher referr eba s metai o exchangest lM iond ran s with + valency n. The thermodynamic equilibrium above constantth ef o , ,K reactio difinee b n : nca as d

___, _ « . ______?~n V+ fH+ H+ M where K„ is the selectivity coefficient, m + and mj.jn+ are the molalities of HNOg and M(NC>3 )nin solution, and YH+ and Y,,n+are the ionic activity coefficients of H and M"+ in solution. XH+ and X^n+are the equivalent fractions of the exchanging H and M in the exchanger phase and T + and ?,, activite exchangee th th e n i ar y + rn coefficientM d an + H f so phase, respectively. The contribution of the ionic activity coefficiens in solutio thermodynamie th o t n c equilibrium e constanb n ca t neglected for uni-univalent exchange reactions, while not r uni-multivalenfo t reactions e ionie th valueTh cf .o s activity coefficient ratio { y^+/y n+)in solution have been used in the following equation [17]: n 2n YH+ '±HN0 n S /I log ———— =log 3 1+n 1.5/+ 1 1 Y ±M(N03)n

6 0 1/2 = 1.825S 2 Xv J_inU (\ > 4 ) e 3T3 > where P is the density of water, £ the dielectric constant oabsolute fth waterT d e ,an temperatur systeme th f o e. The thermodynamic equilibrium constan evaluatee b n tca d by using the simplified treatment of Gains-Thomas equation [18], assuming that the change of water content in the exchange d entrancan r f anioeo n from solution phase ar e negligible.

In K = - ( Z„ - Z„ ) + X d " ( 5 ) n H n M H M s i give , y K, b ne linearversu „ n ar I K e X n I sth , Whe e th n H M n M H

= 4-606 In

266 M where the C is the Kielland coefficient! 19], and (In KH) obtaine$ K n valua iI s extrapolatiny f b deo o zerot M .gX The change in the free energy of the ion exchange reactions calculatewa , ° G r catio ,A pe d n equivalent,

AG°= -( RT/ZM ZH ) in K 7 ) The e enthalpychangth n i e e entropy,, th AH° d ,an P AS were . ] calculate20 [ usuae y th lwa n i d The theoretical cation capacity( 5.057 meq/g ) of the exchanger s evaluateswa assuminy b d g thae antimonon t y gives one hydrogen ion available for cation exchange reaction. 4. Result and Discussion Selectivitie. 1 . 4 differenn so t antimonic acids. 4. 1. 1. Distribution coefficient Kd ) of microamounts. The determination of the Kd values gives useful information for the separation of trace and microamounts of the elements in the fields of analytical chemistry, radiochemistry, environmental chemistry d biochemistryan , . Nitric acid media have been used in order to obtain fundamental ion exchange data in the less complexing media, whic usefue ar hr applicatiofo l e reprocessinth o t n f o g nuclear fuel. Equilibrium distribution coefficients of alkali metal ions on C-SbA, A-SbA, and-G-SbA were plotted in Fig. 1, value n o Bio-Ras G 50W-XA d 8 being includer fo d comparison valued K r alkale fo s.Th i metal ionn o theis r inorganic exchangers, except C-SbA, increas ordee th < rLi n i e Na < K < Rb < Cs as shown with sulfonate-type organic resins e separatioTh . n factor e morar se favorable than

- 3

-2

5 1. W 6 0. • IR Effective cryatal ionic radius ( A ) y 574 Nal.02 K 1.38 Rl>U9 CS170 Fig. 1. The log Kd values of alkali metal ion on different antimonic acids

C-SbA; crystalline, A-SbA; amorphous, G-SbA; glassy alkali metal ions; 10~4M ,tota, M HNQ 1 l, j volmL 5 .2 exchangrer; 0.2. g 5

267 •'chose obtaine n Bio-Rao d d valueK G A 50W-X8 df e o sTh . metal ions on the Bio-Rad AG 50W-X8 are taken from the O-r.t^ reported by Strelow [21]. The increased selectivity was observe Bio-Ras a d d AG50W-X8

Vs 6 c

i i i i i i V i i i i i i i i i l l i i i

OS u S U *> Metal ions Fig. 2. The In KM of various metal ions on C-SbA in IM HNO, H J Conditions; sam s Figa e . 1

On the organic ion exchange resins, the Kd values of alkali metal e ionnearlar s y e solutioconstanth r fo nt containing alkali metal ion at the concentration lower than l m mol/L in a relatively high concentration of acid( > . ) However1 1 s bee 1 ha 0. nt ,i know d nk value thae th st change wit concentratioe th h e metan th loade io f nlo n o d some inorganic ion exchangers, although the acid concentration is unity and relatively high, because of their steric effects log-loe Th .valued K g s plotte v sth f so [H"1], were demonstrate n Figo d e wite , 3 .th changth h n i e n exchangio e Th e . ] 22 [ a B concentratio d an g M f no idealit maintaines i y relativelo t p u d y high concentration of 10~%, ] ,+ H indicatin [ g lo d g/ thaKd slopese g th t lo ,d are about 2. The Kd values increase with decreasing the concentration of the metal ions even in the same concentratio hydroniuf o n m ions. A systematic study of the distribution coefficient of metal ion C-Sbn o ss bee ha A n d co-workers reportean e Ab y db . Girardi and co-workers have reported that about 3,000 adsorption and elution cycle were carried out in a

268 I

10'

W jj o

0.01 0.1 1.0 [HNOj] (mol*/l)

e DependencTh Fig . 3 . e upoe initiath n l concentratioe th f o n metal ions for the Kd of Mg and Ba on C-SbA.

Initial concn.; (---)l(r2M, (——)10~3M, (_._)lCr4M,

standardize y ovewa dr eleven adsorbent s( includin9 g inorganic ion exchangers ) from various media for preliminary screening of possible material to use for radiochemical separation [12] ,comparativA n eio stud f o y 'exchange selectivity betwee P reportenHA C-Sbd y b an dA Girardi et al. was carried out by Abe. The evaluated selectivities on HAP are essentially compatible with those of C-SbA for 18 but out of 4 metal ions [ 23] . . 2 maximu . 1 . m4 uptak metaf o e l ion C-Sbn so A The maximum uptakes were determined by the column techniqu e maximuTh e mentione e m. sectio5 th . t 2 a dn uptake n inorganiso exchangern io c s have been know o nvart y extensively with the nature of cation species, temperature, concentratio f cationo n solutioni n d co-ioan , n present, while the organic ion exchange resins have fairly constant capacity due to their elastic structure. The maximum uptakes for alkali and alkaline earth metal ions, except Li and Mg on the C-SbA decrease with increasing effective ionic crysta e . ( lFigb ) ] n radiThi4 3 .2 ca s [ i explaine e effecth f o y steritb d c effect e maximuTh . m uptakes of Mn, Cd, and Pb was about 6 meq/g, which is higher thatheoreticae th n l capacit f 5.05o y 7 meq/ge Th . additional e capacitb explaine y ma y yb assumin d e th g adsorption of hydroxo-aquo species or polynuclear species in the exchange higt ra h concentratio f thesno e metal ionn i s hig mediH hp a [23] metae Th .l ions having small crystal ionic radii show strong tendenc more b eo t yhighl y hydrated thalargee nth r ions. This causes weak electrostatic force on the surface of the hydrated shell of the small cations. 4. 1. 3. Ion exchange 'isotherms on C-SbA. n exchangIo e isotherms have used to represent graphically experimental data pertineno t n io steri d an c sieve effect ordinate Th . f theseo e plots, e th s i , X

269 6.0

S.O Theoretical Capacity

4.0

2.0

l 1.0 D

I I I I 0.0 O _ ^«POwOCo-iZxiixuZumaXuzuNua— 3C-DJO . Metal ions Fig. Maximu4 . m uptake f metao s l ion n C-Sbo s A

equivalen Mn t n+ exchangee io fractioi nth e th f no r phasd an e the abbcissa, Xfjj , is in the solution. Cation exchange resins of strong acid type have essentially elastic structur d displaan e y swellin d shrinkingan g depending upon the exchanginnature th f eo g cations with hydration. They display isotherm st deviatwhicno o hd e greatly e froth m ideal ones. Inorganic ion exchangers generally have rigid structures and display curve in which KJj varies greatly with the ion exchanging cations. When small cations are exchanged with large cations, the ingoing cation is initially preferred, and the degree of exchange gives a value lower than unity. Thus, the ion exchange isotherm shows S shape curve for rigid exchanger, indicating reversal selectivity e selectivitth d an , y coefficient decrease progressively with increasing concentratio metaf no l ionn i s exchanger. Figs. 5-a ) and 6-a ) showed the typical ion exchange isotherms for alkali metal, alkaline earth metal , and Pb /H+ systems. The selectivity reversal was observed for the metal ions/ H+ ion exchange reactions listed, except for Mgd Li 2an + reaction+V H s [25, 26]. The plots of In KH vs XM , which are referred to Kielland plots d e 6-bshowan ar ,n Figs i e b n. Th 5- . selectivity coefficients decrease with increasing the concentration of metal ions in the exchanger. A fairly good linearlity was observed for the Kielland plots of the exchangn alio l e system listed, excep Ba^f to +Pb^Hd /Han + t systems. It has been known that the absolute values of the Kielland coefficient, C, indicate degree of the steric effect in the exchangers. The C values for different exchange systems are plotted as a function of the effective crystal ionic radius in Fig. 7, including the results for our earlier results. These plots show that cations with an effective crystal ionic radiu f o 1.0-1.s A ente" 2e th r exchanger with minimum steric hindrance. The C-SbA has a rigid structure without swelling or shrinking; the change in the lattice constan e conversioth less i tr sfo than% 2 n

270 -6 -e 0.1 0.2 0.3 0.4 OS exchangn M T Io Figs . v 5 .O e K isothermn plotd I an f ) so a ( s (b) for the system of alkali metal ions/H* on C-SbA.

Open mark; 30°C, closed mark; 45°C.

-5 -

Equivalent fractiosolution i M f no n Equivalent fraction of M in exchanger

M - exchangn ^ X Io s Figv . 6 ^ .eK isothermn plotd I an f so ) a ( s (b) for the systems of alkaline earth metal ions- and Pb^/H* on C-SbA at 30°C

Open mark; forward reaction, closed mark; reverse reaction

271 uo TJ | 10 4l

§ I 5 3 " .pÇa bi-Go ' .In3' Ce3' • .Mn2- . .Cd2' 5 1. 0 1. E.I.R. (A)

Fig. 7. Plots of absolute value of Kieland coefficient vs effective crystal ionic radius of various metals.

frohydrogee th m n for metao io n t m e n forms smalTh io l l. Ni?d e an ar hydrate+cations , Mg t d,Li e stronglth n i y aqueous solution and may be exchanged in keeping with their hydration w leavinn shelhydratei fe r lo a g d water from hydration shell. This brings large steric effectn si the rigid structure of the C-SbA. The exchange reaction of H.+ by the large cations with less hydrated becomes progressively difficult in the rigid structure. Thus, the minimum steric interpretee b effece C-Sb y th e ma Af t o th n i d both contributions. exchangn Io . 2 e. 4 selectivit metaf o y l antimonates. Quadrivalent metal antimonates as ion exchangers have bee o [14]nIt studie d .an de Especially Ab firs y b t , tin(IV d Titaniuan ) m antimonates( d SnSban TiSbAA , respectivel e catio) th behavy s na e exchangers with relatively high exchange capacity, and show the excellent selecivities for alkali metal ions in the order: Na

272 30 SnSbA

,0-—°—d 9--0-""Ro o Bi 0 or' AG50W-X8

1.0

l l -l l l l l l t l l l l l l 3 5 i u z u 2 a X < j < Z Ketal ions valued K e f variouo Th s Fig. 8 . s metal ion n SnSbo s n i A l M HNCL

It has been known that the ion exchange properties including their selectivities are much dependendt on their preparetive conditions and compositions. The Kd value of Li+ shows a maximum at about 1.7 of the Sb/Sn ratio, while increase with increasing Sb/Sn ratio for other alkali metal ions [15] selectivite Th . y sequence alkalf so i metal ions maintain in a range from 0.5 to 2.5 of Sb/Sn ratio. An extremely high selectivit a whic, s observe s i wa yL hwa r fo d good advantage for separation of Li from other alkali metal ions like seawater [29]. 4. 2. 2. Selectivity of TiSbA for microamount of metal ions. TiSbAe Onth , slightly different selectivities wite hth SnSbA was observed; Na n Io exchang . 3 e . 2 isotherm . 4 d an Selectivits y coefficients for alkali metal ions/ H+ system. The ion exchange isotherms observed at 30°C are shown in Fig,10. Eac e e revershresultth th plo f f o o est reactio e e isotherplotteb tn n n ca o dm e curveth f o s forward reaction. Thus, reversible exchange reactions are establishe o hysteresin d an d s effec s observedi t e Th . isotherms observed showeshapS e eth d curves, indicatine th g selectivity reversal [31]. The Kielland_plots showed inflection points at about + + 0.04 and 0.14 of XM for Li /H+ and other metal ions / H systems. An extremely high selectivity was observed for Li at the concentration lower than X = 0.04. This indicates that preference sites for Li are present at lower exchange composition. Other alkali meta t approaclno ionn o t sca h the sites because of large crystal ionic radius. The

273 2.0

--o—o-'"d Ro o Bi 0 AG50W-X8

1.0

TiSbA

-1.0

SA SM S l Metal ions d valueK e f variouo Th s Fig . 9 . s metal ion n TiSbo s n i A H HNOl „

0.3 Li

0.2

n exchangIo Fig. s v 10 . e ^ K isotherm n d plotI an f ) o s (a s (b) for the systems of alkali metal ions on SnSbA at 30°C.

maximum uptake was found to be 1.13, 0.98, 0,77, and 0.73 meq/g for Li, Na, K, Rb, and Cs, respectively. This behavior can be explained in the terms of the selectivity reversal at these infection points.

REFERENCES AT1PHLETT] l Inorgani, [ B. . C ,Exchangersn Io c , Elsevier, Amsterdam, 1964. [2] FULLER, M.J., Chromatogr. Rev., 14 (1971) 45. [3] VESELY, V., Pekarek, V., Talanta, 19 (1972) 219,1245. [4] ABE,IL, Bunseki Kagaku, 23 (1974) 1254, 1561. ALBERTI] 5 [ Ace, ,G. . Chem. Res.,11 (1978) 163. Sep, . Sen, K .K. . A ,. SeiA , .De ] Technol.[6 .(19783 ,1 ) 517. ] CLEARFIELD[7 Inorgani, ,A. c Ion-exchange MaterialsC ,CR Press, Inc., Boca Raton Fl., (1982).

274 [8] ITO, T., ABE, M., Nippon Kagaku Zasshi, 87 (1966) 1174. [9] ASE, II., ITO,T., Bull. Chem. Soc. Jpn., 40 (1967)1013. [10] ABE, II., T. ITO, Bul. Chem. Soc. Jpn., 41 (1968) 333. [il] ABE Bull, M. , . Chem. Soc. Jpn. 2 (19694 , ) 2623. Î12] GIRAPJJI, F., SABBIONI, E., J. Radioanal. Chem., 1 (1968) 169. [13] 3AETSLE, S., HUYS, D., J. Inorg. Nucl. Chem., 30 (1968) 242. [14] ABE ITO, Kogy, M. , T. , o Kagaku Zasshi 0 (1967,7 ) 440. [15] A3E, M., HAYASHI, K., Solv. Extract. Ion Exch., 1 (1983) 97. [16] ABE, M., TSUJI, II., Chem. Lett. 1983, 1561. Î17Î KRAUS, K..A., RARIDON, R. J., J. Phys. Chem., 63(1959) 1901. [18] GAINS Jr.,G THOIÎAS, .L. , H.C. Chem. J , . Phys (19531 .,4 ) 714. [19] BARRER, R. M., FALCONER, J. D., Proc. R. Soc., A236 (1956) 227. [20] BARRER, R..M.,et al., Proc. R. Soc., A273 (1963) 180. [21 al.t e ] , ,STRELOWE. Anal. W . .,F Chem. 7 (1965,3 ) 106. [22] ABE, II., UNO, K., Sepn. Sei. ,Technol. 14, (1979) 355. [23] ABE Sepn, M. , . Sei. Tech. 5 (19801 , . )23 [ 24] SHANNON, D. D., PREWITT, C. T., Acta Crystallogr. B23 (1969) 925. [25] ABE, M., J. Inorg. Nucl. Chem., 41 (1979) 85. [26] ABE, M., SUDOH, K., J. Inorg. Nucl. Chem., 43 (1981) 2537. [27] EISENI1AN 9 (1962) Biophys25 , , G. ,2 ., .J. [28] ABE FURUKI, 44te M. Th , h , Annua,N. l fleetin Jpnf o g . Chem. Soc., paper 2F19, Okayama (1981) [29] AEr, II., HAYASHI Hydrometallgy, ,K. 2 (19841 , . )83 [30] ABE, Tl., CHITRAKAR, R., Unpublished Result. [31] ABE, Tl., FURUKI, N., The 47th Annual Meeting of Jpn. Chem. Soc., paper 3T15 Kyoto (1983)

275 PANEL DISCUSSION

Chairman: H.J. Ache (Federal Republic of Germany)

SUMMARY AND RECOMMENDATIONS

I. Summary of the discussion

Inorganic ion exchangers and adsorbents have proved their potential usefulness in various areas of nuclear fuel cycle technology such as:

1) In the separation of fission products and actinides from medium high-activitd -an y wastes.

2) In the treatment of effluents from nuclear power plants.

chemicas A ) 3 l barriers (backfill material undergrounn )i d repositorie radioactivf so e wastes.

Improvement made b somn i en esca techniques thae alreadar t y being used by the industry, such as: e adsorptioTh ) 1 f iodino n e from dissolver off-gasesd ,an

14 fixatioe Th krypton-8f ) no 2 d C05an 2 .

Further research and development should be directed towards the optimizatio chemicae th f physicand o an l l propertie f theso s e inorganic compounds for their application in the fields listed above. Considerable fundamental research is still necessary to have a better understanding of the mechanisms involved in adsorption and ion-exchange and of the correlation between structur efficiencyd an e .

. RecommendationII s

Regarding future activities participante ,th s recommend that international cooperation shoul encouragee b d maie th n n do topi f o c research on inorganic ion-exchangers and adsorbents for the separation of fission products and actinides from medium-and high-activity wastes and

277 for the treatment of effluents front nuclear power plants, including the following sub-topics:

) a Preparation; b) Characterization; c) Applications; d) Process and engineering aspects, and e) Economics.

Special efforts should be made to promote a greater exchange of information and closer cooperation between the specialists who are working in basic research and those who are primarily concerned with the practical application f thesso e compound nuclean i s r fuel cycle technology.

278 LIST OF PARTICIPANTS

BELGIUM Baetsle, L. Chef du Department Chimie CEN/SCK B-2400 Mol, Belgium

Centner. ,B Ingénieu bureau ra u d'étude d'Electrobel S.A. Place du trône l lOOO Bruxelles, Belgium

Jacqmin. ,J Belgonucleaire Chamu d Mare e pd Ru s , 25 B - 1050 Brussels, Belgium

CHINA. PEOPLE'S REPUBLIC OF Weng Haorain Beijing Normal University Divisio Radiochemistrf no y and Radiation Chemistry Beijing, China, People's Republic of Zhu Yong-jun Institut Nucleaf eo r Energy Technology Tsinghua University Beijing, China, People's Republif o c

EGYPT Misak, N.Z. Prof, of Physical Chemistry Nuclear Chemistry Department Atomic Energy Post 13759 101 Kasr El Einy Street Cairo, Eygpt

FINLAND Lehto, J. University of Helsinki Departmen Radiochemistrf to y Unioninkat5 u3 Helsinki, Finland

FRANCE

Dozol, J.F. CEA/Centre d'Etudes Nucléairee d s Cadarache Services d'Effluents et Déchets de Faible et Moyenne Activité F-1311 Pau. Durances 5St l Le , France Madie. ,C CEA, Section des Transuraniens 6 . B.PNo . F-92260 Fontenay-auz-Roses France

279 Regnaud. ,F CEA/DCAEA/Service Etudes Analytiques B.P. No. 6 F-92260 Fontenay-aux-Roses France

GERMANY. FEDERAL REPUBLIF CO

Ache, H.J. Nuclear Research Center Karlsruhe Institute of Radiochemistry Postfach 3640 750~ D 0 Karlsruhe Germany, Federal Republif o c

Merz, E. Institut Chemicaf eo l Technology Kernforschungsanlage Julich GmbH D-5170 Julich, POB 1913 Germany, Federal Republif co Sameh.A.A. Nuclear Research Center Karlsruhe Institute of Radiochemistry Postfach 3640 D750- 0 Karlsruhe .Germany, Federal Republic of

HUNGARY Szirtes. ,L Institute of Isotopes of the Hungarian Academy of Sciences P.O.Bo7 x7 H-1S25 Budapest Hungary

INDIA lyer, R.M. Head of Chemistry Division Bhabha Atomic Research Coentre Trombay Bomba 0850 y40 , India

ITALY Alberti. ,G Dipartimento di Chimica Universit Perugii ad a Via Elc Sotti ed 0 o1 1-06100 Perugia, Italy

JAPAN Abe, M. Departmen Chemistrf to y Tokyo Institut Technologf eo y 2-12-1 Ookayama Heguro-ku Tokyo, Japan Ito, T. Deputy Director Departmen Radioisotopf to e Production Radioisotope Center, JAERI Tokai-mura, Ibaraki-ken, Japan

280 Kanno. ,T Research Institut Mineraf eo l Dressing and Metallurgy Tohoku University 1 1 Katahira-Chome Sendai, Japan

Saturai, T. General Manager, Physical Chemistry Lab. Departmen Chemistrf to y Tokai Research Establishment JAERI Tokai-mura, Ibaraki-ken Japan

PARISTAU Chaudhri, S.A. Alternate to the Resident Representative Permanent Mission of Pakistan to the IAEA Hofzeile A-1190 Vienna

SWEDEN Airola. ,C Nuclear Division, Waste Systems Studsvik Enefgiteknik AB 6118- S 2 Nykoeping Sweden

UNITED KINGDOM

Harding. ,K UKAEA ABE Winfrith Dorchester H Dorse8D 2 tDT U. K. Hooper, E.W. Separation Processes Group Chemical Technology Division AERE Harwell Ozon 0X1A 1OR U. K. Eccles, H. Building 643, BNFL Springfield Works Salwick Preston, Lanes.J OX 4 ,PR . K . U

UNITED STATES OF AMERICA

Carl, D. Senior Engineer/Solidification Engineer West Valley Nuclear Services Co. Inc. P.O. Box 191 West Valley, New York 14171-0191, U.S.A.

281 Clearfield, A. Department of Chemistry Texa UniversitM sA& y College Station X 7784,T 3 U.S.A. Collins, E.D. Manager Three Mile Island Assistance Programs Ridgk Oa e National Laboratory Oak Ridge 3783N ,T 1 U.S.A.

YUGOSLAVIA Ruvarac, A. The Boris Kidric Institute of Nuclear Sciences 11001 Belgrade 2 52 x Bo . P0 o Yugoslavia

ORGANIZATIONS

INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA) Ajuria . ,S Divisiof no Nuclear Fuel Cycle Rybalchenko, I. Division of Nuclear Fuel Cycle Ugajia, M. (Scientific Secretary) Division of Nuclear Fuel Cycle

282