Advances in Uranium Refining and Conversion Iaea, Vienna, 1987 Iaea-Tecdoc-420

IAEA-TECDOC-420

PROCEEDINGS OF A TECHNICAL COMMITTEE MEETING ON ADVANCE URANIUSN I M REFININ CONVERSIOD GAN N ORGANIZEE TH Y DB INTERNATIONAL ATOMIC ENERGY AGENCY AND HEL VIENNAN D» , 7-10 APRIL 1986

A TECHNICAL DOCUMENT ISSUEE TH Y DB INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 198? The IAEA doe t normallsno y maintain stock f reportso thisn i s series. However, microfiche copies of these reports can be obtained from

ClearinghousS I N I e Internationa! Atomic Energy Agency Wagramerstrasse5 P.O. Box 100 A-1400 Vienna, Austria

Orders shoul accompaniee db prepaymeny db Austriaf to n Schillings 100,- in the form of a cheque or in the form of IAEA microfiche service coupons which may be ordered separately from the I MIS Clearinghouse. ADVANCES IN URANIUM REFINING AND CONVERSION IAEA, VIENNA, 1987 IAEA-TECDOC-420

Printe IAEe th AustriAn i y d b a May 1987 FOREWORD

One of the most important steps in the Nuclear Fuel Cycle is the Uranium Refining and conversion which goes from the yellow-cake to three different products: uranium dioxid, natura) O (U el metallic uraniu) (U m and uranium hexafluoride (UF ) . Although its added value is small, may 6 be 2-4% of the total cost of front-end, it is decisive for producing fuel r botfo h type f reactorso s : natural uranium (Magno d Canduan x d )an enriched uranium (LWR) n thiI . e cycls e uraniuth ste th ef o p m hexafluoride is produced which is very important for the enrichment of uranium.

e totaTh l volum f thio e s industry e presenth t a , t time nearls i , f o y 40,000t U per year and at the end of the present century it would have reached the 60.000t U/a level.

e AgencTh y hel n Advisora d y Group meetine productioth n o g f o n yellowcake and uranium fluorides in Paris, June 1979 and its proceedings were published in 1980. During these seven years, significant improvements have been mad n technologi e equipmenn i e d th an yr fo t uranium refinin d conversionan g , particularly froe environmentalth m , safety and economic viewpoints. The refining and conversion of reprocessed uraniu e mextracteb than ca t y treatinb d g irradiated fuel become equally importan r recyclinfo t g recovered fuel.

In response to the growing interest in these topics, the IAEA convened a Technical Committee Meeting on "Advances in Uranium Refining and Conversion Headquarters it t "a s fro , 198m10 6 Aprio t wit e 7 l th h attendanc 7 expert3 f o e s fro 1 countries2 m . This Technical Document contains the 20 papers presented during the meeting.

The Agency wishes to thank all the scientists, engineers and institutions who contributed to this Meeting with their papers and their participation. Chairmene Speciath o t le ,thankdu Messrs e ar s . PagH . e (U.K.), A.W. Ashbrook (Canada) Faro. ,R n (France) Leyse. ,E r (Federal Republi Germany)f o c , A.G.M. Jackson (South Africa), I.S. Chang (Republic f Koreao d J.A)an . Vercellone (Argentina) e officer IAEAe Th .th ,f o s responsibl e organizatioth . Ugajir M meetinge fo e. d th Mr an nf o s n ,wa r editinfo documene . J.Lth gMr . Nucleas e Rojatwa th f so r Materiald an s Fuel Cycle Technology Section. EDITORIAL NOTE

preparingIn this materialpress,the International the for staff of Atomic Energy Agency have mounted and paginated the original manuscripts as submitted by the authors and given some attention presentation.the to The views expressed in the papers, the statements made and the general style adopted are responsibilitythe namedthe of authors. necessarilyviewsnot The do reflect govern-thosethe of 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 and institutions or of the delimitation of their 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

PAR RECEN— TI T ADVANCE REFININE TH N SI FRESF GO H URANIUM MATERIALS e refininTh conversiod an g f uraniuno m yellowcak uraniuo et m dioxidd ean uranium hexafluoride fuels in Canada: Current processes...... 9 AshbrookW. A. Recent advance presend an s t statu f uraniuo s m refinin India...... n gi 1 2 . N. Swaminathan, S.M. Rao, A.K. Sridharan, M. Sampath, V. Suryanarayanan, KansalK. V. Operating experience of a pilot plant for the production of uranium dioxide from uranium ore concentrate...... 29 M. Shabbir Purification and conversion of uranium from iron and thorium containing deposits...... 43 H. Movaseghi, MeissamiN.

Developmen technologa f o t mako y2 t startin eUO g from "yellowcake" refined with amine sulphuria n i s c environment...... 9 4 . J.A. Vercellone Status of uranium refining and conversion process technology in Korea...... 63 7.5. Chang, S.T. Hwang, J.H. Park

Research and development of UF6 conversion in Japan now and subjects in future...... 81 Y. Hashimoto, L Iwala, T. Nagasaki Experience in yellowcake refining and its conversion to uranium tetrafluoride at IPEN-CNEN/SP...... 103 A. Abrao

Conversio f non-nucleano r grade feedstoc UFo kt 4...... l Il . A.A. Ponelïs, M.N. Slabber, C.H.E. Zimmer The conversion from uranium tetrafluoride info hexafluoride in a vertical fluorization reactor...... 1 14 . Zhang Zhi-Hua, Chao Le-Bao e solvent-containinTh g resi r removinnfo g uranium from effluen f refinino t g process...... 9 14 . Zhu Chang-en, Zhou De-yu Waste managemen e refininth n conversiod i t an g f uraniuno Canada...... mn i 7 15 . A.W. Ashbrook e regulatioTh f uraniuno m refinerie conversiod an s n facilitie Canada...... n i s 7 16 . J. P. Didyk Conversio f uraniuno e concentratemor reprocessed san d uraniu mnucleao t r fuel intermediate t BNFsa L Springfields Part A: Uranium ore concentrates...... 185 H. Page

PAR REFININ— I TI IRRADIATEF GO D URANIUM MATERIALS

Conversion of uranium ore concentrates and reprocessed uranium to nuclear fuel intermediate t BNFa s L Springfields Part B: Reprocessed uranium...... 207 . Page// Nitrox process: A process developed by COMURHEX of continuous denitration...... 223 R. Romano Recent development e purificatioth n i s f uraniuo n m recovered from irradiated materials...... 1 23 . Regge,P.de Collard,G. Daniels,A. Huys,D. SannenL. e reprocesseTh d uranium conversion yearn te :f operatioo s f COMURHEX...... no 5 24 . R. Faron Problem impuritieo t e sdu uraniun i s m recovered fro reprocessine mth f go fuelR use,dLW fro e poinmth f vieo t f recycling...... wo 5 25 . E. Leyser Conversio f reprocesseno d uraniu Japan...... n mi 7 26 . /. Yasuda, Y, Miyamoto, T. Mochiji

Panel Discussion...... 9 27 .

List of Participants...... 283 PARTI RECENT ADVANCES E INTH REFININ FRESF GO H URANIUM MATERIALS THE REFINING AND CONVERSION OF URANIUM YELLOWCAKE TO URANIUM DIOXIDE AND URANIUM HEXAFLUORIDE FUEL CANADAN SI : CURRENT PROCESSES

A.W. ASHBROOK Eldorado Resources Ltd, Ottawa, Ontario, Canada

Abstract

In Canada, uranium yellowcake is now refined at Blind River and e uraniuth m trioxide produce souts n i shipped ki o Port h0 50 td Hope where it is converted to uranium dioxide and hexafluoride fuels.

There have been some significant changes in the processes used n thesi plantsw ne e . Perhap e mosth s t significan e productioth s i t n of uranium tetrafluoride from uranium trioxide using a "wet" process. In this paper the current processing methods are reviewed, with special emphasi n changeo s s froe previouth m s methods.

INTRODUCTION Eldorado Resources Ltd. has operated uranium refining facilitie t Pora s t Hope, Ontario, sinc s note wa .1942 t I . however, until 1955 that solvent extraction was used to produce nuclear grade uranium trioxide. In 1958 the production of uranium dioxid s commencewa e o providt d e fuer domestifo l c CANDU reactors. Conversion facilities were adde n 197i d o t 0 produce uranium hexafluorid r exportfo e . n 197I 5a decisio s mad refininwa o nw built e ne a d d an g conversion facility in the Port Hope area. The outcome was. however, that a new refinery was built at Blind River in northern Ontari nortom k (somf Poro h0 w ne 50 et a Hope d an ) uranium hexafluoride plant built at the existing Port Hope facility. The Blind River refinery came on-streara in 1983 and e Porth t Hope conversion facilit n 1984i y . Today, in Canada, uranium yellowcake is refined at Blind River and the uranium trioxide produced is shipped 500 km south to Port Hope s wheri converte t i e o uraniut d m dioxidd an e hexafluoride fuels. UO3 Powder lo Port Hop« I___ j Nmogen Ondes 10 ! Acid Recovery

Fig. I. Refining of Uranium Yellcwcak, UO o t e

Acid Recycle lo Mill Recycle j lo Digestion Tnick- There have been some significant changee processeth n i s s used in thes w plantsne e . Perhap e mosth st significane th s i t production of uranium tetrafluoride from uranium trioxide using a ''wet'' process n thiI .s pape e currenth r t processing method e reviewedar s , with special emphasi n changeo s s froe th m previous methods.

THE REFINING PROCESS The refining of uranium yellowcakes (concentrates) to nuclear grade uranium trioxide e Blin(UOth dt 3 )a Rive r facilitf o y ERL is shown schematically in Figure 1. It varies little from the process used previously at Port Hope. The component steps in this process are: o weighing, sampling, analysing and blending of the concentrates o digestio e concentrateth f o n n nitrii s c acid o purificatio e digesteth f o n d concentrate y solvenb s t extraction o evaporatio e purth e f o uranyn l nitrate solution resulting from solvent extraction to uranyl nitrate hexahydrate (UNH) o denitration of the UNH to produce UO3 o recovery of nitric acid from the SX raffinate by evaporation with sulphuric acid o recycle of the concentrated raffinate to uranium mills o shipping UO3 to Port Hope Each of the above component parts of the process is briefly described below. Weighing and sampling, etc. This is generally the normal approach taken in the industry. An auger sampling technique was adopted for the BRR. although both auger and falling stream methods had been employed at Port Hope. Digestion Concentrate e dumpe ar so digestio t dd frofe e drum th d mn an s tanks in a continuous two-tank cascade system. Dissolution is with concentrated (13M) nitric aci o product d a eslurr y containin a fre t a e gU acig m~- k *abou d0 concentratio40 t f o n less than 2M. Solids concentration in the slurry is normally les s comprisei r centtha pe d 1 an n, d largel f silicao y . Phosphori e addecb acit thiy a d ma ds stag o reduct e e thorium extractio e solventh n i nt extraction circuit. e two-tanTh k system operates continuousl d feedan y s four hold (feed) tanks, which then feee solventh d t extraction circuit. Off-gases from digestion - largely nitrogen oxides - are fed throug a venturh i scrubber e resultanth d an , t liguor returno t s e digestioth n tanks. Non-condensible e nitrith o t c o acig s d recovery circuit (see below).

11 Solvent Extraction This circuit differs from that employed previousl e Porth tt a y Hope facility in that it uses Mixco (Oldshue-Rushton) - rather than pulse - columns, which are arranged in two parallel trains. Each train contain 3 columnss e eacr on ,fo h extraction, scrubbin d strippingan g . Dimensions of the extraction columns are 11.3 m in height and n diameteri m e 1 scru Th 1. d stri. an b p column e smallerar s , n diameteri m 1 o reducn heighT 1. i . d m bein ean t0 9. g sntrainment carryover between columns, solvent settling tanks are employed. These also provid r somfo ee surge capacitn i y the circuit. The digestion feed is contacted with a solvent comprising 22 vol per cent tributylphosphate (TBP) in a saturated hydrocarbon diluenA ratiO/ e o - ISOPAtTh use n extractio i . dM R s aboui n t 4, and at a temperature of approximately 60°C. Scrubbin f impuritieo g s froe loadeth m d solvent e occurth n i s second columns achievei d an , d usin A ratiO/ g f n wateo oa t a r . Scri-15 b liquo s recycledi r . Finally, the uranium is stripped from the solvent phase, again with water but at an O/A ratio of 1.2, to produce a pure uranyl nitrate solution normally K liquorreferreO s a o .t d As in all solvent extraction processes used in the refining of uranium, thera nee o maintais t i de n solvent purity A blee. d (-10%) of stripped solvent is treated with a sodiuia carbonate solution prior to recycle to extraction. After contacting with the carbonate solution the aqueous and solvent phases are separated by centrifuges. The solvent phase is returned to the extraction circuit, while the aqueous phase containing the impurities is acidified and the resultant organic (non-soluble) produc s filterei t a pre-coa n o d t pressure tube filter.

Evaporatio K LiquoO e th r f o n The OK liquor resulting from stripping the purified uranium from the solvent normally contains about 100 kg m~3 of uranium. Thi s evaporatei s d concentratean d o somt d e

1200 kg m~ uranium (uranyl nitrate hexahydrate, UNH) in a three-stag3 e evaporator. Evaporation is achieved in the first siag y steab e m heatinge seconth d thirn an I d .d stages. overhead steam from the preceding stage is used as the energy source. After the third stage the steam is condensed, and used as the strip water in the solvent extraction circuit. Product from the third evaporator stage, molten UNH. is contained in a heated tank which feeds the denitration pots. Denitration s carrieheavy-walledn i i 3 t 00 ou d o t Denitratio H ,UN e th f o n semi-spherical vessels stirred by an agitator. This is the same technology employed previousl e Porth t t Hopa y e facility. d continuouslfe s i e denitratio H th UN o t ye Th n pots (currently 12 in all) where it is thermally decomposed at about 280°C to e reactioUOTh 3. n products, nitrogen oxides, acidd an , water, are scrubbed to condense water and acid. The scrubbed vapors o absorbert the o g n o product s e nitric acid. Raffinate Treatment and Recycle Raffinate froe solventh m t extraction circuit accountr fo s almost all o£ the wastes generated at the BRR. It is recycled, after treatment o uraniut , m mill t Ellioa s t Lake,m k som 0 5 e north of the Refinery. Because of the importance of raffinate recycle, a special circuit was installed in the BRR. Raffinate from the solvent extraction circuit contains nitrate salts and impurities from the uranium concentrates. It is necessar o removt y e nitric acid d thi an s ,achievei s y b d evaporatio e raffinatth f o n e wite additioth h f sulphurio n c aci o decompost d e nitrate salt d producan s e sulphated an s nitric acid. Froe extractioth m n e columsolventh n i nt extraction circuit the raffinate is pumped to hold tanks where it is settled, and then pumpe a raffinat o t d e evaporation circuit comprising three evaporation stages. Partial evaporation in the first stage is effected in a shell and tube heat exchanger employing the excess heat in the overheads from the third stage of the OK liquor boildown evaporator. Vapors from this stage (dilute nitric acid) go to the first nitric acid concentrator. e seconIth n d stag f raffinato e e evaporation, sulphuric acis i d added resulting in the conversion of nitrates into nitric acid waich is sent to the second concentrator for recovery. In the final stage of evaporation, the concentrated raffinate is drawn off int og ns~ k tank a densit3.7 t 1. a s d f betweeo yan 6 1. n Generall e sulphurith y c acid concentratio s aboui n0 g/L45 t , nitrate content les sr cent thape d contain 1 an ,n s uraniut a m g ra~k 33 . 1- The concentrated raffinate is transferred to tank trucks and transported to the uranium mills at Elliot Lake to be introduced inte millinth o g circuits. Here e sulphurith , c acid in the raffinate contributes to the dissolution of ore, and the uranium in the raffinate is recovered in the mill. Impurities e raffinatith n o witg e emil th h l tailing e tailingth o t s s pond. Nitric Acid Recovery Recover s achieve i o f aci nitrio R ytw BR dn i e dc th aci n i d absorber o acitw d d concentratorsan s , eact operatinse h n i g parallel. Off-gases from the concentrate digester and denitration fume scrubbers pass throug e acith hd absorbers where n contaci , t with water, nitric aci s producedi d . Uncondensed nitrogen oxide from the absorbers enter a catalytic converter which converts the NOX to nitrogen and water before exhaust to atmosphère e aqueouTh . s (dilute nitric acid) phase from each

13 o—o-^ ^=^ ^—^ F'suiter

Fig. 2. Conversion of UO_ to UF, acid absorber, together with vapors from the second and third stage raffinate evaporators, is sent to the acid concentrators. Here e nitrith , c aci s concentratei d o about d t 13M. which is then returned to the digestion stage. Uranium Trioxide Transportation e 110Th 3 product froe refinerth m s shippei y o Port d t Hopn i e t tot10 e bins .s fillei Eacn dbi h froo denitratiotw m n pots (concurrently), and is sampled continuously. Bins stand on load cells during fillinge accuratelar d an , y weighed prioo t r shipping. After analysis of the sample for each tote bin, they are shipped to Port Hope on a specially designed flat-bed truck, which hauls three bins, or about 30t (net) of 1103 (39 gross weight).

E CONVERSIOTH N PROCESSES The processes used for the conversion of 003 to both UFg e showar n2 an UÛ schematicalld e Th . 3 Figuren i yd an 2 s major differenc n conversioi e w ne e nth carrien i t ou d conversion facilit t Pora a y tf o Hop e eus e (UFgWth s )i "wet-way" e productioprocesth r fo s f UF4o n . Developey b d Eldorado, the process provides several advantages over the dry process employed in the east UF$ plant, which was brought on-strea n 1970i m .

Off-Gas Off-Gas

Aqueous Ammonia J

Ammonium Diuranate Ammonium Slurry Nitrate for Fertilizer

Blende Drumn i a dsUO to Canadian Fuel Fabricators

2 FueFig U0 . Conversio3 .lo t Powde 3 UO rf o n

15 COMPONENT STEPS IN THE CONVERSION OF UP3 TO UF6 ^ procesA schematiUF e s showth i s e f n Figuro Th i cn . 2 e component steps in this process comprise: o UC>3 receiving o conversioo UC>t 2^ UO f o n o conversio4 UF f UC>o o nt 2 5 UF o conversioo t 4 UF f o n o hydrogen fluoride recovery o effluent treatment Each unit proces s describei s d briefly below. UO3 Receiving/Feeding 1103 is shipped from Blind River to Port Hope in 10t tote bins. When receive t Pora d t Hope e coveth , r cape replacear s d y valves b s thei e totn Th .bi einverte d ovee receivinth r g screw conveyer and emptied by gravity. The screw conveyers fee a bucked t elevator, which convey e 110a primarth s 3 o t y feed surge hopper, throug a tramh p scree o removt n e lumps. Material passin e screeth g s a fedsurgi n a vi .e hoppera o t , pulveriser and thence via a bucket elevator to the final 1103 feed hopper.

UO3 to UO? Pulverised 1103 froe finath m l feed hoppe o s t screi r d fe w fluid bed whicn i se reductio th h 2 occursUO o t n . Hydroges i n used as the reducing gas. The UC>2 product is then conveyed throug a hcoolin g screa fee o dt w hopper.

UO2 to UF4 a screvi d w fe conveyer s i 2 UÛ s int a serieo f threo s e reaction tanks s slurriei 2 . UO Herde witth e h hydrofluoric aci o convert do UF4 t e conversiot .Th i t s carriei n t ou d continuously in a series of three tanks in a cascade process at about 90°C. Product slurry from the last tank is fed to a drum drier where the UF4 is dried, and the aqueous HF driven e vaporth d ofsan f scrubbe o recovet d r dilut F solutionH e . 4 Thireactio UF s returne i se th no t dtanks . Froe druth mm s calcinei 4 UF drieo remov t de e remaininth r th e g water.

6 UF o t 4 UF UF4 is transported to the fluorination process via screw conveyers and discharged into a bucket elevator. The elevator carries the UF4 to the eighth floor, where it is discharged ta tramo p scree o removt n 7 1/8e " material. After screening the UF4 falls to a surge bin. From here, it is fed via screw conveyers to both the flame reactor and the clean-up reactor surge bins. Flame reactord frofe m e scresar w feeders, which also maintain a seal against the flame reactor pressure. UF4 and fluorine (pre-heated to 450°C) are reacted at the UF4 powder

16 dispersers at the top of the reactor. Most of the reaction of UF4 with F2 occurs in the top 2m of the reactor, at 455-549°C. unde a positivr e pressur f somo e e 2kPa. Unreacted products and ash fall to the bottom of the reactors e collecteanar d n heatei d h cansas d . Thes e changear e s a d required. After cooling for several months the material is recycled to the flame reactors.

Product UF6. containing small amounts of UF4. HF and non-condensible gases and particulates leaves the reactor through primary sintered metal filter o secondart s y filters. Particulate matter froe filterth m s collectei s h cansas n .i d o colFilteret s thei d fe ntrapg UF d s s wherseparatei t i e d 2 whic¥ e clean-us d routei hth an fro o F t dH m p reactors. In the clean-up reactors excess UF4 is reacted with the gases £ colUF frode trapth ma manne n i s r similae th o that rf o t flame reactors, and operate at about 400°C. Products from the clean-up reactors (UF6 gas, unreacted UF4 and ash) are discharged by a screw conveyer to seal hoppers, where UFg, HF and non-condensibles is drawn through clean-up reactor primary and secondary filters, the gases going to the cold traps. Unreacted UF4 is then fed to the flame reactors. 5 collecteUF e primarth d n secondari dan y y cold traps i s heated and fed to the final cold traps, from where it goes to cylinder filling. Fluorine Production Fluorine is produced by the electrolysis of anhydrous HF in KHF n 12,00i 2 p cell0am s with individual rectifierse th n i , usual way. Fluorin s collectei e a surg n i d e s thei drum d n an ,compresse d through sintered monel demisters to the feed drum, and thence e flamtth o e reactors. Hydrogen produce e electrolysith n i d s a surggoe o t se drum, through demister d compressorsan s s i d an , watea n scrubbei r F H scrubber f o d e scrubbeTh . d hydrogen then passes throug a potassiuh m hydroxid o atmosphereet seat po l . Hydrogen Fluoride Recovery e majoth rf o reason e On r adoptinfo s e "wet-wayth g 4 UF " abilits procesit s o recovert ywa s d reus an . ThiHF et onl no sy reduced operating costs but also the amount of waste being generated. The main HF recovery system employs water scrubbing of HF-containing gaseous streams e generaTh . l schem s showi e n in Figur. 4 e

17 oo

, F A4MO , UO F WH

Polit»! Stil WH« Willi

DO, F,

KOM CM***«K l—— i

' 1 1 • i . t sJw«t« 1 HFV«o« faul! SCMJBMMKH 5 Q ^ PtöCrt!l Wtlff

HF MCYCL0 E1 AQUKUI NF , ^_ /UtwniMKovH i ' ^~ ^ «XWC N 1 ^ C«IH«H«

1 U^«OH 1 (KF 1 1, COJ

1' 1r" T ^

Fig. 4. HF Recovery System Effluent Treatment e overalTh l effluent treatment proces s dividei s d into four functional areas - o uranium recovery and equipment cleaning o fluorid d carbonatan e e removal o potassium hydroxide regeneration and storage o area fume removal and scrubbing These area e showar s n schematicall Figurn i y . 5 e 2 UO Conversioo t ^ UO f o n Uranium trioxide froe Blinth m d River refiner s dissolvei y n i d nitric aci o product d a uranye l nitrate solutio f abouo n t U (Figur L . g/ Thi 3) e0 s20 solutio precipitates i n d with anhydrous ammoni o product a n ammoniua e m diuranate (ADU) slurry. The ADU is separated by centrifugation» and the wet soli s driei d d usin a Weismong t drier. After drying, the JJDU is fed to rotary kilns where it is y hydrogeb reduce2 UO no t d(cracke d ammonia) e producTh . t s packagei U(>2 r shipmenfo d o fuet t l fabricator r fuefo s l pellet productio d encapsulatioan n n into fuel bundle r CANDfo s U reactors. e ammoniuTh m diuranat U este AD s evaporateproduce i pe th n i d d to 550-60 L ammoniu0g/ m nitrat d solan ed localla s a y fertiliser.

19 e

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20 RECENT ADVANCE PRESEND SAN T STATUS OF URANIUM REFINING IN INDIA

N. SWAMINATHAN, S.M. RAO, A.K. SRIDHARAN, M. SAMPATH Nuclear Fuel Complex, Hyderabad V. SURYANARAYANAN, V.K. KANSAL Bhabha Atomic Research Centre, Bombay India

Abstract e presenTh t requiremen f reactoto r grade natural uranium-di-oxid metad ean l for power and research reactors are met by indigenous production. The purity of the final produc achieves ti solveny db t extraction process usin i butyl-phosphategTr . This involves handlin largf go e quantitie slurrf so pulpingd yan , filtratiod nan evaporation operations. The calcination and reduction are carried out in rotary furnaces which are more amenable for smaller scale of operations. An extraction uni bees tha n develope uraniur dfo m extraction directly from slurries, thus eliminatin numbea g procesf ro s steps thanormalle ar t y involved wit conventionae hth l perforated plate pulsed columns. Developmental efforts were also directed towards the studies on recovery of Uranium by Ammonium-Uranyl- Carbonate (AUC) route & calcination of AUC & reduction of UCL (obtained by ADU routefluidizee th n ) i d reactiodbe n systems. This paper proces e bringth t ssou developments, which have taken placn ei recene Indith n tai past, that would mak refinine eth g operations much more simpler d adoptablar r largeefo r scal operationsf eo .

INTRODUCTION India, like other countrie worlde th followes n ,si ha conventionae dth l route for refining of Uranium to get the reactor grade Uranium-dl-Oxide and metal. The refining process essentially involves purification by selective extraction of uranium using tri-butyl-phosphate (TBP), precipitatio ammonium-di-uranats na d ean subsequent calcination and reduction to U0„ followed by stabilization under controlled condition handlee b enabl o so t t open di t ei n atmosphere U0~powdee Th . r thus obtained, is finally pulverized and blended to get a homogeneous product, free from agglomerates ard suitable for pelletizing. A flowsheet of the process, starting from yellow cake, is given in fig.i. For making the metal, the UQ~ is hydro- fLuorinated and the UF, is reduced with magnesium. extractior Fo uraniuf no m from crude uranyl nitrate solution, perforated platej pulsed columns are used which are operable with clear filtered solution, However, magnesia used at earlier stage of precipitation of yellow cake, has been leading to fine silica particulates being carried over to the filtered solution, This gives rise to some operating problems in the extraction columns. The silicious material accumulates at the interphase, resulting in entrainment in the extract phase. The pipe lines, sieve -plates, valves and even the column heads get deposited with silica resulting in frequent cleaning and washing operations. Further, the cake obtained after filtration of crude uranyl nitrate solution contains appreciable amount of soluble uranium. The cake has, therefore, to be pulped and filtered three to four times to recover uranium and for decontamination

21 MAGNESIUM DlURANATE CONCENTRATES 1 NITRIC ACID DISSOLUTION ^ URANIUM SOLUTION URANYL NITR/liTE SLURRY ^ AGING | ACT1V FILTRATION , ———CAKE ^REPULP.MG E CAKE - CAUSTIf C SODA CRUDE URANY L NITRATI E i » — —— — —— — - I —————— f ÏAFFINÂTE r pppr^TATinw ACTIV TBP DILUTE »_ D._ SOLVENT —— — ^ WA^Tt I E WITH KEROSENE I EXTRATION " DISPOSAL! I h | SODIUMT NITRA ' 1 r r PURE URANYL N/TRÄTE ,. AMMONIUM DlURANATE 1 AMMONIUM. NITRAT1 E ' — AQUEOUS AMMONIA-*- PRECIPITATION f ADU CAKE

DRYING AND CALCINATION AMMONIA U03 | T STABILISATION N2fH U02 | C*ACKfai 2 ,. REDUCTION " PULVERISATION D STORAGAN E

Fig.l. Production of nuclear grade uranium oxide.

of the cake to satisfactory levels- The leach liquor, thus obtained, is concentrated by evaporation and then taken for extraction along with the main stream. In view of the above problems, there was a long felt need to evolve a system where the slurry can directly be processed. A new extraction unit has been develope Procese th y db s Development Grou Nucleaf po r Fuel Complex (NFCr )fo extractio uraniuf o n m directly from such slurr whic] [1 y h wil discussee lb n di textdetaie th .n li Presently the pure uranyl nitrate solution is precipitated with ammonia as ammonium di-uranate precipitate Th . e thus obtaine amorphous i d naturn si d ean lead formatioo st agglomeratef no s durin furthes git r processing realises wa t I .d that if uranium is precipitated as ammonium uranyl carbonate the precipitate would crystalline b naturn ei e with better flow characteristic amenabld san adoptinr efo g fluidize operationd dsubsequenbe e th r fo s t stage processingf so . Development work carried out in this regard at Nuclear Fuel Complex is also described in this paper. As mentioned earlier, the calcination and reduction operations for the ajononium-diuranate are presently carried out in rotary furnaces. Efforts were also directed towards the development of fluidized bed reactor system in view of its obvious advantages for reduction of DO., to U0~. The work carried out in this aspect Chemicae atth l Engineering Divisio Bhabhf no a Atomic Research Centr alss ei o covered. CONIKACTDA EXTRACTINR RFO G URANIUM DIRECTLY FSOM SLURRIES The development of this contractor for extraction,directly from a slurry, was carried out in three steps. Experiments were carried out on a single stage unit first, the pilona t scale solvent extraction unit with seven stage designes swa d and operated for about one year (1). The salient features of the unit are : neithes ha movint y I ran controy g an par ) i r ltno slurre valvth n yei line,

22 lifr ai t pum uses ) inter pi il dfo r stage aqueous transfe alsd r ran ofo mixing the aqueous and organic phases for mass transfer, iii) organic flow takes plac naturally eb y available hydraulic gradient, iv) high organic/aqueous ratio is maintained to ensure aqueous phase dispersal and minimum entrainment loss of solvent in the raffinate. Experiments with slurry containin solid% hig15 s ga s hsa were conducten do the pilot scale unit with no operating problems, though the solid content in the crude slurry obtained after dissolution of the yellow cake is of much lower order. Based on these experiments, a plant scale unit capable of handling about 350 solvenf o aquouf o h designeds h Ip twa s Ip 0 slurr70 d , yan fabricate d commissioneddan . operatio n unii e Th s ti halmora r nfo d fe an yeartha e non s (2). Descriptio operatiod nan n schematie Th c diagranewle th yf mo develope d slurry extractoe r th uni d tan flow arrangement are given in fig.2. The plant scale unit also consists of seven stages. Each mixera stag s eha , disengagement sectio settlind nan g tanke Th . disengagement sections have been provided with mist eliminators with their exhaust connected to a common exhaust header. The slurry feed is metered and fed at the extract end while the raffinate is pumped out in metered quantity at the other end. aqueoue Th s strea transferres mi d from stag lif r stago eai t t y pumeb p from extract end to raffinate end, while the organic flows from raffinate end to extract end counter currently by gravity. The overall level difference in the 7 stage settling tanks is about 300 mm.

U.ECE.NJ) LUS/hou0 70 CTjrfA) > Q r £ ) qr(O

('hour ["'S 4700 Llrs/hour (O'g)

Fig.2. Schematic diagram of slurry extraction system (U.O.P)

23 DIUBANATE

NOTE_; A OPERATION;AJ - . E SHOWBLOCTH m ( KNCho m dotteE AR ) d ELIMINATED SY THE SINGLE OPERATION OF SLUSRV EXTRACTION

Fig.3. Solvent extraction of uranium.

Lean solven admittes ti e sevent th t a dh desiree stagth t ea d flow rate through a rotameter to get about 85 to 90% saturation of the solvent with respect uraniumo t . Slurry fee checkes di uraniur dfo m concentration, free aciditd yan percentag solif eo d befor admittes i slurr e t eth i o ydt extractor. Stage efficiency of the order of 85 to 907« has been achieved, though the contact time is of the order of a few seconds. 350 Iph of feed is admitted at the extract end which gets mixed with about 3500 Iph of recirculated solvent from the first stage. The mixed phase is pumped into the disengagement section by the first stage air lift pump. The air from the disengagement sectio ventes ni througt dou mise hth t eliminator mixee Ih .d phase, then,flows into the settling tank through a tangential entry. Organic gets separated settlerfroe mixee th m th f dflowo d phase.an t sp ou to , e riseth o st

24 From this, 700 Iph overflows as an extract while the rest (about 3500 Iph) gets recycle comed dan contacn si t witfrese hth h slurry. Similarly abou tcleaf o 420 h r0Ip organie th f o c p flowto frot e sou m th second stage settlin g suctioe secone tanth th y kb f dn o stag lifr eai t pumpt Ou . flowh Ip thif s0 o upwars70 d int firse oth t stage settling tank throug dowe hth n corner due to gravity. The down comer has been sized such that the upward velocity of the organic is low enough to permit the aqueous flow downwards. About organif o 350 p 0Ih c aqueouf alono h gIp s wit 0 froh35 m second stage settling tank thire pumpes i th do t dstag e disengagement sectio settline theth d nan o n t g tank and thus the flow continues. A clear cat interphase is allowed only in the last stage settling tank where the raffinate is withdrawn from the bottom by an airlift arrangement. The recycle rate of organic at each stage is very important for efficient operation of the systems. The organic recirculation flow is controlled such tha A ratit0/ always oi s more than 10: 1mixinn i g stages. Operating conditions remain stable once the flow pattern is set in each stage. Fig.3 gives the operations involved in the conventional process that are eliminated by the newly developed slurry extraction unit.

Advantages slurre Th directln yca ) extractee i yb d withou filtrationy tan , repeated pulping and evaporation of wash liquor; thus a huge saving on labour, handling operations and energy. ) ii Uraniu raffinate th n mi controlles ei permissible th o dt e o n leve d lan further treatmen necessarys ti . iii) The system can be operated over a wide range of plant capacity. iv) No moving parts are involved and so practically no maintenance and operation is more reliable. stoppee b unie n restarteTh d tca dan ) wilt v a d l without disturbine gth equillibrium. vi) As the number of operations are grossly reduced the plant is compact and neat. vii) Huge energy savin alsd an go savin spacn gi civid ean l work. viii bornr Ai ) e activit mise th t t yeliminatoa r exhaust duc wels ti l withie nth permissible limit.

AMgNIUM ORANYL CARBONATE (AUC) PROCES REOOVERR SFO URANIUF YO M FROM PURE URANYL NITRATE SOLUTION

In view of the excellent flowability of powder by carbonate route, its amenability to fluidized bed operations and single stage pelletization, an experimenta installes wa p u operated t dan l se studo dt procese yth s parameters. schematie Th c diagraprocese th f mo s steps involve shows di fig.n ni r 4fo precipitatio froC AU m f npuro e uranyl-nitrate.Ammonia, carbon di-oxid werr ai e d ean passed through the uranyl nitrate solution, which was also added continuously in tank,th x solutioe emi Th . stirres nwa alsd dan o continuously recirculated through a circulation pump mix-tane Th . provides kwa d with internal steam heating coils and temperatur abouf eo t 60° s 70° o maintaineCt wa s CH wa p tanke e th Th .n di controlled betwee 8.5d an . n8 Uranyl nitrate solution with uranium concentration gm/0 gm/0 20 o f10 1d 1weran slurrC e AU continuouslused s e ywa Th . y withdrawn from the recirculation stream filtered in pan filters and dried in tray driers.

25 S -OFGA F

UNH

CALCINATION REACTOR

AUC OFF-GAS H -»-N PRECIPATION 2 2

cm a

REDUCTION "l

REACTOR U02 PRODUCT

X3ÖO PREHEATEH

C rout AU r préparatioFigfo e. .4 f l>0_o n .

driee fluidizea Th powdeC di d AU m calcines m rreactorwa d 5 d7 be n di . s ga « N d an maintained s be wa C e ° th 0 n 65 di Temperatur o t 0 range 60 th f eo n e i velocity of 3 to 5 cm/sec, was used. The studies were conducted with powder feed rate from 5 to 16 kg/hr. The 'JO^ powder was subsequently reduced in rotary furnace, The UO~ , thus obtained, was observed to have better tap density and required only single stage palletization. The pellets were also of better finish and their acceptability was better. However, the process variables need further studies for large scale operation powde „ obtaio UÖ t acceptablf e d ro ns th an consistend ean t metallurgical characteristics suitable for sintering.

FLUIDIZED BED REDUCTION OF URANIUM TRI-OXIDE TO URANIUM PI-OXIDE

Rotary furnace is presently being used for reduction of uranium trioxide to uraniu mhydrofluorinatior oxidfo d ean dioxide e fluidizee th th f no s A .d dbe reactor offers a more efficient and compact system, studies were conducted to evaluate various operating parameter asseso t feasibilite d sth san s it f yo replacin presentle gth y employed rotary furnac reductior efo n operation (and subsequently for hydrofluorination as well). The Uranium tri -oxide powder (UOo) used was composed of extremely fine particles, the average particle size being about 19 /u. The loose packing density was 2.2 gm/cc and the tap density was 2.6 gm/cc approximately. The composition of the powder varied somewhat with different feed stocks used for different experimental runs (may be because of variations in ADU - calcination conditions). A powder composition with 0/U ratio of about 2.75 could be taken as the average. A 75 mm diameter glass column was used for determining the fluidization characteristic UO.e ,th f powderso date Th .a thus obtained, forme basie dth s for the selection of operating gas velocity range and bed height for reduction reactor sintereA . d porous meta mealu / disn0 porcs (2 ega sizee useth s )s wa da distributo thesr rfo e studie subsequend san t reduction trials.

Description and Operations (3) The reduction reaction with hydrogen is exothermic and the rate of reaction becomes significant onl temperaturet ya s about 350°C and, therefore finae ,th l runs

26 were conducte operatine th n di g temperature rang ASu^f ec 600°o Ct C where eth reaction rat quits ei e higmasd han s transfer become controlline sth g factor. Fig.5 show schematie sth c diagraexperimentae th f mo employep u t lse r dfo fluidizee th reductiod dbe n N.Bm trialsm 5 .reactoe 7 Th a pip. s erwa 0 wit15 ha mm disengagement section at the top. The l£U powder was admitted in the bed above distributoe th r whil produce eth t overflowd dischargbe y b reactoe s Th .ewa s rwa provided with suitable heating and temperature control arrangement. The IX^roduct receive cylindricaa s rwa l vessel equipped with porous metal filters feee .Th d gas, a mixture of (-L and N~ in the ratio 3 : 1 was preheated to about 300 to 350°C before it was admitted to the reactor. Once the steady state operation had been achieved a few pouder samples were withdrawn from the product overflow line for determination of 0/U ratio such that they are not exposed to the atmospheric air. In certain trial runs no on-line samples powde ~ werU0 ee r drawth produce d nan directls dwa y hydrofluorinaten di rotary furnace reactor amoune Th . UO~F f to produc, ~UF formee takes tth wa s n a di a measurconversioe th f eo n efficienc fluidizee tn n yi réductiod dbe n reactor. Table 1 shows a summary of a few of the trial runs conducted. Samples for the first three runs were analyse onliny db e sampling determinationU (0/ ) while eth product for the last two runs was directly hydrofluorinated and analysed for l3QjF~ content. e resultTh s obtained from these preliminary trial runs encouraging e loob o kt . However, more studies are required before the process can be adopted for production operation.

OFF-GAS

FEED POWDER

MOTOR

W\AA/VV\AA/vH BUNKER

FF GAS

Fig.5. FLuidized bed reduction set-up.

27 Tabll e-

AveragU eIX Total gas Feed gas Feed gas Product Hydrogen Bed-Temp. composition excess Run No. feed rate Feed rate H»/N„molar velocity <°C) (kg/hr.) (1/m) ratio (cm/sec) (Wt U02) 7.

1 8 28 3:1 5.6 430-460 86.8 71.4 2 10 35 3:1 7.0 450-600 84 73.1 3 10 32.5 3:1 6.3 550-600 99.5 61.5 4 6 25.5 3:1 5.1 550-600 99.5 106 5 8 28 3:1 5.6 500-600 99.5 71.4

Acknowledgement

authore Th s convey their sincere thank Shro st i R.K. Garg, Director, Chemical Engineering Group, BARShrd BalaramamoorthyC. an iK , Chief Executive their fo rC ,NF keen interes constand tan t encouragemen thesr tfo e studies.

References

1. "A new solvent extractor for slurries" Swaminathab. yN S.Md nan . Rao, Nuclear Fuel Complex, Hyderabad to be published 2. "Solvent extraction of Uranium from Uranyl nitrate slurry" by S.M. Rao, A.K. Sridharan, M. Sampath and N. Swaminathan, Nuclear Fuel Complex, Hyderabad

3. "Fluidize reductiod dbe U0f no 3" Suryanarayanab. yV BARA n- C internal report

28 OPERATING EXPERIENC PILOA F EO T PLANT FOR THE PRODUCTION OF URANIUM DIOXIDE FROM URANIUM ORE CONCENTRATE

. SHABBIM R Pakistan Atomic Energy Commission, Islamabad, Pakistan

Abstract

Pakistan's heavy-water moderated power reactor (CANDU TYP) E s fuellei d with natural uranium dioxide. Chemically pure, compact-

able and sinterable uranium dioxide powder is required to suit the production of reactor grade fuel pellets.

A pilot plant for the production of natural uranium dioxide from the indigenous uranium ore concentrate to meet the CANDU fuel specifications has been established. This paper concerns with the operating experience e piloth tf so plant.

1. INTRODUCTION

Uranium dioxide (U^) powder for the production of fuel

pellets with density close to the theoretical, is produced from

the ammonium diuranate or ammonium uranyl carbonate. In the early

studies refinin f yellogo w cak solveny b e t extractioe th d an n

production of small quantities of UÛ2 for laboratory studies

from ammonium diuranat s reporteewa d [1]. Since thestudiee th n s

were further extende o product d e natural uranium dioxide, suitable e fuefoth rl fabricatio r Karachfo n i Nuclear Power Plant (KANUPP).

Various studies have been reported on the influence of precipitation, calcination and reduction conditions on the sinter- abilit f 110o v 2 powder [2,3], These unit operations influence th e physical propertie 2 nowdeU0 f so r (i.e. surface area, bulk densitv,

29 tap densit d particlan y e size). These physical properties affect the sintered densit2 fueU0 lf yo pellets .

The physical properties of IJ02 powder have been controlled bv optimisation of the process and operations parameters at each stage. Equally importan e chemicath s i t l 2 puritpowdeU0 i f o yfo r neutron economy.

In the present studies the parameters for the extraction of uraniu usiny mb g TBP-Kerosene mixture, subsequent precipitation, calcination and reduction processes have been optimised. (JO2 powder produceo s bees e reactoha dn th user rfo d grade pellet production.

2. MATERIAL SPECIFICATIONS Specifications of the uranium dioxide powder include requiremen chemicaf to l purit d sinterabilityan y e chemica.Th l puritf o y U02 required for the KANUPP fuel is given in Table-I. Impurities like boron, cadmium and gadolinium are highly deleterious due to their higher neutron absorption cross section areas i.e. 755; 2,550 and 46,000 barns respectively. Equivalent boron content (EBC) of U02 powder impurities is determined and the total EBC shall not be >1.2.

Anionic impurities e.g. carbon and fluorine are also undes- irabl havd an ee marked e effecfabricatioth n i t f fueno l pellets d theian r subsequent irradiation.

30 TABLE- e PermissiblTh I e Impurities 2 LevePowdeU0 n r i l fo r KANUPP Fuel Fabrication.

Impurity Max. Leve ppirn i l basiU , s M 25 B 0.3 C 200 Ca 50 Cd 0. 2 Cr 15 C.- IV

Thoug physicae th h l propertie powderf w o sno e ar s well founde d theian d r maximu minimud an m m limits have been laid down but their requirement varies with the fuel manufacturer. Nevertheles acceptabilite th s f o y powde entirels i r y"Sinterine baseth n o d g Performance Test" powdee .Th r characterizatio "performance th d nan e test" are given in Table-II.

31 . CharacterizatioA n

TABLE-II. Natural U02 Powder Characterization and Performance Test.

Parameter Required range 1. Surface area m2/g 5.0- 7.5 2. Moisture (Wt.l) 0.4 max. U ratio-correcte30/ . r fo d 2.1- 2.047 moisture 4. Bulk density (g/cm3) 0.75 - 1.25 p densityTa . 5 ' (g/cm3) 1.7- 2.155 6. Fisher Particle sizef^n) 0.8 - 1.2 3 7.Qrsen density (e/cm ) a) inside die at 2?6 MN/nv5 5.6 - 6.0 b) out of die at 276 MN/m3 5.1 - 5.4 8. Compressibility factot ra 276 MN/m3 1.8 -2.2 9.Biggest particle size CD-80 0.015" max.

B. Performance Test A total of 10 test pellets shall be cold pressed at 276 MN/m2 (40,000 Psi) wit dwelha lsecond0 3 tim f o e s maximu pressurt ma e

without addition of any binder or lubricant to a density of

3 4 g/cm5. 5. - 3 1wit variatioa h t morno e f o ntha n ±0.05 g/cm from the average.

The pellets shall be fired in an atmosphere of H? of dissociated NH, for two hrs. at 1650 ± 2S°C. The temperature profile

shall have a gradient of :00°-300°C. The pellets shall give a

sinter density of more than 10.53 g/m 3 (96% TDS). Average sintered density variatic.' .more b siu'i et no thal 0.0± n 5 g/cm.3

3. PROCESS DESCRIPTION 3.1 Dissolution Stainless steel dissolvers fitted with paddle type agitators are used to dissolve yellow cake (Y.C.) and uranium dioxide pellets pellete .Th s before dissolutio preheatee nar tempa t da -

32 erature of 350-400°C for about 12 hours. Heating and cooling jackets have been provide r steafo d m heatin cold an gd water circulation dissolvee .Th d uraniu fore uranyf th mo n i m l nitrate solutio filteres ni d throug platha d framan e e filter press. The feed solution i.e. crude uranyl nitrate prepared for the extraction step contains 300 ± 10 g/1 of uranium and 3±0. freN 1 e acidity.

3.2 Solvent Extraction batterA columnf yo s have been useextractionr fo d , scrubbing d strippingan pulsatior Fo .columne th f o n, oulsating pumps variablf o e stroke lengt d frequenchan y have been usede .Th solution d throug colume fe th e o hnar st proportionating feed dosing pumps. The operational parameters are as follows : a) Extraction Parameters - Solvent composition 301 TBP + 70% Kerosene oil - Flow rate of solvent to 3:1 crude uranyl nitrate - Pulse amplitude 14 - cm c/mi2 - Puls1 ne frequency Scrubbin) b g Parameters - Flow rate of uranium orga- 12:1 approx nie to 4 N HNO-7 solution. m c 1 - Puls1 e amplitude - Pulse frequency 12 c/min.

c) Re-extraction(Stripping) Parameters 1 - Flo: w1 rat f U-organieo c complex to BMW - Pulse amplitude 13-1. cm 4 - Pulse frequency 12 c/min. Refined uranyl nitrate (UNH) solutioo t obtaineo s nd fe s i d the precipitation productioe vesselth r sfo ammoniuf no m diuranate (ADU).

33 3.3 Precipitation Phvsical propertie ~ PowdeU0 dependene f so rar t upoe nth precipitation conditions. The influence of precipitation parameter discusses si section i d . n4 Process/operating parameters have been studied and optimsed; the values are given as below:-

- Precipitation temperature 50°C - Percent uranium precipitation 50±5 (piI 3.5) 1st stage - Percent uranium precipitation 50±5 (pH 7.2) 2nd stage e thickneth o t d subsequentl ran d D slurr fe AD s e i yTh y filtered usirig rotary drum vaccum fliters(RDVF). ADU is then transferred to the repulper, slurry so obtained is again filtere y DRVFb d U cak dries .AD i e t 120°a d d Can ground to (-) 10 mesh. 3.4 Conversion of ADU to UC>3 .ADU is fed to "hree zone furnace. Each zone is controlled by independent neat ing circuit. The temperature of first zone is kept at 260 °C, second zone at 270 °C and third zone at 280 °C. The residence time of powder is kept one hour.

3.5 Calcination of U03 to U30g Orange oxide (U03) is conveted to UsOg. The calcination temper

Thi. °C s0 temperatur55 - 0 atur53 s ei gives U700 which flows J O smoothly when fed to the reduction furnace.

3.6 Reduction of u^Og to U02

rotara o t yd typfe ILO s ei g furnace wit provisioha supplo t n y dissociated ammonia (N2 + 3H2) mixture) stream. The powder is reduced at a temperature of 625 - 650°C.

34 The powder (UCU) so obtained is hihgly pyrophoric and has to be stabilised. Stabilization is done by the addition of dry ice into UC>2 Powde Partia. ] [4 r l pressur oxygef en o inera n i nt atmosphere of solid/gaseous carbon dioxide leads to the formation

of UJDr, ? monolayePowderU0 e powdee th . Th n ro r once stabilised is very stable towards oxidation. The reduction temperature has a direct bearinphysicae th n go l propertieU PowderUC e th . f o s This aspect is discussed in section 4. 4. DISCUSSION 4.1 Extractio Uraniuf no m In the solvent extraction section the free acidity of crude uranyl nitrate solution has marked influence on the extraction efficiency extractioe .Th n efficiency decreases witdecrease hth e of free acidity. The results of studies under pilot plant conditions have bee ns bee giveha n t n FigI i nobserve . 1 . d thae th t free acidit resultN uptake 3 th > ydeleteriouf n o ei s s impurities into the organic complex e.g. boron, cadmium, gadolinium etc. [5].

The uranium concentration of feed solution also affects the extraction efficiency e optimu.Th m uranium contenfeee th dn i tsolutio s i n 0 g/130 .e foun b Furthe o t d r increas uraniuf eo feen i m d solution lead increaso st e uraniu raffinaten mi . The throughput has also a bearing on the extraction efficiency. Throughput will have to be increased, keeping the capacity of the colum viewn ni . Such thaextractioe th t n efficinec maintaines i y d an d the ratio of organic to acqueous flows is such that the uranium is in stochiometric balance organie .Th aqueouo ct s ratio found optimum . e operatio1 th : r 3 fo s i n

35 J 98.0 -

c 9M -

96.0 -

9S.O 2.5 3.0 3.5

Free Acidity —

FIGURE Percent Extraction Efficiency Versus Free Acidity.

Scrubbing of the uranium organic complex is essential for the removal of left over impurities. The strength of nitric acid had to be kept more than 3 N to prevent transfer of uranium into aqueous phase. Nevertheless the practice of scrubbing with UNH solution is preferred [6]. Thi advantageous i s furthen i s r increasin uraniue th g m concentration in the loaded organic. Increasing the uranium saturation e solvenoth f t result reducinn i s g co-extratio f impuritieno s [7].The main disadvantag lose th f refine so s i eH solutio UN n di d nan handling of aqueous obtained from the scrubbing column. For scrubbing 4N nitric acid solution is being used. Re-extraction(stripping) of uranium has been effectively accomplished

in the columns by keeping the U-organic complex and DMW. It has been observed that the increase in temperature to 60CC enhances the efficiency of stripping [8].

36 4.2 Precipitation of ADU from UNH Precipitation step for the production of ADU has a bearing on the physical properties of uranium dioxide produced for the nuclear fuels.

U crystallite e AD siz Th f eo d agglomeratean s> s decreases with increasin . PrecipitatiogpH verf giveo 7 > yU H finsAD p t ena particle size, which poses difficulty in settling and filteration. e settlinTh g rat n seconi e d stagH 7.1±0.p t ea gives i 1 Fig.2n i n . The settling rates are an index to sintered density. Lower the settling rate of ADU slurry higher the sintered density of UO^

prepare dreduction o froThi . 2 U giveit mU0 sAD f so n powder which is difficul proceso t t sintesbu higo rt h sinter densities.

60 -

55 -

! I I i i 1 I I j [

0 W 0.2 03 at as 06 a? 08 09 ID U 12 U

Settling Rate (mm/sec) —————•-

FIGURE 2 : Settling Rate of ADU in 2nd Stage at pH 7ltQli50'C

37 othee th rn O han precipitateU dAD H i.e p give5 w 3. .lo s t da course particle agglomeratesd san powdeL UC e r.Th produced from less i thi sU pyrophorisAD easild can y compactabl givet ew bu lo s sintered densities(Table-III). Such a powder is not suited for fuel pellet fabrication as such. This type of powder has to be ball milled/micronise goot ge d o resultsdt .

TABLE-III. Effec PERCENf to T Precipitatio Physicae Stagt th 1s n eo t na l Characteristics of UCU Powders and Sintered Densities.

Uranium Physical Characteristicf so Sintered Percent U02** Powders Density Precipitation* g/cm3 Bulk Density Tap Density Surface Area g/cm3 g/cm3 mVg 50 1.1- 5 1.25 1.85-2.10 4.5 - 7.0 10,5-10.6 60 1.25 - 1.30 2.00-2.15 3.5 - 5.0 10.3-10.4 70 1.30 - 1.40 2.10-2.30 4. 0- 0 3. 10.2-10.3

* Precipitation Temperature 50°C ** Reduction Temperature 625°-650°C

To achieve a balance, two stage precipitation is preferred [2]. This gives a blend of coarse and fine crystallites that filter easily e secon.Th d stage precipitatio ±0.1 7. H 1p f no also ensures that all the uranium is recovered from the solution.

? powdeU0 r obtaine giveU d AD frose mintermediatth e surface L* ared otheaan r physical properties produceo s t highl L no .UC s >i d v pyropharic but needs stabilisation and sinters to very high sintered densities e.g. greater than 96% of theoretical.

38 4.3 Conversion of ADU to U02

This step has been carried out in three stages :-

) Conversioa o t U AD f no b) Calcination of U03 to

) Reductioc 2 U0 Uf o 3no 0t 8 ensures Ii t d that each materia completels i b sten & i l pa y converted. The process parameters are adjusted to get smooth flowing materia electricae th n i l l resistance furnaces. Care is taken to avoid micro and macro-sintering in steps(a) and (b). Such pre-sintering would tend to reduce the surface area finae th lf o product i.e. UCU e temperatur.Th calcinef eo alss i r o very important facto determininn i r physicae th g l propertiee th f o s

U02 Powder.

The most important step is the reduction of l^Og to U09. This step is carried out in the reducing environment of either hydrogen

or dissociated ammonia. The surface area of U07 powder independent governes parene i reductioe th U th f to AD y db n temperature. Higher reduction temperature results in lowering of surface area, whereas, lower reduction temperature increases it as shown in Table-lV. On

TABLE-IV. Effect of Reduction Temperature on Surface u PowdeAreUC f ao r

Reduction Temperature Surfac Aree a 2 Cm /g)

625 6.0 - 7.0 630 5.0 - 6.0 640 3.5 - 4.5 650 3.0 - 4.0

39 the other hand, high temperature of calcination and reduction will increas densitiesp eta buld kan , whereas temperaturw ,lo e will decrease their values. High reduction temperatures result presinterinn i crystallitee th f go s presenU agglomeratesUC n i t . It has been observed that powder reduced at high temperature i.e. > 650°C manifests marked decrease in surface area, this supports the view point of gross pre-sintering of the fines taking place in the powder.

4.4 Influence of Physical Parameters on the Sintering Characteristic Powder^ UC f so s The powders are characterised generally for the following: Surfac) a e area b) Bulk density c) Tap density Particl) d e size Compactabilit) e y factor

The range of these physical parameters is given in Table-II. The acceptability of powder is not only based on the physical characteristic performancs it n o t sinterinn sbu eo g (performance test mentiones )a Table-IIn i d .

In general powders with high surface area bulw d ,lo k an tap densities and small particle size sinter to high densities. But there are process and economic constraints e.g. powders with high surface area are difficult to compact and handle. Powders with hig densitiep hta buld an k s sinteo t r low sintered densities but are fairly good in compaction. Powders with particle size >1.2pn ten yielo t d d pelletf so low sinter-density, such powders would need ball milling/

40 micronising for use in pellet fabrication. Thus the cost for the entire pellet fabricatio obviousls ni polydimensionaya l function. It is therefore necessary not to optimise the proce-

ss parameters at each step of U02 processing but to corelate the monotonous function of the cost as well [9]. The specifications are thus based on a compromise iiy process contraints, product requirement as well as economic considerations.

5. CONCLUSION By using indigenous yellow cake and recycled material,uranium dioxide to meet the KANUPP requirements has been successfully produced observatione .Th s durin operatione gth s conformed wit e informatiohth n already availabl n litratur. i e 9] - 1 [ e

- 6 ACKNOWLEDGEMENT

e authoTh thankfus i r N.K.Qazio t l , M.T. Qureshd ian N.A.Chughta experimentae th r ifo d technicalan l word kan M.Salim for the valuable suggestions and discussions for the preparation of the manuscript.

REFERENCES

1. M. Yunus, A. Muzaffar, M.T. Oureshi, N.K. Qazi, J.R. Khan, N.A. Chughtai and S.M.H. Zaidi, "Production of Yellow Cake and Uranium Fluorides", Proc. Adv. Group Meeting, Paris 5-8 June, 1979 (P-329). 2. J. Janov, P.G. Alfredson and Y.K. Vilkaitis, "The Influence Precipitatiof o n Condition Propertiee th n so Ammoniuf so m Diuranate and Uranium Dioxide Powders", AAEC/E220, May, 1971. 3. P.G. Alfredson and J. Janov, "Investigation of Batch-Tray Calcination-Reductio Ammoniuf no m Diuranat Uraniuo et m Dioxide", AAEC/TM 599, August, 1971.

41 4. W.T. Bourns and L.C. Watson, AECL 1312, 1961. 5. A. Muzaffar, M.T. Qureshi, N.K. Qazi, J.R. Khan, N.A. Chughtai d S.M.Han . Zaidi, "Productio Nucleaf no r Grade UC>2 Powder", Internal Report, NMD, PINSTECH, 1978. . J.E6 . Littlechild, "Operational Developmen Uraniua f to e mOr Solution Extraction Plant" Chem. ,I Symp. .E . Series, 26 . ,No 107-110, 1967. 7. P.G. Alfredson, "Productio Yellof no w CakUraniud ean m Fluorides", Proc. Adv. Group Meeting, Paris 5-8 June, 1979 (P-149). 8. P.G. Alfredson, E.G. Charlton, R.K. RyaV.Kd nan . Vilkaitis, "Pilot Plant Development of Processes for the Production of Purified Uranyl Nitrate Solutions", AAEC/E 344, January, 1975.

9. U.Runfors, "The Influence of Powder Characteristics on Process

and Product Parameters in U02 Pelletization", AE-415, April, 1971 (Sewden).

42 PURIFICATION AND CONVERSION OF URANIUM FROM IROTHORIUD NAN M CONTAINING DEPOSITS

. MOVASEGHIH . MEISSAMN , I Atomic Energy Organizatio f Iranno , Teheran, Islamic Republic of Iran

Abstract Acid leachin f o guraniu m deposits i s not a selective process. Sulfuric acid solubilizes half or more of the thorium depending on the mineralogy of this element uraniun T . m recover y solvenb y t extraction process, uranium is separated from thorium by an organic phase consisting of 10 vol% tributylphosphate (TBP) in kerosin diluent. Provided thae aaueouth t s phase is saturated with ammoniu solutioe th mf o nitratnH p d an e

is lowered to 0.5 with sulfuric acid. In other words the separation o^ uranium and thorium depends on the way that the relative distributions of the two materials between aqueous solutions and TBP vary with sulfuric acid concentration. Under these conditions the extractio f iron(IIIo n ) intP drasticallTB o y diminishes to a tolerable level. Uranium can be stripped from the organic phase by distilled water,denitrated and delivere r electrolytifo d c reduction. Uraniue b n ca m precipitated as uranium tetrafluoride by the reaction between uranous and hydrofluoric acid. Thorium is

later recovered froe wastth m e leach liqour after removal of sulfate ions.

43 INTRODUCTION

n uraniuI m purification with tributylphosphate from nitric acid solutions, ferric ion and thorium can accompany uranium ( VI ) to the organic phase in significant amounts and may appear as major impurities. The high iron contamination in uranium concentrate is undesirable, because it interferes with enrichment of uranium. The allowable amount f thoriuo n uraniui m m concentrate according with e specificationth s establishe y differenb d t conversion plant s showi s. ) tabln i nI ( e

Table ( I )

The product quality of the salable yellow cake according with the specifications established by the conversion plants . ) 1 (

British Conurhex Eldorado Allied Kerr nuclear chemicals Me Gee fuel (UK ) ( FRANCE) (CANADA) (USA) (USA)

U 40 60 50 75 60

n.a n.a 2.0 2.0 Th

0.15

) limit Î ( s with penaltie r exceedinfo s g values.

An enormous amoun f woro s tgon ha k e inte th o

development of extractive methods for uranium to meet the problems which have arizen in the

44 extraction of uranium from ores, the purification f e uraniurecovero th d an muraniuf o y m from reprocessing stage. Solvent s extractioha P TB y b n long been applied as a means of separating numerous elements from uranium in nitrate solutions. In this method uraniu s effectiveli m y separated from thorium and iron employing TBP in a proper extraction conditions.

EXPERIMENTAL

w materia ra e sulfuri th e s Th i l c acid leach

solution from thorium and iron containing deposits. The deposits lack the presence of considerable amount manf o sy spesific elements suc s molybdea h - num, titanium, vanadiu d boronan m . Leach solution is saturated withe ammoniuth f o H m? nitrate th d an e e additiosolutioth y b f s sulfulowere5 o i n 0. o -t d ric acid. Uranium is fractionated away from thor-

ium and iron by 10 vol% TBP in kerosin diluent. Thoriue recovereb n ca m d froe wastth m e leach liquor, öfter remova f sulfato l e ion y bariub s m nitratd an e using 30 voll TBP/diluent. Uranium is stripped from e organith c phas y distilleb e d wate d deliverean r d for electrolytic reduction.

The solution of uranyl nitrate is denitrated by means of sulfuric acid, crystals of uranyl sulfate,

UO,. (SO. 3H,), obtaineds 0i A solutio .0 gU/ 10 1f o n 2. 4 0 2. Cf 4 % hydrofuori35 d an c aci s reducei d d electrolyticall ia singln e compartment cell heate bata n hi d over

45 e maskinTh . C g 0 effec9 f o tfluorid n decreaseio e s reoxidation rate of U (IV ) on the anode. Graphite is used as anode and rrionel-400 as cathode. current 2 densit s aboui y5 Am/c d potentia0. tan m 7 volts6- l . e solutioTh s stirrei n d vigorously e currenTh . t efficienc e recoverth d f an uranius abouo y i y % 20 t m is more than 95%.

DISCUSSION The fractionation of uranium ( VI ) depends on the concentration of uncomplexed TBP in the organic phase. Thus one of the most important variables is the degree of solvent saturation with uranium whic mors i h e strongly complexe P thaTB ny b diro n and thorium. Thorium is successfully masked by the addition of sulfuric acid. Using no membrane and complex cell causes decrease in current efficiency e anodTh renewes .i e d after each electrolysis in order to prevent crumbling of graphite. Stirrin s continue i ge hou on rr fo d after the current is stoped. Very filtrable and

washable particle f UFo s 4, 0.75H,,O precipitated and studied by X-ray diffraction spectra. Hydrated uranium tetrafluoride grains grow to 70-100 micro meter d kepan s t suspendee th n i d electrolytic celle hydrateTh . d uranium tetrafluoride produced contains no significant impurities of Al, B, Ca, Fe, Th, Mo, V and Ti because the traces of these element e furthesar r remove t uraniua d m tetrafluoride precipitation stage.

46 References

[l] Morrison, G.H, Fraiser. H, Solvent Extraction in Analytical Chemistry, Wiley, New York 1965, p.88.

[2] Chiang, P.T. patenS U , t Doc. 4/255/392/A,

Int. C122B 60/02, 1981.

[3] Marshall . Nuclea,W r Power Technology, vol.2, Claredon press, Oxford 1983,P.398.

] [4 Peacefull use f atomiso c energy UN., CN ,,

vol 4, 1985, P/534.

[5] Takenaka . Kawate,S Uraniu, H. , e Or m Processing, IAEA, Vienna, 1976.

47 DEVELOPMEN TECHNOLOGA F TO MAKO YT 2 EUO STARTING FROM "YELLOWCAKE" REFINED WITH AMINE SULPHURIA N SI C ENVIRONMENT

J.A. VERCELLONE Atomic Energy Commission, Cordoba, Argentina

Abstract

The development carried out at a pilot scale certifies the advantag f purifyino e g "yellow cake y tertiarb " y o aminet d an s obtain AUTC nuclear purity through direct precipitating elution

Furthermore, this product being adequately conditioned in its mother liquo s wela rs straightla l y reduce a batc n i dh furnace by dissociated ammonia produces an UÜ2 free flowing meeting the standard specifications required for nuclear fuel used at Atucha I.

e seconTh d stag f thio e f sbeino technology gwa optimizedn i s i y , although sintered densitie f 10,4o s d 10,60an 0 g/cm e obtainear 3 d e stilar le w having problems wite graith h n growine th n i g pellets which we are at the present time trying to improve it by adjustin e sinterinth g g heat treatment.

e wanW o makt e acquainted that starting from sulphuric solutions of ore treatment it is possible, with the same technology and without intermediate precipitation n AUTa t C ge ,nuclea r purity of same characteristics as the one obtained through the former methodology.

49 F ABRI CAB I L I T Y OF AUTC (Ammonium Uranyl Tricarbonate) AND

NUCLEAR PURITY STARTING FROM ADU TREATMENT IN A SULPHURIC

ENVIRONMENT TERTIARY AMINES AND STRAIGHT PRECIPITATING ELUTION.

Operative Technique

U produce e ConcentratioAD th e t Th a d n Plants wit a standarh d analysis is hereunder shown:

U-jOg according to dry sample 85,74 %

= S04 1,76 V205 < 0,05

3 C0 0,3 8 0, 7 P2°5 Fe 1,18 Zr < 0,1

Si02 1,28 < 0,03 o M

e sam Th s delayei e n watei dd thean r n attacked with concentrated sulphuric solutio o nobtait (S04H2s a a finan o s )l, solution a d an 3 1, o f Uraniut o 2 1 1, g/ f abou mo f 0 o havin 15 tH p a g F.D.R. Powder Oxyreducer highe . FrormV mtha0 thi40 n s mother liquor which is previously filtered through a vacuum filter, an aliquot part is taken and diluted in water and raffinate coming froe Solventh m t Plant. This concentratio hers i n e adjusted to 20 g/1 of uranium and its pH reaches the value of 1,3. Being the same arranged and filtered through a line filter it meets the requirements to enter in the Process of Extraction through Solvents.

) Extraction_throu_2b h Solvents Process A five stages battery of mixers-settlers of the compact type contacte impurth w es no Uranyl Sulphate solution wit a tertiarh y amines solution (Alamin r Adogeo 6 33 ne 364M wite 1 )th h0,

50 aggregat f Isodecanoo e o n Volumdoins i e passag % d th 3 glan e e of uranium froe aqueouth m se organiphasth o t ec phass i e accomplished.

Amines solutio n thii y enriche wa s i nwitU n a concentrahi d - e raffinatth tio, f 1 abouo ng/ e7 twhic h come t froe sou th m Plant after passing throug a safeth y decante n uraniua s ha rm concentratio m (lowepp 0 1 r < than part te r nmillion) pe s ; the 50-60 % of same is recycled in order to condition the liquids in stage a) . What is left runs to the Effluents area e theb o nt neutralized, concentrate d recuperatean d s crystaa d l ized ammonium sulphatO S e

e solutioTh f amineo n s charged with uraniu s alsi m o washen i d mixers-settlers performe o stagetw n f reversi o sd e currenf o t 5 solutio1, = SO4U0H p n whicN P 2 s thei h n being sen o décantt t a tion. Once this has been completed its joining to the following stage is produced.

[2 R N + [H S0 ] 3g ^oror q 2 4 ^a^q

2(R3NH)2 S04]org + [U0< 2(S0 g _42SO+ Qr ) ] 3 ]=43 ) ^ 4 [(RS0 ( 3 NH)2 UO 4

[(R3NH)4 U02(S0 5 [ (NH4 )+ 34J2 ) 2 U0 [(NHT CC> 4 H ) 3] 4 etc.

e presenTh t operatio s performei n batcn i d s followsa h :

A measured quantity of Uranium amine saturated solution is first added and that is maintained to a temperature of 45°C, then a measured quantity of ammonium carbonate solution and

51 AUTC analysis produced at CFC

Elements

Ag 0,05-0,5

5 2 < l A ter 2 0, < B

0 10 < C

Ca 25

1 0, < Cd

Co < 1

Cr 10 - 25

Cu 1-3

Dy < 0,0 3

Eu < 0 , 0 3

Fe < 10 - 50

Gd 0,03 In < 1

Mg 3-10

Mn < 2

4 < o M

Ni < 4 - 20

Pb < 5

Si < 30

Ti 7-15

0 1 < V

Zn < 20

S 30-50

B < 0,2

52 ammonium sulphate saturated in Uranium are added having these determined concentration d bein an swels a g l heate o 45°Ct d . e herb eo t controllel s Al ha . e valu9 H th p f o o e t d Th these are stirred for a time of 30 minutes letting it then drawn off for 30 minutes to favor good crystallization.

d) Crystals arrangements

With the 70% of the mother liquor previously separated from the organic solutio a specia n i n l decanter dispositiof o n crystal s achievedi s A .stirrin f thio g s precipitation during 0 minutea 12 tun f o es obtaining o doings n i , , wort edgeou n s preparin e structurth g e futurth f o e UC>2 with free flowing characteristics.

Filtering_and_washing_gf_AyT) e C

Filtering is put through a centrifuge machine and the washing o stagetw n ii ss performed e firsth n tI . stag e materiath e l is washed wit% ammoniu20 h m bicarbonate solution saturated first in uranium and then with methanol that in addition to y organimovinan t cou g solution traces that migh e lefb t t kept back favors crystal deshydration, this allows the AUTC to hav moistura e t higheno e . r% tha5 0, n

53 SPECTROGRAPHIC ANALYSIS OF AN UO2 SAMPLE.

CFC POWDE 2 PLANU0 - RT NATIONAL TECHNOLOGY s accomplisheIwa t e Chemistrth y b d y Departmen- t Management of Chemistry Processes and CFC Laboratory

Elements Specif Lot 1 Lot 2 Lot 3 Lot 4 RBU 6 32 M M 333 M 337 M 343

Si 0 pg/g10 U = 50 30-40 < 30 D = 30

Cd 1D N 1 0, < < 0,1 ND <0,1 ND < 0,1 ND

B 0,2 < 0,1 D < 0,1 D <0D ,1 < 0,1 D

Ça 100 10-25 < 10 D <10 D < 10 D

Ag 2 < 0,05 D < 0,05 D <0,0D 5 < 0,05 D

Fe 100 15-50 15-25 15-25 25 Mn 50 2 D D 2 < D 2 < < 2 D

Cr 200 < 10 D D 0 1 < <10 D < 10 D

Ni 50 < 4 D < 4 D < 4 D < 4 D

Al 50 5-15 D 5 < D 5 < < 5 D

Mg 50 < 1 D D 1 < < 1 D < 1 D

Cu 25 < 1 D D 1 < < 1 D < 1 D

Mo 50 < 4 D < 4 D < 4 D < 1 D

CO 6 D N 1 < < 1 ND < 1 D < 4 D

D; Detected ND: No Detected

f) Conversio2 U0 o t n

The conversion to IX^ stage has here always been carried

out in a tray furnace. The characteristics of the opera- n generai tion s sha l terms been performe s followsa d :

Racks of four trays are loaded with an AUTC height of m eachc 5 ;1, thee thear yn introduce e vestibulth n i d e

54 without heating stayin n nitrogei g n atmospher r somfo ee minutes. Then they are carried to the heating area at a températur f 700°o e c being them submitte n alternativa o t d e treatment for a period of 20 minutes in dissociated ammonia (NH-j )the; 0 minute4 n n Nitrogei s n (No0 2 d agai an r ) fo n minute n dissociatei s d ammonia. Rack e transferrear s o t d a cooling a temperaturare t a a e samth 120°. ed of asaian C d treatmen n equa i te heatin th l t timea g s a ares s thei a n given. From here the e takea coolinar yo t n g area wita h temperatur f 70-BO°o e d passinan C g finall a passiv o t y e area where the oxygen-uranium rate is adjusted with a nitrogen treatment being they then put into hermetic containers (drums).

Physical characteristics of_UC>2 powder

Test Apparent Flowing Specific Rate Moisture density g/cc sec. surphase 0/U mVg

LG 3 1,80 1,00 5,59 2,12 0,23

LG19 1,88 1,10 5,42 2,09 0,18 LG24-1 1,86 1,10 5,62 2,10 0,20

LG36-1 1,80 1,10 6,76 2,15 0,26

The following are the latest results obtained with the pellets sintered at 1700°C in dry hydrogen atmosphere with 2 m-vh volume and a pressure of 30 milibars in an experimental furnace.

55 RESULTS OBTAINED IN AN EXPERIMENTAL FURNACE

T COMPLEJA O FABRIL CORDOBA (ARGENTINA)

Test N° Pressure Densit, in y Sintered t/cm2 green g/cm^ density g/cm2

LG 3 3,4 5,37 10,32

11 3,34 5,47 10,43

II 3,63 5,58 10,50

M 4,08 5,67 10,56 ii 4,30 5,73 10,58

ti 4,60 5,79 10,60 LG 19 3,04 5,33 10,17

M 3,71 5,54 10,40

If 4,23 5,69 10,50

ft 4,97 5,80 10,58

11 5,92 5,97 10,62 LG 24-1 3,11 5,34 10,35

tl 3,78 5,56 10,49

11 4,23 5,70 10,55

II 4,52 5,79 10,58

II 5,04 5,90 10,62

LG 36-1 3,56 5,42 10,43 D 4,23 5,61 10,55 K 4,37 5,65 10,57

n 4,52 5,67 10,58

. External characteristics of pellets: good . Internal porosit pelletsf o y : good . Size : Specified . Grain size in crossed section of pellets: - external area : 8 y - central area : 70 P

56 RESULTS OBTAINED IN A PRODUCTION FURNACE

T CENTRA O ATOMICO EZEIZA

Green pellets have been obtained through the pressing proces t Compleja s o Fabril Cordoba. Froo differentw m t lots a number of ten green pellets each at its corresponding working pressure have been separated. Half of this material from each group has been sintered at Complejo Fabril Cordoba attaining the results shown in the preceding table, while the other half has been sintered in the Production Furnace at Centro Atomico Ezeiza with the following results:

Test N° Pressure Densitn i y Sintered t/cm2 green c/cm3 density g/cm2

G 24-L 1 3,11 5,34 10,46

It 3,78 5,56 10,50 ri 4,23 5,70 10,52

n 4,52 5,79 10,53

ti 5,04 5,90 10,50

LG 36 3,56 5,42 10,49

ft 4,23 5,61 10,54

It 4,37 5,65 10,54

fl 4,52 5,67 10,56

. External characteristic pelletsf o s : good . Internal porosity of pellets: good Siz. : e Specified . Grain siz crossen i e d sectio pelletsf o n : - external area : 11,9 \V - central area : 12,5 p

57 At this time . '- is being intended to change sintering condition n ordei s o obtait r n sintering densitie a lowe t a sr compacting pressure and procure a better grain growing of the sintered pellets.

) Çf.coyer_Y_of„Reagentg s §nd_Effluents e 50-6f raffinâteth o % ) 0 b s sai i n i dt i s s sulfuriA c solution that come t froe Extractioou s th m n through Solvents Plans i t recovered to uranyl sulfate (SO. UOp); the remainder of this liqueu s use i ro neutraliz t d mothere th e s liqueur from AUTC crystallization t awathapu y t e b havfroe circui o th mt e e du t to its high sulfate content. Furthermore, all acid vapors or ammoniacal vapors coining out of the process and which are absorbed in different absorption towers are added to these solutions, being uranium draw t througou n h Precipitatios a n ADU and solutions where uranium has already been removed are being concentrated getting ammonium sulfate as final product

fertilizing quality meet(S0t (i 4(N s2 ) HIRA M Argentine Specification U ° 2241631.841.1)sN CD - 0 .

REQUIREMENTS x UniMa t n Mi

Ammoniacal Nitrogen 20,5 Sulphur 23,4 Free acidity expressed s sulphuria c acid g/100G 0,1 (H2S04) Moisture 1 Caught in siever IRAM 850/v 15 Sieved IRAM 450 p 30

58 Standard analysi f ammoniuo s m sulfate carriet C.F.Ca t ou .d

REQUIREMENTS Unit Obtained Results

Ammoniacal Nitrogen 21,03 Sulphur 24,03 Free acidity expressed as sulphuric acid g/100 G 0,08 (H2S04) Moisture 0,6 Caugh n Sievei t r IRAM 850" v / 17 Sievey d IRA0 45 M 28

We have arrived to this alternative since Safety Specifica- tions appliee Argentinth n i d e Republi t allono o wd c liqueurs with sulfates higher tha 0 mg/e vacatedb 40 n o 1t d becausan , e from an analysis estimate cost performed in advance offer significant savinge operatioth n i s n cost tha f calciui t m oxides were use o neutralizt d e samth ee liqueurs.

) A_futurh e outlook_of„this_Methodolog_i;

According to tests which have been carried out and opportunitely informed, the present methodology intends to obtain, starting from impure solutions, AUTC which meet Nuclear Purity Specifica- tions. This means that the possibilities offered by the sulfuric leaching treatment of ores appropiately clarified, filtered and conditione e verar dy promissor d theran y e exist e certaith s n possibility that an AUTC nuclear purity may directly be obtained, it means without the previous step through ADU precipitation.

59 Représentât ive_aQalY s_for_liguors_to_beSi _ treated

U308 ——————— 1,8l g/ 0

e F ——————— 11,90" Cu ——————— 5,25 " H p ——————— 1,40" P. O.K. —— — — - 500 mv

From these liquors purifie y amineb d s AUTswa C obtained which representative analysi n threi s e main impurities: Silex, Iron and Copper gave the following results being the remaining impurities in according to Nuclear Purity Specifications;

SX _ — ____ — _ 30 ppm Fe — ——— — -- 80 ppn Cu ---- —— — - 20 ppm

Quality Control, Radioprotection and Safety Specifications taking into account during operation at a Pilot Plant

ontrol

e SectioTh n responsibl e Qualitth r yfo e contro t Conple^a l o Fabril Cordoba, with activities based on verifying the fulfil- men f Proceedino t g Manuals, Chemistr d Physican y s Laboratory

Manual, Inspection and Testing Plan, Operating Manual, Commissions and Functions Manual, as well as others, is at the present time preparin e "Qualitth g y Control Manual" that will govere th n Quality Control Program and to which Plants will have to obey.

60 ADU f l l H20

D1SOLUTION ADU

R AFIN AT..E. H20 f

DILUTION ABSORPTION TOWERS

SOLVENT NEUTRALIZATION EXTRACTION

l f

PRECIPITATING ELUCTION PRECIPITATION ADU MOTHER LIQUOR

ATU CRISTALS FILTRATION ARRANGEMENTS

CONCENTRATION FILTERING AND WASHING AUTC

FILTRATION

CONVERSION 2 U0 TO

S04I

Flow-sheet of the process.

6i/4l STATU URANIUF SO M REFININ CONVERSIOD GAN N PROCESS TECHNOLOGY IN KOREA

I.S. CHANG, S.T, HWANG, J.H. PARK Conversion Process Research Division, Korea Advanced Energy Research Institute, Daejun, Choong-Nam, Republi f Koreco a

Abstract

The nuclear grade uranyl nitrate solution is prepared from yellow cake in the purification pilot plant, at Korea Advanced Energy Institute(KAERI), Korea. This pilot plant has several 15 cm diameter pulsed columns which perform a series of extraction, scrubbing and stripping r pulsatioprocesseai e eh n i e pulses mode th n dI . extraction column, the crude uranyl nitrate solution of 2N free- aciditn contaci t ge yt counter-currently wit e disperseth h d phase of 40% TBP/dodecane mixture. The uranyl nitrate is then extracted into the organic phase and subsequently reextracted into the aqueous phase to yield a purified uranyl nitrate solution in the stripping column e finaTh . l uranyl nitrate solution thus produced is well within a nuclear grade purity and the small uranium content e raffinatth n i e aqueou e strippes th phas d an ed organic phase, less m respectivelypp 0 10 d an ,m pp showtha 0 e processe1 ns th thal al t s in the plant are satisfactory. Moreover, Mo content of 1500 ppm in the yellow cake is reduced down to less than 1 ppm in the purified uranyl nitrate solution. For the production of sinterable U0„ powder, ammonium uranyl carbonate (AUC) is precipitated in the saturated ammonium carbonate solution using nulcear grade uranyl nitrate solution together with ammoni d carboan a n dioxide gases. The free-flowing and granular type AUC powder is then converted into sinterabl „ powdeU0 e y reactiob r n with a hydroge n i s ga n fluidized bed reactor.

All parameters, which can control 0/U ratio, flowability, tap density, specific surface area, pore size and its distribution, have been studied.

63 1. INTRODUCTION

The construction of nuclear power plant in Korea started wit hnuclea1 Kor . iNO r reacto d increaser an 1971 d rapidly through the oil crisis in 1970's. As of 1986, 5 nuclear power reactors are in operation and 4 more reactors are under construction. Along with increased nuclear power generation, Korean government undertook the project for the localization of nuclear fuel manufacturing technology e policTh .f Korea o y n governmenr fo t the nuclear fuel is that nuclear fuel for pressurized water reactor(PWR) wil manufacturee lb d using transfered technology from foreign fuel vendors, and fuels for Wolsung reactor(CANDU) wil producee lb KAERy b d I using domestically developed conversion and fabrication technology. KAERI has established a conversion and fabrication research laboratories in 1979 and 1978, respectively o develot , e fueth pl manufacturing technology and to train man-power in the field of nuclear fuel. Utilizing this facilities and accumulated experience in fuel manufacturing, the development of Wolsung nuclear reactor fuel was started n 1981i .

n SeptembeI r 1984 4 fue2 , l bundles mad KAERy eb I using RBU-U02 powder(West Germany) were loade Wolsunn i d g reacto werd an re discharged after successful performance. In November 1985, another 24 fuel bundles using KAERI-U0„ powder e loadear d agai Wolsunn i n g undee reactoar rd an burn-upr .

Therefore e believ,w e thatechnologe th t y developmenr fo t

the CANDU type nuclear fuel manufacturin s successfullgwa y ful-

e plafillew o expand t n e capacitiean d th d f conversioo s d an n

fabrication facilitie e fuel th n orde i sse o supplth needet r l al yd

for Wolsung reacto y ourselvesb r .

2. DESCRIPTION OF PROCESSES

The processes of the pilot plant at KAERI are consisted of

two parts; first, a 20 kg-U/hr (100 ton-U/yr) capacity of dis-

solution, filtratio purificationd nan , second 5 ton-U/y2 a , r

64 capacit „ powdeU0 f o ry production e blocTh . k diagra f uraniuo m m refining and conversion processes is shown in Fig.l.

Fig.l, Block diagram of uranium refining and conversion processes.

65 3. DISSOLUTIO FILTRATIOD NAN N

It takes about 2.0-2.5 hour eacr fo sh operatio dissolutionn i n .

First, about a 250 1 of 10 N HNO„ is introduced into the dissolver

and preheate 50-6o d thet an p n0C u d slowly added U_0 e tem0.Th - J O peratur dissolvee th f o e r shoul kepe b d70-8t a t 0whilC e paying

attention to the temperature rising due to the exothermic reaction. e finaTh l free acidit s controllei y e fina th o 3.5-2.t dld an 0N

uranium concentration becomes abou 0 g-U/190 t . Durin dise th g-

solutio s scrubbei s e emittega th n dX witNO d h abouNaO% n 0 i H1 t

a packed columnd poinen s chose n NaOi td i ,an 0 H t abou1 a n H p t

solution d the,an n uranium recover wastr o y e treatment process

uses i d accordin uraniue degree th th o f t o egm contente .Th

dissolved uranyl nitrate solutioC 5 hour2 9 s age r i nt sa fo d

for silica agin d proceedean ge filterin th o t d g syste removo t m e

the insoluble materials.

Two set of rotary filters are used for filtration. The first

filter is for the removal of the slurry in the aged solution and

second recovere filteuraniue th th r f fo ro y m containee th n i d

first filter cake and treated as a repulped slurry. The uranyl nitrate solution from the first filter becomes almost clarified.

At this stage, the uranium concentration is about 350 g-U/1

and free acidity about 1.5-2.0 N (1). The cake from the second

filter, after estimating the uranium content, is used for the

uranium recovery or treated as waste through the neutralization

step neutralizatioe .Th carries i n witt ou dh lime.

66 PURIFICATIO. 4 N

Purification process is consisted of extraction, scrubbing,

strippin d solvenan g t regeneration sections extractionn .I , there

o extractioartw e n columnsr stronfo e g ,on liquor (uranyl nitrate

solution from dissolution) and the other for weak liquor (uranyl

nitrate solution from the bottom of the strong liquor extraction

column).

Main equipments are perforated pulse column for purification,

mixe d packean r d colum solvenr nfo t regeneration e leve.Th f lo

interfac eact ea h e controllecolumb n nca d wit pressure hth e

difference checked from the air purge system using two dip

tubes whic e attache e columhar th o t nd disengaging section.

Pulse energy generated by the air pulsation mode is introduced

colume ith n e aqueounTh . throug ) s (2 pulse solutio g th h le e n

continuoue asth organise th phas d n i cean P solventTB % 0 ,4

dodecane by vomule % , as the dispersed phase, are used in this

process (3). All pumps are diaphragm type metering pumps having

UNO, Strong liquor

TO MiVMf «0*k

Solvant tank packed column TO WMt*

~•- To prtclpitftfon prooi«« W«ok liquor tar* Solnnt buff«- lank far UOj pe*«'«r H StorifUN i tank

WEAK LIQUOR STRONG LIQUOR SCRUBBING STRIPP «s tOLVENT REGENERATION EXTRACTION COLUMN EXTRACTION COLUMN COLUMN COLUMN

Fig 2. Purification system

67 a good accuracy and explosion- and acid-proof. Each column

diameter is 15 cm, effective height for both extraction columns

6 m each, scrubbing column 4 m, and stripping column 7 m. The

overall diagra s showi m Fig.2n i n .

4-1. Extraction

e aqueouTh s "strong liquor" solution produced through dis-

solutio d filtrationan n processes which containe g-U/0 35 d 1

and 1.5-2.0 N of free acidity, is fed to the top of the strong

liquor extraction column. This aqueous solution flows center-

currently with the organic solvent. The slightly loaded solvent

from the top of the weak liquor extraction column is fed to

the bottom of the strong liquor extraction column through

the pulsation leg e aqueou.Th s solution froe bottoe th th m f o m

strong weae liquoth k f ro p extractioto e th o t n d columfe s i n

liquor extraction column, and fresh organic solvent which is

pre-equilibrated with nitric aci s enterei d d into bottof o m

this column. Two flows are also contacted counter-currently

each other e disperseTh . d phase hold-up appear parabolia s c

distribution alon colume th g n axid showan s maximusa 3 1/ t a m

position from the bottom of the column and the overall mean hold-

up is about 10%. The uranium concentration of the loaded TBP from

strong liquor extraction colum s abou i n5 g-U/ 14 t 1 corresponding s capacity, it uraniue tth f oo d abou% m an 0 9 contente th f o t

raffinate(liquid waste) from weak liquor column is dropped down

belo ppm0 2 w .

The control element in the extraction process is molyb-

denum content, whic containes hi d about yellon i 150 m 0pp w cake

68 as show e amounn Tablo contennpuri Th e M . th f 1 eo t n uranyi t l

nitrate solution whic bottos i h m produce strippinth f o t g

column is checked every 4 hours. The uranium concentration of the o extractiotw e th f o n p columnloadeto P froe TB ds contith i ms -

nuously determine y recordinb d date th ga with densimeter installed

at the top of the column and converting into concentration with

the known value of the solution density. Samples are taken from

the sampling points along the column axis and local hold-up and

the uranium concentration betwee o phasetw ndeterminee ar s d from

these samples. The uranium content in the raffinate is also

checked every 4 hours interval.

Tabl 1 Compansioe f impuritieo n n purifiei s solutioN dU n and nuclear grad ^ powderUC e .

, , Pun-ie, „ d Nuclear grade w caklo l e Ye UN solution UOj powder

U 135g-U7C *87

Ho *0 15 0 73 2

0 1 Fe 5 3 C * 20

Si *0 40 10 10

Ca 1C 2

Mg 4 1 2

Cu 1 20

Ni 5 5

Cr 5 11

Mn 1 1

B 0 1 0 2

Pb 0 3 1

* Unit • percent, others ppm based on U

69 4-2.Scrubbing

e loadeTh P froextractioe dTB th m n ste scrubbes i p d using

strippee th para f o t d uranyl nitrate solution. Becaus f 150o e 0

o contentM ppf mo acidite ,th scrubbinf yo g solutio adjustes ni d y addin b HNON N 5 „1 g4 solutioo t e p th u f o n p througto e th h

column. The flow rate of the scrubbing solution is about 10 %

d oaqueouloadee fan th thaP f TB dto s solution frobottoe th m m

of this colum s recycle ni strone th o gt d liquor extraction column.

In each phase, uranium concentration coming out of column top and

botto keps i m t nearle inle th same tth s y ea concentratioe th d nan

overall mean hold-up in the column is controlled up to 10 %.

4-3. Stripping

The demineralized water is used as the stripping solution and

the flow rate is adjusted to maintain the uranium concentration

of 130-135 g-U/1 in the pure uranyl nitrate solution. In order to

keep the operation temperature about 50-55 C, the demineralized

respectively, C 5 5 d watean e preheate d loadeC ar an r5 ,P 6 TB do t d

and is fed into the column top and bottom. The uranium concen-

tration in the uranyl nitrate solution from the column is cons-

tantly determined by the measurement of the density and the Mo

content is analyzed every 4 hours. Particularly, when the uranium

concentratio produce th f oves o n0 g-U/1i t 14 r , uranium contenn i t

the stripped TBP is increased by gram-order. In addition, the

stripped TBP solution coming out of the top of the column passes

the coalescencer filled with the teflon wool in order to remove

the very fine particle f saqueouo s phase entrained with organic

70 phase. The purity of uranyl nitrate solution is well agreed with

the nuclear grade of the U0„ powder and Table 1 shows the com-

parison of the feed materials and purified uranyl nitrate solution

wit nucleae th h r grad „ powdereU0 .

4-4. Solvent regeneration

In the solvent regeneration process, about 8 % (wt) sodium

carbonate solution is contacted with the stripped TBP solvent

in order to remove all impurities. The flow rate of the sodium

carbonate solutio f thao o f solvens abou% o ti ntw 0 1 d t an t

phase mixee ar sy recirculatio b d n using centrifugal pumd an p

then sent to the settler through the teflon wool (4), The sol-

vent is contacted conter-currently with 2 N nitric acid solution packee th n di colum orden i n acidifyo t r e aqueou.Th s solution

fropackee th m d colum uses i nweas a d k liquor d acidifiean , d

organic solven s recycle i tweae th ko t dliquo r extraction

column. The sodium carbonate solution used for regeneration is

treated with the sodium hydroxide solution to recover the uranium.

5. EVAPORATION/CONCENTRATION OF URANYL NITRATE SOLUTION

n thiI s process concentratioe th , uranye th f lno nitrate

solution froe solventh m t extraction increases fro0 g-U/13 m o 1t

420 g-U/1. The evaporator is a cylindrical form with 40 cm in

diameter and has a teflon lining in order to prevent the intake

of impurities possibly generated from the cylinder wall and has

an internal electric heatine heateth s a rg source.

The solution temperature is controlled at 103 C and the solu-

tion is kept evaporated at higher concentration without boiling,

71 and one batch operation is terminated by the density measurement

uraniue oth f m nitrate solution.

advantagee Th s obtained from concentration/evaporatio s followsa n ;

-Possible contro particlf lo e characteristics suc s sizha e

distribution of the final U0„ powder.

-Smaller scale equipments, that is, smaller capacities of AUC

precipitator, filtration receive r storago r e tank because

e decreaseoth f d uraniue volumth f o em solutio o handlet n .

r instanceFo , 0 g-U/whe142 na solutio uses i n d insteaf o d

a 130 g-U/1 uranium nitrate solution, then the AUC precipi-

tator 's volume becomes 60-1 instead of 140 1.

-Reduction of gas and power consumption.

bettes Ii t keeo t r p constant ammonium carbonate concentra-

tion in the mother liquor to obtain the constant precipita-

tion conditions resulta s .A amoune ,th f ammonio t d car-an a

n dioxidbo decreases i e . abouo % t d 0 3 t

-Reductio f liquino d waste.

PRECIPITATIOC AU . 6 N

In this process AUC is prepared by the reaction of C0„ and

NH_ wit uraniue th h m nitrate solution accordin totae th lo t g

reaction equation;

2NH.NO- 4 J

The precipitator made from stainless steel, is a cylindrical form

with 40 cm in diameter and the slurry is circulated from the bottom

e precipitatoth par f o te centra th o t rl par y usinb te pump th g .

72 Carbon dioxid ammonid ean a gase introducee sar d into solution

through circulation lines e precipitatio.Th carries i n t ou d

batchwisely and preceeded as following three steps;

-Preheating; In order to obtain ammonium carbonate solution

and to raise the temperature up to 58 C, g'ases are fed into

the precipitator. Solution is heated with both heat of re-

action and external hot water circulation.

-Uranyl nitrate solution feeding; AUC particles are produced

whilC 8 a5 te feeding uranyl nitrate solution together with

ammonia and carbon dioxide gases. Crystals deposit at the

gas nozzl d ofte ean s feedin nga plue gth g line n thi.I s case

the distilled wate automaticalls i r y injected through nozzle

d wasan h thet accordin ou mpressure th o t g ee builth n i t

gas lines.

-Cooling; For easy handling of AUC slurry and for the

reductio uraniuf no m contenmothee th n ri t liquore th ,

reaction solution is cooled down to below 20 C and further-

more ammoni d carboaan n dioxide gase e introducedsar .

One AUC particle is calcined and reduced to one UO particle (5)

C conversioAU e ith n n process particle th , e size distribution

„ particled shapU0 an f eo s depend completel particlesC AU n o y .

Consequently, granular shape of AUC particles, that is, granu-

lation is essential for this AUC conversion process. Hence, it is

very importan controo t siz e particleC d shaplth AU ean f eo s

in the precipitation.

-Morphology; One of the advantages of the AUC conversion

procesexcellene th s i s t flowabilit „ powderU0 f yo , which

s granulaha r shape excellene .Th t granulatio mainls ni y

due to attrition of AUC particle edges by external circu-

73 lation. Durin precipitatione th g lone C crystalth ,gAU s

(show Fig.3)n ni , havin ratie th glengtf o diameteo t h r

over 1:10 e generatear , dt no t dependinthes bu e ar H ep n o g

desirable products. It is necessary to control the ratio

possibl, 3 1: 1:o yt 1 with wide rangvalueH p f eo s between

7.8 and 10.

-Size distribution; As it is known(6), the size distribution

depends on the desupersaturation rate from the showering

e earliepointh n i tr stag f precipitationeo e operatioTh . n

parameter e ratth e r controfo s e temperaturear l s feega ,d

rate, operation time, and uranium concentration in the

uranyl nitrate solution, but the easiest way to control

the rat s uraniui e m concentration itself a result s A . ,

0 g-U/42 f solutio1o uses ni n thi i d s process.

1QO IS HOÖS Q !

Fig.3.Scanning electron micrograph of AUC crystal ( X 3,000)

74 7. FILTRATION

AUC particles produced from the precipitation process are

separated from the mother liquor and dried. The average AUC

Particle siz abous i e /im0 4 t , whic s relativelhi y large o simpl,s e

vacuum filter can be used. In order to increase drying rate,

methyl alcoho uses i lwashins a d g solution.

8. CALCINATION AND REDUCTION

The AUC particles are converted into the UO particles by

reaction with hydrogefluidizea n i s reactod nga be d r according

to the following equation;

This fluidized bed reactor is 17 cm in diameter and 220 cm

in height. The AUC powder are fed from the hopper using the

e bedvibratin th n orde I .f o preveno t rp gto e feedee tth th t ra

O U powder e th los f ,catridga so e filte s used backrwa an d -

flushed ever minutes0 2 y calcinatioe .Th d reductionan cons ni -

sisted of the following three steps;

-AUC feeding e reactio;Th n occur 520°t sa f o C % wit 3 1 h

f steao % hydrogem 7 atmospher8 d e temperaturnan th d ean e

can be adjusted by feeding rate of AUC powder. In this

step, either hydrogen concentration or temperature changes

within 10 % doesn't significantly influence the powder

characteristics .

-Pyrohydr o„ powder lysisU0 e onle ;Th sar y treatee th n i d atmosphere of steam in this step. This treatment influences

75 significantly on the U0„ powder characteristics and it

e ste densityp controo th t p ta s i e th l, specific surface

area, por distributios eit sizd ean n together with time

and temperature. The tap density changes virsus time at showe ar Fig.4n ni C 0 .65

3.2 -

3.1

0) 3.0 11C •a

O. Q. 2.9

2.8

20 40 60 80 Pvrohvdrolvsis time, rnin

Fig. 4. Tapped densitv change of UCH powders MS pyrohydrolysis time.

-Reduction ;e lowerin Thith r sU rati fo ste0/ g o s t i op

0 fro2 resulte2. 2. m d fropyrohydrolysie th m s step.

The UO particle characteristics affect mainly on the sinterability. 2 For instance,specific surface area gives great deal of effect on

the sintered density as shown in Fig.5. Accordingly, it is very

important to control and adjust the characteristics of the U02 pow-

der, possibl n thii y s step. However, these characteristic- in e sar

terrelated impossibls eaci t hi othe d adjuso et ran t independently,

76 10.6

BO 3>•> 10.4 Wc

T) u V

5 lo.o • 4ton/cm

_L. _L 4567

Specific surface area U(>2 powder2 m ,

Fig. 5. Relationship between surface area of UC>2 powder and sintered density of pellets.

Fig.6. Scanning electron micrograp O U powde f o h r 20,000X ( )

77 smallea tha, der^itp is t ta r y mean slarga „ eparticle U0 por n i e s

and also leads to increase in specific surface area. This inter-

relationshi explainee b change n monocrystallina th ca p f y eo b d e

to a polycrystalline U0„ particle. As shown in Fig.6, U0„ particle

has a great number of small crystallites. The larger the size of

the crystallite densityp e smallebiggee ta th th , e e d th rth r,an

specific surface area. Consequently, the crystallite size must be

controlled for the sinterable U0„ powders.

9. STABILIZATION

This step is the partial oxidation process of U0„ to have

a stable oxidized film on the U0„ particle surface (7) and increase

of the ratio of 0/U in the controlled oxygen atmosphere. Since the

„ powderU ratiU0 0/ f o varies i s d wit „ concentrationh0 ,

temperatur controllee b d timet ean U no rati ,0/ n e dca o th onl y b y

operation parameters. Tha, specifiis t c surface area should first

be controlled.

URANIU. 10 M RECOVERY

The filtrate contains 600-800 ppm of uranium, because _ i i CO- ions tend easily to form complex ions with U0„ ions in the

mother liquor (8). Hence essentias i t ,i eliminato t l , ionCO e s

in order to recover uranium from the filtrate. The filtrate is

heate98°o t Cp du ,wit h stirring e finaTh removo . l,t „ C0 e

solutio turnH np abouo t s wit9 t h precipitatio ammoniuf no - di m

uranate(ADU) uraniue .Th m conten finaf o t l liquid waste becomes

belo ppmw2 .

78 11. CONCLUSION

The project to localize uranium refining and conversion technologies, as a part of HWR fuel manufacturing»has been successfully completed throughou e in-filth t e test, out-of-file

tesd evean t n throug actuae th h l irradiatio commerciaa n i n l power reactor. This accumulated technologies and experiences will help us to expand our capacity in order to supply all the fuels

Weisune foth r g reactoneae th r n futurei r .

REFERENCES

(1) Ashbrook, A.W., Uranium refining and conversion practice in the Western World; An overview,(1982)

(2) Srinivasan,N., Kumar,S.V., BARC (1972) 589

) Ferez(3 "Duconcentr, ,A. e d'uraniu l'hexafluorure"a m , Proceeding Advisorn a f o s y Group Meeting, Paris 8 Jun5- , e (1979) 201-228

) Leroy,P.(4 , Etud solvenu d e TBP-dodecane% 0 3 t , SCC1-71,(1966)

(5) Assmann, H., Mathieu, V., AED-conf-76-194-007 Pape re 78t , 197presenteth 5 h t 6y Annuaa Ma n o dl Meeting

Americae oth f n Ceramic Society, Cincinnati, Ohio.

(6) AIChE Sym. Ser. 68(121), (1972) 8-20 Desupersaturation of seeded citric acid solution in a

stirred vessel.

(7) Sonndermann, T., J. Nucl. Mater., 106(1-82)45-52

) Chernyaev(8 , 1.1., "Complex Compound Uraniumf o s " Israel Progra Scientifir fo m c Translation, Jerusalem,(1966)

79 RESEARCH AND DEVELOPMENT OF UF6 CONVERSION IN JAPAN NOW AND SUBJECTS IN FUTURE

Y. HASHIMOTO, I. IWATA, T. NAGASAKI Power Reacto Nuclead an r r Fuel Development Corporation, Tokyo, Japan

Abstract s engagini C e PN followinth g g 6 conversionR&D'UF f o S .

(D Rifining and Conversion Pilot Plant (200TU/Y) which produces UF6 from uranium concentrate C procesPN e th s y b s

D ( IJFs conversio f reprocesseo n d uranium (4TU/Yy procesdr y b ) s

PNC has accoumulated many important experiments and been making many

improvements.

We have to solve the following subjects to commercialize the UFs conversion in

future.

These subjects will be solved by combining reprocessing, conversion and

enrichment facilities at the same site to reduce total fuel cycle cost.

We are willing to contribute the world's UF6 conversion industry by developing

the latest technology.

81 1. 1 ntroduct i on

Originall C procesPN y s develope wa e smeta th r l fo fued l productio n Japani n ,

which produces UF4 directly from ore without the yellow cake production. Wite developmenth h e Lighth f t o tHate r Reactor which uses slightly enriched

uranium, the feed material UF6 became needed to produce after LF4 production process.

o thaS C startePN t o develot d e succeedinth p g proces n 197i s d the6an n

succee o product d e UF6. Based on these success, PNC started to construct the (jFe conversion pilot

plant of 200T/year in 1979 to promote further industrialization and supply UFS e enrichmenth o t t pilo te samplanth en o sitet .

This plant was improved to produce UF6 not only from ore but from the yellow cak ee currenwhicth s i h t materia e worlth n di l

Froe operationth m C proces PN e founw ,e e merit th th sd f froo s m yellow cake

and with enrichment plant.

e otheth n rO hand, witf e Tokastaro th hp u it Reprocessing plant t becami , e

needed to develop the recycling technology of reprocessed uranium to LWR

through UF6 conversion and enrichment. o thae S conversioth t n test facilit f reprocesseo y ) Ï (CTÏ F- d uranius wa m

constructed in the same plant.

Froe testth m e f founCTF-w o s e meri, l th d-f n co-conversio i t n with

reprocessed and natural uranium.

The reseach and development of UF6 conversion by PNC process before the pilot plant constructio s alreadha n y been reporte e previouth t a d s IAEA Aduisory Group

Meeting hel n Pari i , dthereforn 198) i s 1 0C e wilw e l report here th e

experiences at the pilot plant of PNC process and CTF- D of reprocessed

e uraniufuturth d ean m subjects.

82 2. UFe conversion in Japan now.

The following development has been carried out in Japan

(1) The pilot plant test of refining and conversion © Feed Yello• ' s e wworl th d cakdomesti n an di e c ore.

@ Process : PNC process (see Fig-1)

(3) Capacity : 200tu/year

® Start up : March 1982

URANIUM CONCENTRATE

DISSOLUTION

(H2S04)'

(HCI) AMINE EXTRACTION

(N32C031

UO2S04 (H2S04) u ELECTROLYTIC REDUCTION

U(S04)2

HYDROFLUORINATING

UF4 -nH20 T

(N2) DEHYDRATION

J UF4

(F2)——[FLUORINE FLUORINATIONJ

UF6

Fig.1 REFININ CONVERSIOD GAN N PILOT PLANT

83 (2) The U F s conversion test for reprocessed uranium. (CTF-Ï) Ï

(D Feeds : reprocessed uranium, U03 (U235 below 1.6%)

@ Start up : March 1982

A theme (2) will be reported in another session.

REPROCESSED URANIU3 MUO

GRINDING AND CLASSIFICATION

U03-nH20

DEHYDRAT10 REDUCTIOD NAN N (H2)- r

U02

(HF}- HYDROFLUORINATION

(F2)— FLUORINE FLUOR1NATION

UF6

Fig-2 CONVERSION OF REPROCESSED URANIUM

3. The experiences and improvements of the Ri fin ing and Conversion pilot plant.

) (1 Process flow (See Fig-3)

C procesPN e Th s consist e followinth f o s g steps.

Dissolusion with sulfunc acid (See Fig-3)

84 H2SÛ4 (recycled)

Uranium I Concentrate I [ n(U0 2SCM

Filter Extraction Washing Strippin Scrubbing| g| ) 3 steps 2 steps 7 steps Dissolution Tank To Raffinate U : 1mg/J Electrolytic Liquid waste Reduction treatment 60~100gu/f

U02S04 UO2Cl2 H2S04 Fig-3 URANIUM CONCENTRATE DISSOLUTIOD NAN AMINE EXTRACTION

Yellow cak s dissolvei e d with sulfunc acid, thee residuth n s i e

filtrated out.

) Amine extraction (See Fig-3)

Dissolued uraniu s extractei m d into tri-n-octyl amine solvent, selected

from impuriries y forminb , g negativ e2 compleU0 (504)3 [ n io x! The equipment consists of 4 stage extracting, 2 stage scrubbing, 7 stage

strippin 2 stag d an eg solvent scrubbing mixer settlers.

D Electrolytic Reduction (See Fig-4)

2 4 Uranyl (U02 *) solution is reduced into uranous (U *) solution by the

electrolytic method. Electrolytic cell is separated into an anode room and a cathode room by

an anion exchange membrane.

Sulfunc aci s anolyti d d uranyan e l sulphat s catholytei e . U02S04 2 2 60~100gu/£ H2°S 04 Electrolytic UOjSCu II Reduction cell UOzC'z

2 O H2SCU YÎ t /•Electrode uo Electrode H 0 . Electrode ,-Ion-Exchange 2 membrane

Ion-Exchange î i Membrane U(SO

Electrolytic reaction 1 To Hydrofluonnation + ) 2H20-»4H++02+2e~ U(SO„)2 UCU

Fsg-4 ELECTROLYTIC REDUCTION

Hydrogen ion produced at anode passes through the membrane and reduces

U02S04 into U(S04)2 at cathode.

) Hydrofluortnat i ng precipitation (See Fig~5)

Uranous Solution is precipitated to hydrated uranium tetraf 1 uor ide

ÜF nH• « 20 with hydrofluoric a reactoaci n i d r vesse 8 cubi f o lc meter made of teflon coated fiber-reinforced plastic.

) Dehydration (See Fig-5)

Hydrated uranium tetrafluo « -nHUF s dehydrate2i e idized-be0u rid fl a n i dd reacto f 35co r m diameter wit ^ gasN h .

) Fluorine fluorination (See Fig-6)

Dehydrated UF« is converted to UF6 gas with F2 gas, and then UF6 gas is cooled and trapped in cold traps.

86 jj=>To Vent — '-————*>To liquid waste treatment

90'C HYDRO- Fluorination Precipitator

U(SCM

2 Fluorinatiof- fj n

Heated Air Dryer

Fig-5 HYDROFLUORINATIO DEHYDRATIOD NAN N

UFe gas UF4 To Vent Cold Trap

NaF trapLJ A|2p3

\ -5'C —3U'C 100'C

s ga F H + 2 H Fan- s ga HF Fluid Ded Scrubber I Fiuorinator

To Vent

N2 gas 1 gas

H F Electrolyisis Cell

HF Scrubber

Fig-6 Fa FLUORÎNATION

e fluorinatoTh r is' luidized-bef a d reacto f 40cio r i diameter.

) Liquid waste treatment (See Fig-7)

e followinTh g liquid wast b dischargei e d froe processth m .

) 1 Raffinele from awine extractor containing H 2d S0impuritiesan 4 .

) 2 Solvent scrubbing waste containing Sodium carbonate Na2C03.

3) Hydrof1uorination waste containing H2S04, HC1, HF and U.

87 EXTRACTIO SOLVENI 1 N T SCRUBBIN j HYDROFL GI U URINATION DEHYDRAHON F2 FLUORINATION 1 PRODUCTIO2 F 1 2S0N 4 , u£ * U ru l

Raff nate M^ ^^ , H2S04 [ACID RECOVERY

rCa(OH>2 T | H2S04 r n y ...... _ Q H Alkaline 1 Recycle Scrubbing

* NaF I DtCARBONATION 1 U

CaS>O< 1 r ———— Ca(OH>2

CaF: PRECIPITATION

F 7pp U mIpp m Discharge U<0 09ppm - —— 1 URANIUM ADSORBING I- —————— FLUORINE ADSORBING

Chelating ion-exchange " r Chelating ion-exchanger

Fig-7 WASTE TREATMENT

4) Alkaline scrubbing waste in dehydration and Fz fluorination process containing NaF and U

Raffinat s treatei e d with Ca(OH) o removt 2 e containing H2d SOan « impuritie y precipitatinb s s CaSOa gc et «

Hydrofluor mat ing waste is evaporated to recover H2SO< and Uranium

Vapour containing HF and HC1 is condensed and used to decarbonate the

solvent scrubbing waste

The decarbonated waste and alkaline scrubbing waste is treated with

Ca(OH)2 to remove containing fluorine and uranium as precipitated CaF2 roughly, then small amoun f n residuawasto ti s U removei e , F ly b d

chelating ion-exchange adsorbing resi r discharginfo n g from facility

88 (2) Specification of impurities in uranium concenrate

The specification of impurities in feed uranium concenrate is determined by

decontamination factor in erch process and DOE specification of product UF6.

C proces 3 PN purificatio s e ha Th s n step s followsa s .

CD Amin extractio 4 precipitatioUF © n2 f1uorma F D ( n t ion.

e decontaminatioTh n factor d uraniuan s m concenrate specification derived

from them is shown in table-1

The elements which decrease the efficiency of electrolytic reduction are

specified further strictly, but this is not fatal.

e limiteTh d impuritie e nobl d ar , transitiosetc an Ag e. , metalCu f o ns

, etcNi . , Co , metalFe f o s

Since Co3, PCU, Cl disturb the amine extraction and As is strict regulated environ mental material, these materisl e strictlar s y specified.

3 shoe impuritied Tablth wan 2 e s specificatio f convertero n e th n i s

world. [23.

. As d an 2 SeverC0 , Mo e C proces, materialF PN , e e PO,Cl th ar s , n i s

89 Table-1 DECONTAMINATION FACTORS OF EACH STEP IN PNC PROCESS AND URANIUM CONCENTRATE'S (U C ) SPECIFICATION

estimatee u a v d

Deconta~ n factoo a n ri Uranium mpur ty UFe UFa C-, Conc°ntrate b elements Spec j^Ex'ract on *" Prec Dicatcn v Fluor nan to Spec Sb 1 ppm/U 4001 ( 1 1 100) 2ppm/U NON Nb " \ 8 000) ( 100) 4 NON Ru » 8 0002 ( ) ( 1 000) 3 NON Ta t 1 ( 8 000) 1 100) ( D NOM Ti 1 « 1 200 1 000 I 1) NON C- 10 7 30 000 ( 1 4) ( 1) NON Mo 4 1 2 7 1 200 ( 1) 4 500 W 4 1 0003 ( ) " 200 ) ' 1 NON V 4 1 80 200 ( ) 96 000 Ca > 21 000 4 70 NON Mg 650 000 445 3 NON | Th 1 3 300 26 (100) NON Zr 50 100 (100) N0\ Total Na 1 800 20 20 NON 300 K 1 300 50 60 NON

z Fe 28 600 2 000 00 NON LU u g Ni 878 000 400) 00 ' NON 2 u Co 21 i 0 2000 (100) 2 100 a D er Q Ag 1 000 1 300 100! 1 000 LJ £ Cu ,. 51 200 1 300 (100! 25 000 Z Er 5 (10 000) 1 ^ C o \ 10 000 CO 1 K- CI 100 CO 900) LO U LJ F NON 1 1 0 000 U O ~ PO* 50 (10 000) 1 10 030 X U. LJ C03 (10 000) 15 000 s A i r NON 1 400 15 000 D ( 0 00 i LJ 3 ) i I 8 4 200 4 000 NOM H O Si 00 37 0 8000 05 1 2 0 S •'OO ) i

Table-2 COMPARISO CONVERTERSF NO ' MAXIMUM ALLOWABLE SPECIFICATION LIMITS (MAXIMUM PERCENTAGES)

Constituent PNC A lied Kerr VlcGee BNFt Comurhex Eldorado Vanadiu) m(V 6 9 0 75 0 85 0 70 0 30 0 85 Phosphorus (P) 1 OIPO4) 1 00 0 70 1 00 0 0 1 0 70 Halogens (C! Br 1) 10 0 10 0 50 0 30 0 25 0 30 Flourm) (F e 1 0 0 10 0 15 0 30 0 30 0 20 Molybdenum (Mo) 0 45 0 30 O 75 0 6O 0 60 0 45 Sulfur (S) —— 12 00 4 50 —— — — — Iron (Fe) —— 1 00 2 00 —— —— —— Arsenic {As} 0 1 0 10 2 00 2 5O 2 50 2 00 Carbonate (COai 1 5 0 50 3 00 4 00 3 00 3 00 Calcium (Ca) —— 1 00 1 50 5 00 5 00 1 50 Boron (B) —— 0 10 0 15 1 00 0 20 0 20 Silicon (SO 8 O(SiCh) 2 50 1 50 8 00 5 OO —— Magnesium iMg) —— 0 50 1 50 —— —— —— Thorium (Th) —— —— 2 50 —— —— 2 50 Zirconium (Zr) —— —— 0 50 0 50 2 00 —— HN03 Insoluable ______0 10 0 10 Uranium Exîractabie Organic __ __ 0 10 0 10 0 10 Material Water (HaOi '5 0 5 00 7 50 15 00 10 00 b 00 Sodium (Na) —— 7 50 —— —— 15 00 —— Potassiu) m(K —— 3 00 —— —— —— —— Titanium (TO —— 0 05 —— —— —— —— Sulfates (S04) —— —— —— —— 10 00 10 50

90 Table-3 COMPARISON OF CONVERTERS' IMPURITY SPECIFICATION LIMITS WITHOUT SURCHARGE (MAXIMUM PERCENTAGES)

Constituent PNC Allied Keir-McGee BNFL Comurhex Eldordoo Vanadium (V) 1 0 0 10 0 10 0 20 —— 0 10 Phosphorus (P) 0 3(PO4) 0 10 O 35 0 50 —— 0 35 Halogens (CI Br 1) 0 31CD 0 05 0 2b 0 10 —— 0 25 Flounne (F) 0 3(f-~) 0 01 0 15 0 01 0 15 0 15 Molybdenum (Mo) 0 1 0 10 0 15 0 20 0 20 0 15 Sulfur (S) —— 3 00 3 50 —— —— —— Iron (Fe) —— 0 15 1 50 —— —— —— Arsenic (As} 0 05 0 05 1 00 1 00 1 00 1 00 Carbonate (COj) 0 5 0 20 2 00 2 00 2 00 2 00 Calcium (Ca) —— 0 05 1 00 1 00 1 00 1 00 Boron (B) —— 0 005 0 15 0 2O —— 0 15 SilicoO n(S 4 OISiOz) 0 50 1 00 4 00 —— —— Magnesium (Mg) —— 0 02 1 00 —— —— —— Thoriu) Th m { —— — — 2 00 —— —— 2 00 Zirconium (Zr) —— —— 0 50 0 10 0 20 —— IIN03 Insoluable __ __ 0 10 _ __ 0 10 Uranium Extractable Organic ______0 10 0 10 Material Water (HiO) 10 0 2 00 7 50 10 00 —— 5 00 Sodium (Na} —— 0 50 —— —— 1 00 —— Potassiu) m(K —— 0 20 —— —— —— —— Titanium (Ti) —— 0 01 —— —— —— —— Sul'ates (SCu) — —— —— — — 10 50

(3) The purity of product L'

e converteW e typicath d l uranium concentrat e worlth s followa n di e s

(D Ranger min n Australariai e , U308 type.

y lakKe e min n Canadai e , U308 type Akouta mine in Niger, MDU,SOU type

The impurities in these uranium concentrate is shown in table 4, and among

these e Akoutath , 's uraniu m concentrate contain d almosan o M ts impuritiee th s

most.

So, the purity of UF product converted from Akouta s uranium concentrate 6 1 it ssatisfiei E show specificatiod DO n tablan i e n , th s5 e n

91 Table-4 IMPURITY CONCENTRAT FEEF EO D URANIUM (PPM) Spec CHINA Australia CANADA Niger Impurity Standard Maximum Pingxiang Ranger Key lake Akouta V 10,000 96,000 5 660

SlO2 40,000 80,000 60 1,500 2,200 5,800 Mg —— —— 5 1,000 23 2,900 Th 20 4 Zr 5 20 Q A 7,700 Na —— —— 0 4 170 230 15,200 K —— —— 10 400 13 1.000 Ti <1 <5 <5

Table-5 IMPURITY CONCENTRATE OF PRODUCT UFe PPM/U Feed U Dissolving Stripping Impurity UF4 UF6 DOE spec (Akouta) Solution Solution 3.4 5 48 A! 1,700 1,600 38.3 > Ca 2,000 1,300 45.5 26 2 1 85 Cd .-5 <1 0 <0.3 <0 2 <.0 2 Cu <5 7.7 <0 6 < 0 4 <0 2 K 1,000 552 3.9 5 5 3 6 Total Mg 2,900 3,100 12 2 0 2 0 52 SOOppm Mn 300 217 1.9 <0 2 <0 2 /U Na 15,200 19,700 495 1 51 1 21 Ni <10 3 01 0 66 2 3 0 33 Pb <10 6 7 0.56 <0 2 <0.2 Zn 56 16.9 5 6 0 38 4 4 , P 2,900 44 0.5 0 5 50 Si 5,800 —— 0.2 <0 2 <0 2 100 Ti 101 2.5 < 1.0 <1 0 1 0 Mo 2,400 1,800 800 < 1.0 <1 0 1 4 V 3,300 482 <5 <4 <1 1 4

92 (4) Loss of uraniurn

The possible outlets of uranium from the PNC process are as follows.

® Raffinat cimin i e n extraction

D ( Solvent scrubbing waste.

(3) Hydrof1uorinat i ng waste.

6 colUF s dfroe ga trapsth m f Of . ®

There is very few uranium loss from the extration process CD, (D.

g wastn i ® t econtainHydrona i r uo % f1 uraniu 10 s m whic s recoverei h n a y b d

evaporator with H2SU4 to recycle to uranium concenrate dissolusion process. Therefore there is no loss of uranium in this process ©.

UF$ gas in off gas @ is recovered by Nap chemical traps now, but NaF is

expensive.

So we are now going to construct a water scrubber which absorbs UFS gas in

off gas to recycle uranium.

e procesTh ) s (5 economy

e maiTh n expendeture e presenth n o st process followsa e ar s .

CD Steam for acid recovery

® Hydrofluorine

® Na3 C03 for solvent's scrubbing

@ N2 gas for fluidizing inert gas. stripr-ingr fo 1 HC . ©

6 recovery pelleF UF r Na fo t . ®

93 URANIUM CONCENTRATE

H2SCU

N32CO3

ELECTROLYTIC REDUCTION

Discharge

UF6 Fig-8 ADVANCED PNC PROCESS

Therefor s developeha C e PN economicae th d l hydrofluorin d uraniuan e m

recovery . metho® d dan whic ® h, Q reduce e th s

As shown in fig-8, this advanced process treats the waste containing F, Cl,

SO«, Na, U • etc and recovers HF and U by precipitation and dissolution and

evaporation, therefore this process doet generatno s e solith e d waste which

raise e cos y th stragesb t .

We will demonstrate this advanced process by the pilot plant in this year.

e performinar e W g another improvement s followsa s .

recycls ga 2 N e ® @ I'Fs recover watea y b yr scrubbe replaco t r trapsF Na e .

extract ion.

94 The concept ® is derived from the merits of the PNC process which has an

unique excess purification proces 4 sF precipitation U tha s i t .

Therefore amine extraction can be replaced by more rough and cheap impurity

removal method likn exchangio e e which removes onle impuritieth y s whicn ca h e removeb t no d afte d disturbsan r , after process.

In the case of UF6 convesion by the PNC process the concept of uranium

extraction should be improved by the concept of small amount of impurities

removal which leads to compact, cheap equipments and operation cost.

) C proces(6 PN Feature e th s f o s

As describing above, the merits and demerits of the PNC process are

summarized as follows.

(Merits)

® Triple purification

1) Amin extraction 2) hydrofluor mat ing precipitation

3) F2 fluormation

) (2 Eas o t controprocest ywe e th ls unti 4 precipitatioUF l n becausf o e

homogen iouw temperaturlo s e system.

@ Easy to recycle the fluoride waste.

(Demerits)

system.

95 e developinar e W e recyclinth g g syste e compac f th wasto m d an te equipments

for wet process like pulse column extractor etc, and we will overcome these

demer its in future.

4. The subjects of UFs conversion to commercialize in Japan.

[ 1 ] Needs

e commercializinTh g plan f nucleao s r fuel cycl s follows a n Japa i e ar n .

capacity : 800tli/year

start up : 1995

(2) Enrichment plant (at Shimokita)

Capacity up of 150TSWU, year by year, from 1991 to 2000.

e finaTh l capacit s 1500tSWU/yeari y .

Therefore the demand of UF6 conversion will grow large on near 1995. e problemTh domestin ] i s [2 c industrialization

There are 5 commercial converters in the world with capacity of

52,290tU/year which is surficient to supply world's demands, and conversior,

price is cheap.

As shown in Fig-9, the conversion price depends on the plant scale largely.

The world's converter with the plant scale of about 10,OOOtU/year can

afford the conversion service on the cheap price of 7 $/kgU (1.3 milliones yen

e domestith n o t c/tU> bu cas, e wit0 e demantil/year20 th h f o e dpric th , e will

e 11.b 1 $/kgU (2.0 millions yen/tUt seemi d s an )bette o dependt r s upoe th n

96 1995 2000

Year of Demand generation 3.5-

c o c 05 The Meri f Domestio t c

CD O

Q) oC 1.5- u

1000 5000 10000

PLANT SCALE [TU/YEAR]

Fig-9 CONVERSION PRICE AND PLANT SCALE

world's converter's service.

But as shown below and Fig-10 the conversion price of uranium concentrate i

is chea d comparablan p e wite transportatioth h ne interespricth r o ef o t

excess uranium preparation during UF6 transportation etc.

Terms Pr ice (million yen/tU)

(D Transportation of UF6 to Japan 0.3 (2) Transportation of reprocessed uranium from Japan 0.8

to fore ign converter's

period from conversio o enrichment n t (abou 3 monthst )

97 DOMESTIC CONVERSION FOREIGN CONVERSION

JAPAN- -JAPAN-

IIIMOKITA-

Uranium mine

Fig-10

e cas th f conversioo en O f reprocesseo n d uranium ,o supplien ther e ar en i r

the world except COMLRHEX with small capacity of about 250 Tonsil/year and with

about six times of natural uranium conversion.

On the case of domestic conversion, these expense can cut down and total

econome achievedb n ca y .

3] The policy to commercialize the U F G conversion.

The following policy is necessary to commercialize the l F B conversion in

Japan.

D C Mak a strone g combination with enrichmen t dowcu no t t

1) transportation cost

2) excess uranium preparation

98 3) excess time loss just like transportation period.

4) excess job like UF6 analysis. ) 5 exces n powema s r just like managemeut.

6) excess facility just like utility.

® Combine also with UU2 conversion of enriched uranium to cut down the same anD cost( @ ds a s

@ Dilut e reprocesseth e d uranium wite naturath h l uraniu o compensatt m e th e

following demerit se excusivowinth o t g e conversion.

1) scale demerit

2) high r -ray activity by U232 daughter

) 3 batch treatmen y heterogeneoub t s enriched uranium.

) 4 critical limitatio e enricheth y b n d uranium ove % U231 r 5

e pacon ke constructioth y B f relateo n d plane s abovea s e followinth , g

economy can be achieved, in the case of domestic 1JF6 conversion.

Case Sav i ng cost (million yens/ton U)

7 1. ) (2 Reprocessed Uranium

Therefore the establishment of the domestic conversion saves the total fuel

cycle cost in Japan. (See Fig-9). But these savin s effectivi g e domestieth onlr fo yc demand.

e followinTh g technology shoul e developeb d o construct d e economicath t l

conversion plant.

99 (D Fluorine recovery from UFG © Mixed conversio f naturao n d reprocessean l d uranium.

If the AVLIS is industrialized in future, the merit of thePNC process for

metal production wil e recognizedb l , scince almost mill t mina s e b e sitn ca f

exchange e procesth d C procesPN s e easilth s o becaust y e they have anune

extraction process already.

But the following technology should be developed to reconstruct economical

rifining plant of D metal at mine site.

CD Hydrofluorine recovery from the process, UF< to metal.

@ Economical fuel battery for electrolytic reduction.

PNC is developing the hydrofluorine recovery technololgy © now, and fuel

battery wil e economicab l n e futuremine-siti lth o s ,C procese P\ b e y ma s

hopefu n futurei l .

. 5 Conclusions

The PNC process has been demonstrated by the pilot plant.

The purity of product U F 6 satisfied the DOE specification, even with the

worst feed just like Akouta's uraniu ^centrale> c m . The uranium lose was negligible small, but uranium recovery systems like MaF

trap e beinar s g improve o economicat d l system like water scrubber.

We found that the non-sludge treatment of fluoride waste improves, the

economy of chemicals, power and steam.

e morTh e advanced proces s beini s g developed.

e mine-sitTh C procesPN e s will gain advantage e futurth n i es Laser enrichment

system.

100 The conversion of reprocessed uranium by dry conventional process has been

tested and the improved scale up equipments have been developed.

The domestic conversion plant will convert principally the reprocessed

uranium.

o construcl e economicath t l conversion syste e F conversio n L (apani me th , n

plant should be combined with enrichment, tail UF6 conversion and enriched UFC reconvesion plant, etc d hydrofluorin.an e recovering technology shoule b d

developed.

e willinar e W o contributt g e world'th e s IIFS conversion industry b y

developing the latest technology.

References

. TAKENAKAS ) . 1 NAGASAKI T ,; , "Studie r producinfo s 6 froUF g m LF«-nHn 2i 0 Japan" PRODUCTION OP YELLOW CAKE AND URANIUM FLUOR1DES (Proc. Advisory

Group Meeting Paris June 1979) IAEA, Vienna (1980) 309.

) C2 Fuel-trac, November 1977 Nuclear Assurance Corporation.

101 EXPERIENC YELLOWCAKN EI E REFININS IT D GAN CONVERSION TO URANIUM TETRAFLUORIDE AT IPEN-CNEN/SP

A. ABRAO Institute de Pesquisas Energéticas e Nucleares, Comissao Naciona e Energfd l a Nuclear, Sao Paulo, Brazil

Abstract

This paper will focus the experience acquired during the operatio a pulse f o n d columns solvent extraction pilot e purificatioplanth n i t yellowcaka f o n e produced e froth m industrial treatment of monazite sand. Special care was devo- e rarth e teo t earthd s elements, thoriu d zirconiuan m m decon. tamination. Intermediate produc n uraniua s i t m trioxidb o e taine y dewatterinb d d thermaan g l decompositio f o ndiuranat e and its con- arsion to uranium tetrafluoride. The experience developed and the establishment of the quality control procedures to follow up all steps on both pilot unities as an important support to the technical work is emphasised.

1. INTRODUCTION

s contributioA nationae th o nt l progra r develomfo p ping atomic energ r peacefufo y l uses, e headeth y b d Brazilian Nuclear Energy Commission (CNEN) e ,th Institute de Pesquisas Energéticas e Nucleares (IPEN), S.Paulo, has give a ngrea t dea f efforo l t concernin ga systemati c develo£ men f researce o testablishmen th r fo he technologth f o t f o y uraniu d thoriuman m e prograTh . vers i m y much dedicateo t d the educatio d traininan n f chemisto g d engineersan s o t d ,an the productio f somo n e nuclear materia r furthefo l r metal_ lurgic work and fabrication of fuel elements for research reactors. In this paper we summarize the main activities on the purification of uranium raw concentrates and their conve_r_ sion into nuclear grade compounds e desigTh . d assemblagan n e of pilot facilities for pure ammonium diuranate (ADU), ur£ nium tetrafluoride and uranyl nitrate (UN) and its further denitration to trioxide are discussed. e developmenTh d adaptatioan t f analyticao n e c o pr l dure d theian s r applicabilit n importana s a y t e supporth o t t technical work and the quality control of the abovementioned nuclear grade materials is emphasied as well.

103 2. THE FIRST YELLOWCAKE

e firsTh t yellowcake tha wore tw k with since several yeara sodiu s i s m diuranate (SOU) produced froe industriath m l processin monazitf o g e sand [1] e chemica.Th l treatmenr fo t breaking up monazite sand by alkaline process has been in pra£ tic Brazin i e l (S.Paulo) since 194n industriaa 8n o l scale [2]. The production capacity is about 3000 metric tons of monazite per year for the production of thorium, rare earth chlorides (2000 tons) and phosphate as the main products. After decontami_ nated from radium and its descendents by coprecipitation with barium sulphate the rare earth chlorides are commercialized. Thorium is stocked mainly as a crude hydroxide (thorium slud ge). Uranium is recovered as a by-produtct in the form of sodium diuranate. The main impurities considered in this yellow cake are sodium, phosphate, silica, iron and f courseo , , d thoriuan m rare earth elements (RE). Great concern was given to the decon tamination of thorium and rare earths. In Table I is presented a representative compositio f thio n s SOU,

TABL CHEMICA- I E L COMPOSITIO SODIUF NO M DIURANATE

ELEMENT %

Us 0a U0 Q 79.5 B 3 8 0.0002 Cu 0.001 V 0.004 Mo 0.0005 As 0.01 P as PO, 0.3 S as SO, 1.5 _ 4 F 0.02 Halogens 0.015 ThOs a „h T 3.0* Rare Earths 0.2 Sm + Eu + Gd + Dy 0.02 max. Fe 0.1 Cd 0.007 Pb 0.0015 Ti 0.0015 s SSiOa i „ 1.4 Na as Na^O 9.2 * Variable from 0.3 to 8.0%

As the abovementioned SDU at the monazite plant is only dewatered at 110-120°C and usually contains some organic mater, as a not controlled impurity, before its dissolution the yellowcake is calcined at 450°C during two hours. This trea_t men s introducetwa e flow-sheeth n i d o avoit t d evolutiof o n X gasesNO workee .W d with this uranium concentrat r severafo e l years, as it was the unique raw material at hand.

104 3. A NEW YELLOWCAKE

Recentl a secony d yellowcake coul e usedb d t I . came from the Poços de Caldas Industrial Complex, at Poços de Caldas, Minas Gérais State. This industrial plan ownes i td operate an d d y NUCLEBRASb e yellowcakTh . n ammoniua s i e m diuranat f o ever y good quality. The unique difficulty we had to cope with is the presenc f zirconiuo e m contamination. Afte s dissolutioit r n with nitric acid, the clear uranyl nitrate solution was treated for the removal of the great majority of zirconium. It is clear that this yellowcak ee calcined b coul t no d , otherwis t i ewil l be convert into uranium oxide which solubilization will gen_e rate high evolution of NOX gases and the installation is not ready to absorve them.

4. DISSOLUTION OF YELLOWCAKE

e initiaTh e dissolutiol th ste s e concentrati p th f o n e with nitric acid for the obtaintion of a clear uranyl nitrate. Durin e dissolutioth g e grosth n s amoun f silico t s removei a y b d dehidratio f siliciouo n s acid e dissolutioTh . s i accomplishen d into a stainless steel reactor of 300 L capacity in a batchwise fashion. The yellowcake is poured direct and slowly into the nitric acid e yello. th Afte l wal r caks introduceewa diges it d s tio mads i n e wit HNO.M 2 h t 90-100°a e complet, th r Cfo e flocullï tion of silica. If zirconium is present, the uranyl nitrate is treated with controlled amount of phosphoric acid. Both silica and zi_r conyl phosphat e separatear e d together t pul ho s i pe filt_Th . e red into a canvas filter and the residue thoroughly washed for e removath f solublo l e uranyl nitrate e filtereTh . d uranyi n_ l träte d aftesolutioan a concentratiorL s U/ ha ng coc 5 > 47 f o n led is adjusted to 200 g U/L and IM HNO,, previously to the so_l vent extraction. Sodium or ammonium nitrate, not less than IM used as salting cut agent is formed during the dissolution. In the case of the SDU produced from the monazite sand the trouble some presence of Th and RE is minimized by the controlled add^ tion of sodium sulfate [3,4].

5. TSF EXTRACTION PILOT PLANT FOR PURIFICATION OF URANIUM

A pilot e purificatioplanth r t fo facilit p u f uranyo t n se y l nitrat e conventiona s baseth i e n o d l liquid-liquid extraction technique using three pulsed e extractioncolumnth r fo s , scrub- bind strippingan g , respectively e facilits Th .it equi d an yp ment, operational flow-sheet, performanc d gainean e d experience were published [5,6] e facilit.Th y comprisee a ssectioth r fo n safety opening of the drums and for the dissolution of the yellowcake. The organic phase is a (v/v) 35% TBP-varsol. The three columns have perforated plates of about 23% area. The e_x traction is accomplished in coutercurrent using an organic to aqueous ration of 2.2 to 1. The loaded organic phase containing L leaveU/ 13 e extractiog 5 th s e n th s admitecolumi d o an nt d scrubbing column ns scrubbe i wher t i e d with 0.2n Ma HNO n -i organic/aqueous phases ratio of 1:1. The washed organic phase

105 (110-11 g U/L5 )s strippei d with water usin d aqueous/organian g c , resultin1 o t 6 n uranyratia g1. f o ol nitrate solutiof o n 70-105 g U/L. The third column can be steam heated and operated at 40-60°C resulting in an uranyl solution of mean 100 g U/L va lue. This solution is filtered through a celite layer for the coalescence os small droplets of TBP and then forced into a layer of pure diluent to remove the last traces of TBP,

6. PRECIPITATIO F AMMONIUO N M DIURANATE (ADU)

e piloTh t plan s equippei t o perfort d e precipitatioth m n i n a batchwise way L reactor ,s continuou0 a int50 d a oan , s opera tio wellas n purThe . e uranyl nitrate solution (100 U/Lg is ) heated to about 60°C and the ADU is obtained by bubbling _ luted anhydrou e controlleE b N sgas_ n va e finaca Th . H o p lt d lues raging from 4,0 to 7,5 for ADU to be send to the UF7 unity H aboue p s directe c_ i o 0 (excest 1 tU r AD o o t e ds th NH_ f i ), ramie grade U0„ pellets. For the batchwise precipitation the ADU is dewatered int a vacuuo m canva n umidita ss n filteha ra yd an r ging from 45 to 50%. In the continuous precipitation using one step reaction the reactor (027 cm x 100 cm, 56L) is fed with uranyl nitrate at a 2-2. ~ bubble1 rat . NH f OL/mina rac o e d t ea an d , ,C heate0 5 t a d of 60-80 L/min s usualle finai Th .H p l y abou 0 (exces1 t s NH_, ) s depositei e e bottoe U continuouslurrth th Th AD f n f o i mdo y s filter and is sucked by Vacuum into the rotating drum. The cake leave e druth s m with abou % umidity50 t e e thicknesth th , f o s layer. mm rangin 5 3. o gt fro5 2. m

7. TR10XIDE FACILITY o curr.prise UO A facilit conversioe o t th U r a s AD fo y f o n continuous, electrically heatedi d ,d belfe s ti furnaceU AD e Th . rectly froe rotatinth m g e stainlesfilteth o t r s steel conveyor belt moving insid e furnacth e e with zones having different tempe ratures, the gradient ranging from 110° to 500°C. The final pro duct is an UO-^ oxide used as feeding material for the tetrafluo ride plant or sent to further conversion to ceramic grade diox^i de. The residence time during the dewatering is 2-3 hours.

8. TREATMENT OF AQUEOUS EFFLUENTS The uranium purification pilot plant gives rise to some solutions containing uranium, thorium, rare earthsi t , irod an n tanium maie th ,n stream coming froe extractioth m n columnl Al . effluents are collected and treated with sodium hydroxide. The precipitate is filtered out and returned to the dissolution se_£ tion with nitric acid and sent to the extraction again.

9. TETRAFLUORIDE PILOT PLANT After some previous work [7,8 a ]pilo t plant facilito yt acquir necessare th e y ^ productiotechnologUF e th futher n fo o ny r uses in the reduction to U metal and preparation of UFg, was set up [9] e establishmen.Th f thio t e collaboratios th unit d ha y d nan technical assistance from the International Atomic Energy Agency (IAEA) .

106 The starting material is UO-^, reduced to UC>2 by cracked NHg and the use of anhydrous hydrogenfluoride for the conversion to UF^. During some operatio f thio n s pilob o e procesth t f o s taintion of UO^ was changed in the sense that the ADU precipitate at pH 7.5 for ceramic grade was calcined to VO^ in the contji nuous belt furnace and then pellotized in spheres of about 4-6 , driemm d agai d use s feean na d d L reactormateria e th n i .l This o exhibete w UO typ ne f o e d excellent mechanical properties, Nowa day e changew s f directlo d e onc us o produce e UO y th mor o s t a ed flakes of 3-4 mm thickness by filtration of ADU in the vacuum ro_ tating filter and dried and calcined in the moving belt conve yor. This type of UO-^ exhibeted excellent mechanical properties. This trioxide is contacted with anhydrous hydrogen fluoride for the conversion to UF^. The green salt produced is of good qua. lity, assessing a minimum of 95% UF^, ~~

10- WET WAY UF, PREPARATION

A bench scale facility for preparation of uranium tetr_a fluorid a aqueouvi e s conversio 2 powdeUÛ f o rn with hydrofluoric acid is under experimental test. The dioxide is obtained in the y movindr d reactobe g r thas bee ha n operatioti n e th n alsr fo o UO? production as well. The UOj is reduced again by hydrogen g_e nerate y ammonib d a cracking A .secon 2 typ UÛ s producedi e y b d direct reduction of ADU in a moving belt furnace with hydrogen. e conversioTh f boto n h_ wit U0 typ hf o ehydrofluori c aci s quiti d e simpl d effectivean e e tetr?.fluoridth , e being fi_l_ tered usin vacuua g m canvas filte d driean r d stepwise, first a t 110-120° d the an Ct highe a n r temperature e qualitTh . f o y this green salt for uranium reduction or hexafluoride preparation is under investigation e firsth , t results being promissor.

11. DENITRATIO F URANYO N L NITRATE

Development studie a fluidize on s procesbed d confor s - version of uranyl nitrate solution to uranium trioxide and recovery of nitric acid is under investigation. A pilot plant is being set up comprising an unity for concentration of uranyl ni_ träte solution from 100 g U/L to about 900 g U/L, followed by denitration of the melteded uranyl nitrate to UO.,. This new uni_ e firsth ts tha tesy t schedule r seconfo d d semester 1986,

12. QUALITY CONTROL

Mention wil made r somb l fo ee procedures speciall^ d y veloped to assist the pilot plant work. A rapid routine determl nation of uranium content in uranyl solution is done by gamma- ray spectrometry [10] v photopeak, usinKe 2£ e ^55 pr th gu18 .A cedur s outlinedirece wa eth r t fo ddeterminatio U conten f o n f o t uranyl nitrate-TBP-organic phase [11] e thermogravimetriTh . s bj c havior of ADU samples, and specially the pyrophoricity grade of 2 powderUÛ d thei an sU rati 0/ r UC^n i o - powde d pelletan rs wa s developed [12] e analytica.Th _ se s lmadwa controy , b e UF f o l quential analysis of the most probable products existing with the tetrafluoride [13]. The determination of microquantities of

107 B in highly pure uranium and thorium compounds is done through the extraction of the colored complex of BF. -monomethyl tnionine [14,15]. 4 -* t°C

660

970 .1000

<0 490 610 730620 90O 0 *OO96 G

340

170

THERMOGRAVIMETRIC CURVE F URANIUO S M DIOXIDE [12]

) U0,(A (normal) obtaine y reductiob d f o n ;

770°Ct a , H .U 3n Ui Q

(B) DO (pyrophoric) obtained by direct reduction of ADU in II- at 770 C.

A procedur e separatioth r fo ed concentratioan n x e f o n tremely low amounts of thorium and rare earths from uranyl so lution s developpeswa d based upo e sorptioth n f thoso n e elements from solutions containing 0,3M HF into a small column of alumi a [16]n . Using this techniqu individuae th e E havR l e a beean n lysed by emission spectrography [17]. A semiquantitative rout_i_ e spectrometrin c metho s outlinee direcwa dth r t fo d determine^ element8 1 tio f o n n uraniui s m compounds, including UF,., using gallium oxide and sodium fluoride. Vanishing small amounts of RE in uranium are determined by fluorescende spectrometry after separation int aluminn a o a column [18], Zirconiu s i m analysed by direct spectrof luor:'metric determinatio n uranyi n l chloride using morin [19 d spectrophotometricallan ] y with chloroanilic acid [20].

108 Procedure e determinatioth r fo se compositio th f o n f o n the cell electrolyte were developed based on the alkalimetric determination of HF and the total determination of free hydro fluoric acid liberated after percolation on a strong cationic ion-exchange resin, H-form, and on the determination of melting point of the mixture. The presence of residual HF in uranium hexafluoride is determined after its hydrolysis and measurement of total uraniu d totaan m l hydrofluoric acid.

REFERENCES

[1] BRIL & KRUMHOLZ P. Producäo de öxido de tôrio nucleamiente puro. Sao Paulo, Institute de Energia Atômica, dez. 1965. (IEA-115). [2] KRUMHOLZ, P. and GOTTENKER, F. The extraction of thorium ' and uranium from monazite. In: United Nations, New York, Proceedings of the international conference on the Pe_a ceful Uses of Atomic Energy, held in Geneva 8 August 20 August 1955, 4-8: Production technology of the material ' used for nuclear energy. New York, 1956, p.126-8. [3] BRIL, K.J. & KRUMHOLZ, P. Production of nuclearly pure ura niu me decontaminatio studth n o y f uraniuno m from thorium and rare earth y extractiob s n with tributylphosphat: in e INTERAMERICAN NUCLEAR ENERGY COMMISSION. Proceedings of the 3rd Interamerican symposium on the peaceful applica tions of nuclear energy, Rio de Janeiro, 1960. Washington, D.C., Pan American Union, 1961. p.37-59. ] BRI[4 KRUMHOLZ& L processn Ur , ,P o industria produçâe d l e d o urânio nuclearraente puro o PauloSä . , ORQUIMA, Lab. Pesqui- sas, 1960. (LPO-9). ] FRANÇ[5 A JR., J.M. Usina pilot purificaçàe d o urânie d o o pelo processo de colunas pulsadas em operaçao no Institute de Energia Atômica o PauloSâ , , Institut e Energid e a Atômica, out. 1972, (IEA-277). [6] FRANÇA JR., J.M. & MESSANO, J. Dimensionamento de colunas ' pulsadas industrials na purificaçào de urânio para fins nucleares, pelo U metodindiretoHT o d o o PauloSâ . , Institu_ to de Energia Atômica, maio 1974, (IEA-343). [7] CUSSIOi, FILHO, A. & ABRÂO, A, Tecnologia para a preparaçlo de tetrafluoret e urlnid o r fluoridretaçapo o „ obtiU0 e -d o do de diuranato de amonio. Sâo Paulo, Institute de Energia Atômica, ja. 1975. (IEA-379). ] RIBAS[8 , A.G.S ABRÂOd an . . Preparaçâ,A „ apropriadU0 e d o . pa o a obtençâr e UF.d o. S.Paulo, Institut Energie d e a Atômica, nov. 1973 (IEA-318). [91 FRANÇ . J.M.JR A Unidade pilot e tetrafluoretd o urânie d o ' o pelo process e leitd o o môve m operaçae l o Paulo o Sä IEAn o , , Institut Energie d e a Atômica, jan. 1975. (IEA-381).

109 [10] ABRÂO, A. & TAMURA, H. Routine radiometric determination of uraniu y gamina-rab m y spectrometry o PauloSa . , Institute d o Energia Atômica, ago. 1968. (IEA-170). [11] FEDERGRUN ABRÂO& . , L . Determinaçâ,A o espectrofotométri_ ça direta de urlnio na fase orgânica fosfato de n-tribu tilo-nitrato de uranilo. Sao Paulo, Instituto de Energia AtÔmica, jul. 1971. (IEA-242). [12] ABRÂO, A. Thermogravimetric behavior of some uranium coin pounds;applicatio U rati0/ o ot n determination o PauloSa , , Instituto de Energia AtÔmica, ago, 1965. (IEA-105), [13] FEDERGRUN, L. & ABRÂO, A. Determinaçâo dos conteûdos de m tetrafluorete ^ UF e d e e urânio d o2 o UO„FU0 Sa e .d 2 Paulo, Instituto Energie d a AtÔmica, maio 1974.(IEA-341). [14] FEDERGRUN ABRÂO& . ,L Determinaçâ. ,A o espectrofotométri_ ça de boro em sulfato de tôrio. Sao Paulo, Instituto de Energia Atômica, jun. 1976. (IEA-420), [15] FEDERGRUN ABRÂO& . ,L . Determinaçâ ,A o espectrofotométri_ ça de boro em urânio, alumînio e magnesio: extraçao de tetrafluoreto de monometiltionina. Sao Paulo, Instituto de Energia AtÔmica, jun. 1968. (IEA-165). [16] ABRÂO . Chromatographi,A e separatio d concentratioan n n of thoriu d raran m e earths from uranium using alumina- hydrofluoric acid. Preparation of carrier-free radio_ thorium and contribution to the fission rare earths.Sao Paulo, Institut Energie d o a Atômica, jun, 1970.(1EA-217). [17] LORDELLO, A.R. Determinaçâo espectroquimica dos elementos lantanldeo m compostoe s e urâniod s a separaça,vi o croma tografic m colune a e alumina-acidd a o fluoridricoo Sa . Paulo, 1972. [Master Thesis], [18] CAZOTTI, R.I .ABRÄO& . Spectrofluorimetri,A c determina- « tio f raro n e earth U afte n i s r separatio d concentran n a tion of total lanthanides onto an alumina column, Sao Paulo, Institut Energie d o a Atômica, jun, 1973,(IEA-295), [19] CAZOTTI, R.I .t aliie . Determinaçâo espectrofluorimétri- ca direta de microquantidades de zircônio em urlnio.Sao Paulo, Institut Energie d o a Atômica, fev.1976.(IEA-401). [20] FLOHt aliie . .,B Separaça e zircônid o r extraçapo o m e o meio cloridrico corn tri-n-octilamin a determinaçasu e a o espectrofotométrica corn âcido cloroanîlico o PauloSa . , Institut Energie d o a AtÔmica, ago.1976. (IEA-427).

110 CONVERSION OF NON-NUCLEAR GRADE FEEDSTOCK TO UF4

A.A. PONELIS, M.N. SLABBER, C.H.E. ZIMMER Atomic Energy Corporatio f Soutno h Africa Ltd, Pretoria, South Africa

Abstract e SoutTh h African Conversion rout e direc s baseth i e n o td fee f ammoniuo d m di-uranat a numbe f o f o e re on produce y an y b d different mines. The physical and chemical characteristics of e feedstocth n thuca k s vary considerabl d influencan y e th e conversion e finarat th V s welproduca eUF ls a l t purity.

, conversioUF e Th n reacto a Movin s d i Reactor Be g r (MBR) with countercurren e reactint th flo f s o phasesw ga g .

Initial problem o continuouslt s y R operatwerMB ee th e mostly concerned with the physical nature of the U0„ feed particles. Different approaches to eventually obtain a suc- e discussedcessfuar R MB l . Besides obtainin L feeUC g d particles with certain physical attributes e chemicath , l impurities also have an effect on the operability of the MBR.

e influencTh e feedstocth f o e k variable e reductioth n o s n d hydrofluorinatioan n rates after calcinin s largelha g y been determined from laboratory and pilot studies.

The effect of chemical impurities such as sodium and potassium on the sinterability of the reacting particles and therefore the optimum temperature range in the MBR is also discussed.

Confirmatio e effecth f sodiuf o to n d potassiuan m m impurities on the conversion rate has been obtained from large scale reactor operation.

11 1. INTRODUCTION

1.1 General

The South African process is based on the conversion to e sulphatth f o U produce minese AD , th UF y ,b d followea y b d UF, flame reactor and subsequent purification of the UF,- by distillation.

n thiI s pape histore conversioe th rth f o y n step s trai s - ced, the present experiences and development regarding the impure feedstoc discusses i ke futur th d e an ddevelopmen t outlined.

1.2 The four nuclear sectors

Four independent groups are responsible for the various aspects of nuclear material production.

These four are:

i) the gold producing mines that produce sulphate by-producta ADs Ua ,

ii) the Nuclear Fuels Corporation of SA (NUFCOR) which produces Uranium Oxide Concentrate r expor fo sd als an to handles the transport of ADU slurry to the Atomic Energy Cor- poratio Limite, SA f o nd (AEC),

C whicd the- AE an en n e h , th convertUF iii o t ) U AD s riches this materiald ,an

iv) the Electricity Supply Commission (ESCOM) which generates electric power and also operates the Koeberg Nuclear Power Station.

112 w commentfe A s regarding each activity will clarife th y AEG's responsibility.

uraniu7 1 o Ut p m extraction plant e operatesar - va y b d rious gold mines. The processing of the uranyl sulphate and subsequen U precipitatioAD t d filtratioan n s dependeni n n o t the specific mine's requirements [1].

The clearing house of ADU is NUFCOR and the NUFCOR plant s produceha d uranium oxide concentrate r exporfo s t since 1953. A wealt f experienco h s beeha e n t builthia p su c plant, which s mad ha a significane t contributio C conversioAE e th o t n plant.

1.3 The AEC operations

s sitei t ValindabC a d AE e d produceTh an a s enriched uranium A pilo. t plan s beeha t n user initiafo d , producUF l - tion and development work but this will be shut down during 1986.

A bigge , conversioUF r e production th plan r fo t f o n C enrichmenAE e th suitablr tfo plan , s beeUF eha t n build an t wil commissionee b l d during 1986.

The raw material inputs to the plant are ADU from the

mines, CaF9 and sulphuric acid for HF and the subsequent

F2 production. The ADU ^o U03> the U03 to UF4 the UF, to UF, followed by UF, distillation steps then pro- enrichmene ducth a suitabler fo , t UF eplant . (See Figures . 2) d an 1

113 Fig. l Schematic representatio, productionUF o t U AD ,f o n

UF6 CRYSTALLIZE RS PRIMARY SFCl JDARY

— -fsCRUPETR I

FLUORIC CELLS

Fig 2 . Schematic representatio, productionUF o t , UF f .o n

114 This paper emphasizes the ADU to UF, process steps, although the restrictions or requirements set by the ADU propertie r UF.so , specification e alsar s o mentioned.

2. UF4 CONVERSION HISTORY

1 , 2 Initial objectives

The AEG utilizes a moving bed reactor (MBR) for the con- e firso UF,t th versio tn . I U0 sectioe reac f th o n -f o n 0 3 4 tor UCL is reduced to UCL by means of NH„ and in the s i hydrofluorinate L seconUC . e meany HF b th d f , o s UF o t d d solidan s sga flo e wTh countercurrentl e reactorth n i y .

The following description of how the process was develope de non-nuclea e lighmusth th seee f b to tn ni rd an grad U AD e thus the impure UCL used as feedstock. It was known that successful commercial conversio d beeha nn obtained R witMB a h but with nitrate base d nucleaan d r puret feedstockno s wa t I . always clear during development work whether adverse effects such as sintering or unreactivity were caused by certain impuritiee sulphat th , etc. y Ca b s , K )er (sa o , contentyNa .

t firsa s tI wa t thought thae initiath t l purification steps of ADU were expensive and that simple UF,- distillation at the tail-end was more cost effective. Environmental factors relating to effluent treatment of a possible TBP plant for ADU purification also influenced the initial decision to opt for tail-end purification. However, as the cost of the UF,- is high, any processing steps must minimise UF, losses.

In order to obtain UF,- suitable for enrichment at the o G distillatioAE s utilizei n o reduct d e volatilth e e fluorides such as MoF,,. o

115 2 2. Process development

e SoutTh h Africa , conversioUF n n backgroun arbin ca d - trarily be divided into four periods.

2.2.1 The first period (- December 1972)

NUFCQR originally carrie t feasibilitou d y studie- de o t s termine whether South African material coul e converteb d a n i d MBR. During these initial studies emphasis was placed on the ADU preparation conditions such as precipitation pH, precipitation temperature arid precipitatio- n al rate e s welth a , s la lowable sulphate content. A limit of 4 % sulphate in the ADU was specified and ADU preparation conditions were established. Since the ADU precipitation particle size determines the suc- ces f e conversiospecificatioso th , UF o t n n requires that particlee th f o greatee % b s 0 5 r than abou t. Preci5-1ym 0 - pitation condition minee th e therefor sar t a s s followsa e a : temperature between 30 and 40 °C, a uranium concentration in the mother liquor of between 7 and 9 g/£, a residence time of about 17 to 24 minutes and a pH between 7,0 and 7,4.

Furthermor e filterabilitth e y prio o dryint r d calcian g - ning is important and the ADU slurry density is specified to 3 be between 1300 and 1450 kg/m at delivery.

Additional studies to correlate the ADU precipitation particle size, specific surfac ~ pow U0„e U0 - th ,are f o a der bulk density and operational conditions such as temperature were unsuccessful.

Howeve r s successfullabouwa 0 tonne, 2 tUF f so y produced by NUFCOR and it was decided to proceed with this process

116 2.2.2 The second period (1973 - December 1982)

During this perio a pilod , reactoUF t s constructewa r d and commissione e AEGth .t a dEmpirica l studie o correlatt s e thermogravimetric analysis with reactor operation were unsuccessful.

The feed U0„ was material as produced by the drier- calciner. Various configurations to the ADD extruder were made in an attempt to produce acceptable U0~ feed particles.

Also various mechanical devices were introduce. d UF inte th o 4 d movemene mechanicae effecbe th th reacto d f t o tai bu t o t rl o caust aid s e swa powderin e UCL/UCth f o gL particles with subsequent unpredictabilit d movementbe f o y .

Although the reactor runs were unsuccessful, valuable experience was gained in the operation of drier-calciner during this period.

2.2.3 The third period (1983 - July 1985)

e beginninAth t s apparen f wa 198o gt 3i , tUF thae th t reactor required ~ feedstocU0 a bette f o k r physical (mechanical) quality.

A compactor (pelletizer) was acquired to obtain UCL feed particles with a predictable sieve analysis.

Table I shows a typical sieve analysis. It can be seen that about 90 % of the UCL feed has particles greater than 4,75 mm.

Also, all mechanical aids to bed movement were removed e reactoanth d r converte a conventiona o t d l MBR.

17 TABLE I

Typical 00,, feed ,,article sieve analysis for UF, reactor

Size (mm) >9,5 5 9, 6, - 7 4,75 - 6,7 - 4,73,35 2,36 - 3,36

Mass (%) 20 48 23 7 2

In additio e acquisitioth o e t compactorn th f o n a ,detai - led semi-empirical laboratory stud s undertakewa y o investit n - gate the effect of sodium and potassium on the sintering characteristics of the UCL - UCL - UF, particles.

During the pilot reactor runs from April 1985 to K Juld an y e compacte a N 198th e 5 th ~ fee s useU0 d wa dan d content checked n additionI . e previousl, th som f o e y unsuccessful operating conditions were changed reductioThe . n temperature was kept to about 700 °C and the hydrofluorination temperature profile reactoth n i e r limitet a C ° o about d0 60 t th es t introduced wa no exi F s H t wa % where F • ; " -T 0 .10 diluted with nitrogen, as had previously been the case to counteract the high reactivity of the sulphate material.

TABLI EI

Summary of UF, pilot reactor runs for April to July 1985

Total runtim shutdowo t e - 156n 2 hours

% 0 8 Operatin - g time

Downtime - 20 % 3 tonne6 - sUF, produced

80 - 90 60 - 70 50 - 60 . contenUF 4 t >90 70-80 40 - 50 <40

Amount of material (%) 26 39 15

118 Typical results of these runs are presented in Table II. From this table it can be seen that about 70 % of the material had a conversion of greater than 70 % to UF,. Additional discussion follow n Sectioi s . 4 n

2.2.4 The fourth period (August 1985 -)

e piloTh t reactor runs establishe e requireth d L UC d particle size distributio e correcth d tan n reactor operating procedures. Together with knowledge obtained from empirical laboratory studie n "goodo s U feedstoc"AD commerciae th k l conversion plant reactor coul e re-evaluateb d d modifiedan d .

Whereas the pilot reactor had a throughput of about 5 kg/4 h (U0~ productioe )th n reactor's throughpus i t approximately 175 kg/h. Typical residence times are 11 and 7 hours respectively.

e reactoTh s constructei r o materialstw f o de reduc th : - tion section of Stainless Steel 309 and the hydrofluorination section of Monel 400.

The reactor is equipped with vibrators that are intermittently switche o facilitatt n o d d movementbe e .

e productioTh n reacto s successfullwa r y commissioned an d has produced UF, similar in quality to the pilot reactor.

3. RESEARCH AND DEVELOPMENT

o maitw Thern e o operatrequirementt ar e R MB e a suc r fo s- cessfull r hig- yieldsfo yUF h .

119 Firstl e soli d musth y be dt move consistently without blocking or channelling. To satisfy this requirement the feed UOo particles must comply with the specified sieve analysis.

Secondly the individual particle must be porous, reac- tive, have low sinterability, have a wide operating temperature rang d must powdean eno t r easily.

The ADU feed and drier calciner operating conditions determine the feed characteristics of the U0„ compactor. The UCL compactor operation determines to a large extent the particle characteristics.

Laboratory investigations were principallye aimeth t a d

ADU calcining conditions to obtain U0 with low IL0 and

0 + 0 J J 8 NH, content. In addition, detailed investigations were carried out into the effects of the impurities Na and K on the reactivity and consequently on the temperature range which the UCL-UCL-UF, particle could withstand. Reduced reactivity can be caused by a reduced specific surface area (sintering by forming of a low melting point NaF eutectic or collapsing of pore structure y formatiob r )o f sodiuo n r potassiuo m m fluorides on which the product gases such as water adsorp more strongly, preventin highea g r reaction rate.

e relativTh e contribution f theso s e factor o reducet s d reactivity has not yet been established and thus cannot be quantifie t presenta d .

3.1 U0j0 productio—————————n

3.1.1 Plant operation

In orde obtaio ammoniut d r an Uw 0 nU0 wit lo a hm

0 0 0 O JO content the operating conditions presented in Table III are normally used.

120 TABLE III

Typical drier-calciner operational conditions

_ productioU0 n rate 160kg/h

r inleai Driet- temperaturr e 290 °C r outleai t temperature 65 °C residence time 1 hour D thicknesAD beln so t 30 mm r flowratai e 4 ton/2, h

Calcine r inleai t- r temperature 410 °C air outlet temperature 290 °C residence time 1 hour L thicknesUC n belso t 30 mm air flowrate ton/4 2, h

Typical ADU properties from 6 mines are given in Table e resultan IVproperties Th .ha 3 U0 t shows a s Tabln ni . V e From this table it can be seen that the U0„ feed particle typicall abouporosita f wit% o s 0 specifia ) h6 tha y e ( y c surface area (SSA. /g abouf m o ) 0 2 t 2 These tables forbasie th mf furthe o s r discussions.

TABLE IV

ADU Propertie materialf so s investigated

Material pH Moisture SG Dp50 Total SO^ Total U

Mine (%) (um) (%) (%)

1 7,6 35,0 1,30 9,6 3,74 71

2 7,5 34,4 1,40 8,6 2,60 73

3 7,0 27,8 1,28 9,2 1,93 74

4 7,6 40,2 1,37 6,4 0,81 73

5 2,0

6 4,0

121 TABLEV

, PropertieDO f materialo s s investigated

Material Total U soj- Na K Na+K 30 3 SSA U3°8 V " CP Mine (%) (%) (%) (ug/gU) (Ug/gU) (Ug/gU) (kg/m3) (n>2/g) 1 80 3,10 2, A 30 11 Al 2,01 0,67 18,3

2 80 2, A3 1,1 72 1AO 212 2,10 0,57 19,0 3 81 1,97 6,1 151 356 507 2,05 0,66 17,3

4 81 1,04 A, A 256 558 81A 2,06 0,66 18,3

5 79 1,26 3,7 1200 177 1277 1,99 0,53 18,7 6 78 0,59 6,5 220 7066 7286 2,08 0,65 9,2

7 81 0 8, A 97 168 265 2,08 0,63 17,7

8 82 0,5 5,6 523 328 851 2,21 0,57 20,8 i

3.1.2 Thermal decomposition of U00

Sinc e thermath e l histor f feedstoco y L determineUC k s the amount of other oxides, and therefore reactivity characteristics? it is necessary

) durin(a g plant operatio keeo t n p operating conditions closely controlled, and

(b) for investigation purposes to have repeatable characteristics.

A typical mine-produced material (Min Tabl, 8 e wit) V e h medium-high Na+K was thermally decomposed in air for periods of one hour to obtain equilibrium with atmospheric oxygen.

Figure 3 (curve A) shows the fourfold reduction in specific surface area (SSA) between 500 and 700 °C. The accompanying curv B eshow oxygee sth uraniuo t n m ratio (0/U) change. At an 0/U ratio of about 2,7 a phase change occurs and the SSA change coincides with this change.

122 3O _

O/U SSA RATIO (m2 /g)

5 -

O 5O 100 0 30 700 9OO TEMPERATURE i°o

Fig3 . Thermal decompositio UO,f o nn air i ^ .

A few comments pertaining to curve A are relevant.

It is known that the reduction rate is directly propor- feedstoce tionath o t lA [2]SS k . Therefore feedstocL UC k which has been heated to about 700 °C prior to reduction will have substantially lower reactivity.

Further, if rapid reduction occurs on the outer crust of large non-porous particles in the reactor, inner particle temperatures will rise, leading to decreased SSA with subsequent more difficult reduction at later stages in the reduction zone.

The loss of SSA is an indication of sintering and is in- s compositiodependenga a temperatur e s i th t f o tbu n e sensi- tive phenomenon.

123 Preliminary tests in a nitrogen atmosphere indicate that A firsthSS e t increase beforC ° 0 se 50 betwee decrea- 0 40 n - sin t highea g r temperatures.

Similar curve e beinsar g material r drawfo p u n s containing different quantitie f Na+K o sn iner a alsd - n ,an at i to mosphere.

2 . 3 Particle reactivity investigation (Na-fK)

s knowi t nI that sodiu d potassiuan m m have deleterious effects during reductio hydrofluorinatiod an ] [3 n n [4]- .La boratory studie e beins ar wer d g an econducte o levelstw n o d. Firstly empirical tests were developed to check material for possible plant conversion, primaril o determint y e effectth e s of different Na+K contents.

Since "good material" conversion characteristice ar s known both from the plant and the laboratory, the results of these tests are then used for comparative rather than quanti- tive purposes.

Secondly, reaction rate ki'netic studie e beinsar g conducted with both thermogravimetric analysis and small static beds n ordei o obtait r n rate equation r developmenfo s a reacto f o t r model e reactoTh . r modellin s basei g n solvino d e reactioth g n kinetics differential equations for the particles without sintering effects (Appendix). Later studies will include the sintering effect followed by full modelling of the entire reactor.

3.2.1 Stati testd be c methos- d

In order to obtain comparative results on widely diffe- ring materials the following empirical method was developed for laboratory use.

124 A sample of sieved U0~ with a mesh size between 2,36 and s obtaine3,3i m 6m d froe planth m t compacted material. The sample of about 5 g is spread as a single layer of parti- a gricle n do s contained withi a ntubee tub s Th heatei .e n i d n ovea o obtait n e requireth n d bulk reaction temperature.

For all the reduction and hydrofluorination tests the UO,, is first converted to lUCL in a nitrogen atmosphere at 650 °C for 60 minutes. The sample is then taken to the reduction temperatur ~ reductioNH r fo e n followee th y b d hydrofluorination temperatur r hydrofluorinatinfo e g wit. HF h

3.2.1.1 Reduction tests

After the sample is taken to the test reduction temperature from the 650 °C level, it is reduced with NHL for only 6 minutes. Since the determination of the UO^ content is difficult to establish directly the sample is then hydrofluo- a perio0 minute 2 r f fo o d ensuro C t s° rinate0 e55 comt a d - plete hydrofluorination of all UCL. The amount of UF, formed is then an indication of the UCL obtained during reduction.

This specifia tes r s carriei tfo t c ou dmateria e th t a l following reduction temperatures: 600, 620,d an 640 0 66 , 680 °C.

This series of tests is then repeated for the other materials.

e resultreductioe Th th f o s n test e presentesar n i d Figure A.

125 ICO

% UR

40 MATERIAL Na. K (jjq/gll) TEST CONDITIONS .— _ — 1 41 REDUCTION TIME 6mlns .—— — 2 212 HYDROFLUORINATION — — 3 507 TIME 20mins ———— 4 814 HYDROFLUORINATION 55O*C ———— 5 I177 TEMP GPAJ4ULE SIZE «2,36mm 6 7286 -3,36mm — —— 7 265

600 620 640 66O 660 REDUCTION TEMPERATURo c E

Fig. 4 Optimum reduction temperature.

3.2.1.2 Hydrofluorination tests

These tests were similar to the reduction tests except that complete reduction without sintering was ensured by longea r fo r C ° perioreducin 0 0 minutes)60 (3 d t a g .

Subsequent hydrofluorination is then carried out for a shorte o obtaiT r perio . curva n0 minutes °C r (1 d0 fo e 45 t )a a specific materia e tes s repeateth li t d with fresh samplet sa 500, 550, 600, 650 and 700 °C to obtain an "optimum" temperature profile for hydrofluorination. The entire sequence is then repeate r anothefo d r material.

The results are shown in Figure 5.

126 lOO

MATERIAL No« K (pg/gU) TEST CONDITIONS REDUCTION TEMP 600"C _ _ _ 1 41 REDUCTION TIME 30mlni ———— 2 212 — —— 3 507 HYDROFLUORINATION TIME lOmlns — — —— 4 814 GRANULE SIZE «2.36mm 5 1277 -3,36mm 6 72B6 — — 7 265

450 5OO 55O 600 650 TOO HYDROFLUORINATION TEMs ce P

Fig. 5 Optimum hydrofluorination temp.

3.2.2 Static bed tests - temperature dependence

3.2.2.1 Reduction

Two main conclusions emerge from the temperature conversion curves for the different materials (Figure 4).

e materialTh sw Na+ witlo K a h content hav widea e r temperature rang whicn i e h reductio n occu ca d wiln an rl therefore convert more easily in a MBR with moderate channelling.

e materialTh s wit a high h Na+K content hav a lowee r conversio d narrowean n r effective temperature range.

It mus e pointeb t t thaou dt these temperaturee ar s average oven temperature d cannosan e relateb t d directlo t y reactor bed measured values.

127 3.2.2.2 Hydrofluorination

Hydrofluorination temperature curves (Figure 5) indicate a similar patter e reductio th o that nf o t n tests. Materials w Na-fwitlo Ka h value have bot a wideh r temperature range (450 °C to 650 °C) and higher UF, conversion (80 % to 10. 0%)

e materialTh s wit higa h h Na-t- K) hav 6 valu a narroe, (5 e w temperature range (550 °C to 650 °C) and a low UF^ yield (60 % to 70 %).

3.2.2.3 Preliminary conclusions

For both figures 4 and 5 the increasing UF^ yield at low temperatures result from kinetic considerations. At the high temperatures the particle reactivity decreases due to increased sintering and/or a decrease in available surface area.

Tabl showV I e s thadifficule th t t material ) hav6 a e, (5 s low ADU precipitation particle size (2-4 ym), whereas the better material precipitatio% s 0 hav5 a e n particle diameter greate. ym 0 1 r o that 6 n

Besides this complicating factor, Table V indicates that the materials with higher Na+K hav a lowee r sulphate content.

Increased reactivity during reductio d hydrofluorinaan n - tion of materials with low sulphate levels (up to 700 ppm) is known [4].

A test material (number 7) was prepared to check if the same Na+K with zero sulphate would behav accordancn i e e with

128 sulphat r Na+o e K rankings. This material 5 yg/gU(Na+26 = K ) behaved very similarly to material 2 (Na+K = 212 ug/gU). See . Tabld Figure5 an d V an e 4 s

e higheAth t re temperatureth l al , yiel r UF fo de th s materials decreased. (Figures 4 and 5).

This decreased reactivit o increaset e du y d Na+ s knowi K n maie th n t [4]problebu , n laboratori m y o obtaitestt s i sn data which can be reliably used for interpretation of plant operations.

From a MBR point of view the wider and higher the temperature conversion profiles, the better the material which will be converte plane th tn i . d(Figure 5) d an 4 s

This means that materials 3 and 4 would reduce better than material 6 which has a high Na+K content.

Furthermore, these temperature conversion profiles give n indicatioa e temperaturth f o n e excursion e reactoth s n ca r sustain. This indicatio n allowabla f o n e temperature excursion is, of course, more pertinent in the reduction reaction, sinc e hydrofluorinatioth e n reactio s reversibli n s i d an e conducted countercurrently, and has a reasonable residence time.

3.2.3 Particle reactivity - Na+K rate dependence

In order to compare the effect of Na+K, the rates of hydrofluorinatio e differenth f o n t material e "optimumth t sa " hydrofluorination temperature of approximately 550 °C were determine averagn a r fo de particle siz f 2,8o e . 6mm

129 e ratTh e constant fit a sfirs t order reactios ni ratd an e shown in Figure 6.

It can be seen that the increased amounts of Na+K reduce the reaction rate considerably.

0.6O

MATERIAL Na+K(pg^U) 7 265 0-50 3 507 4 814 5 1277 6 7286

O4O TEST CONDITIONS REDUCTION TiME 30MNS REDUCTION TEM0 P60 °C _c E 030 HF TEMP 550°C n (l-XKk.-I t

O20

040

0.00 40 80 120 160 200 Na+K(;jgmol/gU)

Fig. 6 Hydrofluorination rate as a factor of Na -f- K.

3.2.4 Particle reactivit siz- y e dependence

wer5 Figure d e an obtaine s4 d from tests carrien o t ou d material of about 2,86 mm particle size.

Preliminary results on the zero order rate constant for reduction show the particle size dependence in Figure 7.

e bul Th f materia o ke reacto th n i l r a siz fees eha d greater d m experimentatha(Tablm an 5 n ) eI l data wile b l obtaine n thii d s particle size range.

130 It appears from Figur tha7 epreliminare th t y results from the thermobalance give far higher reaction rates. The stati d tesbe ct parameter e beinsar g checke o obtait d n better ammonia circulation and closer temperature measurement of the layer.

030 MATERIA. 7 L TEST CONDITIONS 025 HF TIME 20mins HF TEMP 400-c O20 REDUCTION TEMP 60o«c

X-k.T 0.15 THERMOBALANCE

0.10

Q05

STATIC BED 0.00 000 1.00 2.0O 300 4.OO 5.OO 6.00 7.00 8.00 9.010.000

Dp (mm) Fig. 7 Reduction rate on thermobalance and static bed.

Since a thermobalance run is comparatively fast, it would be useful to correlate these results with the static bed and finally with plant operating conditions.

e particlTh e size dependenc hydrofluorinatioe th f o y n reaction must stil carriee lb d out.

4 REACTOUF . R4 RUNS

As indicated in an earlier section and Table II the UF, pilot reactor runs were relatively successful in terms of reactor availability and UF, yield.

131 O 200 40O 600 800 OFF 0*3 TEMP -AXIAL TEMP PROFILE ro eoo MAXIMUM i rf REDUCTION REDUCTION SECTION TEMPERATURE l, 3m TOO

-NM3

600 COO.INO -HZ MAXIMUM SECTION AXIAL 1.4m T HYDRO FLUORINATION TEMP TEMPERATUR:- j OFF 3*3 PROFILE 500 C7 MNE 6 (No.K- 7286 HF SECTIO* N 2,3 r. PRODUCT CONTENT (%) "0 20 r

4 REACTOUF R SCO 4QO 600 800 TIME TEMPERATURE >c*i

f'ig 8 - Pilot reacto witn ru rK (Minh . + hig 6) ea N h

The reactor run with material 6, high in Na+K, (7286 Ug/gU s depicte)i n Figuri d . 8 e

From the static bed tests (Figure 4) for material 6 it is clear that sinterin r loso gf reactivito s y during reduction face . th t Thi°C y b borns i s0 t 70 ou e - 0 65 t a n i wilt se l thae reductioth t n temperatur ee sustaine b coul e t th no d n i d reactor (Figur . Not8) e e als conversio, o UF thae th ts wa n e higth h d IKLFan w « lo content indicates incomplete reduction.

Als oe figur showa typicath s n i ei n l reactor temperature profile at a given moment. The typical sharp reduction temperature vs reactor length indicates a narrow leaction front.

132 o ao 0 60 o lo 0 20 o

TE**- CCI

TEMPERATURO C E

Fig 9 . Pilo K (Mint . + reacto 7) ea N witn w ru rlo h

In contrast, a reactor run with material low in Na-t-K s show(Mini Figurn ) i n7 e . 9 Despite a varyine g maximum reduction temperature, a high yield of UF, was obtained. A typical reactor temperature profile is also given. In this cas e breath e d reduction temperature n contrasfroni s i t o t the sharp reduction reaction zone seen in materials containing high Na+K.

5. UF,, FLAME REACTOR AND PURIFICATION o

, materiaUF s e beeha Th l n G produceAE e th dt a thur fa s oa variablf e e operatinqualitth d an y g requiremente th r fo s flame reactor reasonabla hav o t e e extent been determined.

^ powdes UF Aftei reactes e A ha rth rUF e d th wit ~ F h filtered and condensed. These steps reduce the non-volatile fluoride impurities. Typical changes in impurity level obtained so far are reflected in Table VI.

133 TABLE VI

Typical impurity levels of non-volatile fluorides in UF, produced from UF, before distillation in comparison with the Enrichment Plant feed Specification.

Enrich= ment Element Compound Bpt UF4 UF6 Spec

(Ug/gU)

Cu CuF2 >1000 48 <0,4 5,40

Mn MnF_ >1000 33 <0,2 3S60 Ni NiF~ >1000 250 <0,6 7,10

Co CoF2 >1100 11 <0,5 0,64

Fe FeF2 >1100 1260 20,0 50,00

Pb PbF2 1293 4

Zn ZnF2 1500 29 0,3 14,00

Mg MgF2 2260 182 14,0 3,60

Ca CaF2 2500 1054 54,0 14,00

TABLI EVI

Typical flag element impurity levels in UF, produced from , beforUF e distillatio comparison i n n witEnrichmene th h t Plant feed Specification

Enrich= ment Element Compound Bpt UF 4 Spec. (ug/gU)

B BF3 -100 - - 0,11

W WF6 18 3 0,7 7,10 Mo MoF, 35 16 5,5 11,00 o V 111 24 1,2 0,11 VF5

Cr Cr02F2 200 23 16 14

Th ThF4 200 - - 0,71

Bi BiF3 1000 3,5 0,5 0,21 K KP 1500 390 0,3 - Na NaF 1700 470 30 -

134 Since the more volatile fluorides such as WF, . MoF, D D and VF_ flow with the UF, a distillation step is required for purification.

These impurities have been used as flag compounds for distillation design. Typical value f theso s e compounde ar s show n Tabli n e VII t mus I .notee b t d that these results, which were obtained during development of the distillation, are not consistent. However, it can be seen that in general the UF, s closi o enrichment e t feed material quality.

6. CONCLUSIONS AND FUTURE DEVELOPMENT

e compilatio Th U specificatioAD n a f o n n ensuring thaa t high UF, conversion is obtained with minimum distillation of the resultan , produc tUF n progress i s i t .

To convert mine-produced ADU, the contributory effects of specific surface area, impurity levels of sodium, potassium and sulphate must be quantified. A reaction kinetic modelling , reactoUF e s beestudth ha r nf o ycommence achievo t d e these goals.

7. ACKNOWLEDGEMENTS

e authorTh s wis o exprest h s their appreciatioC AE e th o t n e opportunitth r fo f submittino y g this papere th d als ,an o t o operations and analytical staff for their contributions to the work. Special mentio F Kruger J s mad" i n f o eR ,Viljoen , D Mitchell J Steynberg JacksoM B , G d A an nr D , A Odendaal W r D . (Appendix)

135 APPENDIX

MODELLING OF THE REDUCTION AND FLUORINATION OF URANIUM PARTICLES

1. SIMULATION APPROACH

(a) From the pore size distribution (PSD), assuming ideal cylindrical pores catalyse th , t particl s discretizei e d into group f poro s e geometries. (Fiftee r thifo ns example).

(b) The behaviour of each of these groups of pores are simulated and the global conversion of U0„ to products calculated.

mechanise Th ) sinterinf (d mo fairls i g descril il y - bed in literature and experimental data (to be collected) will be employed to define the relationship between the particle temperatur e sinterinth d an e g activity.

2. MODEL

2.1 Reduction reaction

e reactioTh ) s assumei n(a tako t d ee por placth e n i e according to

k A /dC= ' U03 NH3

136 PRODUCT

Rr R'

Fig. 10 Definition of pore parameters.

(b) The differential equation describing the reduction reactio pore s th showi e n n Figuri ni n . 10 e

p * , KI Q / d 2 2'ARL*4> *P*6 ds2 l+c}52s'p*(p-l)

with

C NH, L e = - ;A s• = — L ' RL ~ R NH,

k'R k-R 11 __J ; P = D R Du

and the boundary conditions

4> = l, s = 0 and 0' = 0, s = 1. (Side reactions are not included in this brief note).

137 2 n -reactio2 Fluoio t a rin n

The more simple model for the equilibrium reaction in e particleth , postulate e porth se reaction modes a l

°'5 S Z. f\ r\ i~t 1 lir,\J U i / ____d D . 2. .Z r rj i Z _____4_T . 2 RL K ^-,0 ^TTn ds c C HF U02

R / L = A with T n RL P P

2 2 d) = R -k-Snn /D p U02 p

e = CHF/C°HF

= surfacS. e i are f o a l

= equilibriuK m constant

. 1 = s , 0 = ' 6 d an 0 = s l, = 6 and

(This differential equatioF H e solves th i n r fo d reactio e othenth onlrd an sidy e fluorination reactione ar s evaluated on an ad-hoc basis according to the kinetic data available.)

0 3. PRELIMINARY RESULTS

Based on kinetic data obtained from Harrington and Ruehl d alsan eo Gmelin differentiae ,th l equations were solved for two particle sizes but no compensation was made for the sintering effect e resultTh . e depictesar Figurn i d . 11 e

138 too

090 -

080 -

070 -

060 - CALCULATION TIME INTERVALS oois D DlSCRETIZEPS D INT INTER5 O1 - VALS 050 - 0 CONVERSION 040

100 REACTION TIM» is E

Fig. 11 Conversion profile for UO-, U0„, UoO„.

REFERENCES

[1] COLBORN, R.P., et al, "Uranium Refining in South Africa", Productio Yellof o n w CakUraniud an e m Fluorides (Proc. Adv. Group Meeting, Paris 1979), IAEA, Vienna (1980) 229.

[2] NOTZ, K.J., MENDEL, M.G., X-Ray and kinetic study of the hydrogen reductio y-U03f o n . Inorg,J . Nucl. Chem4 1 . (1960. )55

] GMELIN[3 , "Gmelin Handbuc Anorganisher de h n Chemie", Uran, Nummer 55, C2. Springer-Verlag, Berlin Heidelberg, (1978) 217.

[4] HARRINGTON AND RUEHLE, "Uranium Production Technology", Van Nostrand, (1959).

139 THE CONVERSION FROM URANIUM TETRAFLUORÏ7>E INTO HEXAFLUORIE VERTICAA N EI L FLUORIZATION REACTOR

ZHANG ZHI-HUA, CHAO LE-BAO Bureau of Nuclear Fuels, Beijing, China

Abstract

Experiments conducted in a vertical fluorization reactor of . diaoetemm r 0 convertinfo 20 r g uranium tetrafluorid r trio e - uraniuiB octa-oxide into uranium hexaf]uoride are reviewed. A brief acouot of the process flow sheet» major equipments employe d operationan d s procedures' s ,i presented e Th . advantages of euch a reactor are discussed. s i concludeIt d that suc a hvertica l fluorization reactor posesses certain advantages over conventional fluorization equipments e employedb n e conversioca th t t onlI ,nu r . fo y f o n uranium letrafluoride into uranium hexafluoride t alsr bu ,fo o the direct transposition of tri-uranium octa-oxide into uraniuœ hexa-fluoride. Even tri-uranium octa-oxide * witlo h concentration, of itf y can be effectively converted into hexafluoride by using this technique. II is hereby claimed that the verticil reactor in question can serve as a l fluor iy.ation reactor with, good adaptability.

Introduction

e produc Th f uraniuo D tJO m hcxa fluorid a ver s i y c important lin K in the processing and conversion of nuclear fuels. "Fixed bed reactors- horizonta reactorsg in t ta i ,laS fluidr/.ation reactors, flame reactors etc. have been employed for this purpose, with fluidualion reactors and flame reactors being currently used in many nuclear fuel plants. Selection of the appropriate reactor for the production of uranium hexafiuoride from uranius tetrafluoride and/or tri-uranium octa-oxide depends very much upon the characteristics of the reaction as well as upon the safety measures demanded. In general.the fiuoriza- tion of *ii i-uranium octa-oxide is usually carried out in a series of small-sized horizontal agitating reactors r yearsa resulFo s .a f ,o t continuous experimentation and innovation,a new model vertical fluorization reactor, specially designe o envelopt d e majoth e r advantagef so most conventional reactors mentioned above, has eventually come into existence. Experimental results of the conversion of uranium tetrafluoride and/or tri-uranium octa-oxide into uranium hexafluoride

141 ia vertican l fluorizatio m nm diamete reacto0 20 s provef o ha r n that e productivitth e reactoth f o s comparativelyi r y high» with excellent direct yield f metao sd comparativelan l w residuelo y . Furthermore,the reactor itself is structurally sisple, relatively lo? cost» easy to operat d easan o emaintaint y s alsi t oI observe. d thae verticath t l reactor is adaptable to most raw ssterials,adaptable to the conversion f uraniuo m tetmluoride into uraniuœ hexafluorid o t s tha»ela es a tl of tri-uraniuip octs-oxide into hexafluoride.

Description of Flo» Sheet and Process Equipants

Preheated and pressurized fluorine g&c is passed into a gas distributor located at the bottom of the vertical reactor. e sa»Ath t e time- uraniu» tetrafluorid r triuraniuo c n oota- oxide d inte ireactofe s th o r throug a feedeh a feedin d an r g pipee Th . e solith d d an fee s dga meet countcrcurrcntb

UP.

UP;

1—Buffering Tank: 2—Cleaner; 3—Pressurizer! F=\ 4—BafferiDf Tank; 5—Prebeater! 6—Vertical Fluorization Reactor; 7—Feeding Pipe! 8—Feeder! §—Hopper ! 10—Filtering Trap: Fig. l Plot Sheet of Fluorization 11—Sla« Collector

inside the reactor and the feed is converted into uraniu« hcxsfiuoridc. The outer shell of the reactor is electrically heated to provide the temperature needed for the reaction to proceed. A continuous stream of compressed air is introduced between the shell wall of the reactor and the electrical heating device to cool down the »all tesperature. Functioning of the electrical heating device and cool ing syste» is automatically controled through necessary instrusentation. Reactor gas.

142 F etcH s directe. composei , 02 - 2 f UFçN o d , ^ int e filterinth o g trao t removp e solid dust entrappe d proceed an e dcondensin th o t s g system (not shown in Fig. 1) , where entrained uranium tiexafluoride is recovered. The filtering trap is periodically blown with flourine gas o effect a tlowerin s resistancit n i g o flowt e . Slag foned durine th g reaction accumulate e reactine trttoth th f o t a s s gdumpei zon d dan e e intconnectinth o g slag collecto y b invertinr e turnablth g s ga e distributor. Reactos œonitorei s d ga ranalyse an d y automatib d s ga c Chromatograph)1. Referring to Fig. 1, it will be noticed that the major pieces of equipments employed in the experimental systeu are as followslvertical fluorization reactor, pressurizer, feeding device for solid feeds, electrical heating device and filtering trap. The vertical reactor of 200 DI diameter (Fig. 2) is made of Monel alloy.Five temperature measuring tappings, with thermocouples inserted, are located along the length of the reactor body to register the temperature e correspondinth t a s g e sectionreactoe th Th f o . rs pressurize s fundamentalli r a diaphragy n compressor.The feeding system r soli fo w materiara d s composei l a hopper f o d a feede, d feedinan r g pipe. The feeder is driven by a variable speed motor, allowing remote and automatic control of feeding rate ia the central control rooi. The reactor gas filtering trap is of inserted design, so that it may be e reactoth f o o allot r p wto direcinstallee th tt a droppinde th f o g blown dust inte mai th oe reacto n th bod r f furtheo yfo r r reactione Th . electrical healing devic s dividei e d into five different sectionf o s

_n !ir-i AA,

1— Inler Elenenlafo t l Fluorin\ e 2,—Gas Distributor; 3—Main Body of reactor: 4—FeedinS Pipe! 5—Filtering Trap.

Fig. 2 Vertical Flaoriwtion Reactor

143 equal power supply s distributoga A . e s locatee i bottorth th f t o a id reactor to ensure even distribution of the gas stream within the reactor. Us distributing plate is -made fron Monel alloy. Elemental. Fluorine is pressurized through the diaphragn compressor before entering the reactor. The operation procedure begins with the switching on the electrical heating eleaents to allow the temperature within the reactor to rise to a pre-set value. Then, elemental fluorine is- introduced for a few minutes. After that» switch on the feeder motor, lettin a pre-se n i g t amoun f o uraniut m tetrafluorid r tri-uraniuio e s octa-oxide intermittently» to initiate 'ignition' for the chemical reaction e temperatur th s sooA s .a e nreacto th f o e r body reachea s certain value» feed in uraniua tetraflucride or tri-uraniun octa-oxide continuously» and the system will be directed to continuous processing. Durin e coursth g f continuouo e s operation»the rat f floo e f flourino w e gas in respect to the amount of uranium tetrafluoride or tri-uranium octa-oxid s closeli d fe ey adjuste n accordanci d e with indicationf o s analytical result s Chromatographga se showth y b o .

Result d Discussionan s s

w Material1Ra . s The two species of feed are ^ed in the experiment oce is refined neT uraniua tetraflaoride and tne °^ tri-araniui octa-oxide. The tri- uraniua octa-oxide is a purified product froa waste containing uranium in uranium enrichment plants. Evidently, the two species are differen n i theirt ' respective p|ysica d cneœicaan l l properties» hence different technological parameters have to be eiployed for their processing, yet the resulting products obtained are both of extremely high quality. For uroniun telrafluoride» the direct'rield of netal is over 99.5* with 0.02—Ü.05X residue; and. for Iri-uraniun octa-oxide» the' respective d 0.5Xfigurean , e abovar X Evidentlys97 e , suca h vertical fluoriiation reactor proves to be very well adapted to the processing eithe w materialsra r » with steady proceedings. Hences i t i , hereby daisied that such a vertical fluoriz,at ion reactor can be used for the two different purposes advantageously.

2- Reaction Temperature

Reaction tenperature is. of course» one of the important parameters for the reaction of urlanius tetrafluoride or .tri-uraium octa-oxide with elemental fluorine to form nraniuœ hexafluoride» as it s weli l acknowledged e reactiothath t y cheiiicanan rat f o e l reaction depends very ouch on its reaction teœperatore. /U roon temperature» araniui tetrafluoride reacts rather slowly with eleeenta] fluorinet I .

144 is found that onl t temperaturea y s over ZoO^C. woule reactioth d n procee a significantl t a d y faste e temperaturrth rate n I . e rangf o e 0 -SOO'c.25 e rat th f fluorizatio»o e s i apparentln y independenf o t reaction temperature s pointei t t I tha .t ou 250°Cda t e interactioth . n f tetrafluorido d elementaan e l fluorin a ver s yi complexe - process. Intermediate products such as \^^j> l^Pg, ^5 etc-ar e formed. Further fluorization of these indteriediates leads to the formation of uranium hexafluoride s suggestei t I . d that such interaediate product e mosar s t unstable and tends to 'sinter' within a certain temperature range. Further heating , such intermediate products can occ«r disputation reactions» forming uraniua tetrafluorid d uraniuan e m bexafliioride. The rate of dissiutation holds a linear relationship with reaction temperature. Consequently, a higher reaction temperature not only would favor an. increase in the rate of fluorization but also woul r«mulp e d tutHi f o edlsmulntion e . This would mcnn loss intermediate products formed, shorter existence of such intcrmcdilc product e reactioth n I d sclluinatlo an n f theio n r 'sinteringa s A . ' natter of fact. it Is observed that at températures above 'J5()°G, 'sintering' of Intermediate products can be essentially avoided. Howcvcr.it mus o pointeb t t thaou d t thoug n increasa h n reactioi e n temperature does favor the processing of-the fluorlzfltion. too high a temperature is regarded as inappropriate. This is particularly true dealing wite fluorizatioth h f tri-uraniuo n m oeta-oxide, whics i h higiiîy exothermic,otherwise unforeseen' operating problei y conia se into existence.-Furthermore, extreiely high teiperatures are capable of accelerating corrosive effecte e reactordetcriouth ar d o t san »o t s e reactor th e lif th f o e . The principal rule for the choice of reaction temperatures for the flnorizatio f uraniuro n a tetrafluorid r tri-uraniu/o d an e a octa-oxide in a vertical reactor should be as follows! select the lowest possible temperature but be sure that the rale of fluorization reaction is comparatively high at this chosen temperature. Experinental results has indicated that when appropriate temperatures are chosen» the rate of fluorization can be optimized and 'sintering' of intermediate products can be avoided. Furthermore, the exothermic behaioar of the flcorizatio f tri-uraniuo n m octa-oxid e kepb n t ca eunde r control, thus eliminating possible operating problens. Vith appropriate teiperatures under control, the operating process with proceed steadily and keeping the reactor wall temperature will not be of any problei.it is observed that with appropriate reaction temperature chosen, the experimental reactor does not shown any significant corrosion during an extended period of repeated experiaentation. This happens to be a prominent advantage for employing a vertical fluorization reactor, which is seldom attained by a flame reactor or a horizontal agitating reactor.

145 . Los3 f Element«o s ! Fluorine Ia vertican l fluorizatlon reactor e soli,th d feed drops down fro» e reactoth f d elc»cnt«o an rp to e lth fluorine enter e reactoth s t a r its bottoi part through a gas distributor. Solid feeds of uraniua tctrafluoride or tri-uraniui octa-oxide fall down the lain body e reactoth y f b o gravitr d folloan y a comparativelw y evenly distributed path. At the same tiie» elemental fluorine flows up eountercurrcntly and evenly distributed by passing through the gas distributor. Consequently» bolter autual contact between the two phases within the reactor can be «aterialized. Purtheriore, excess elemental fluorine wile ablb o havlt e e contact with fresh uranim tetrafluoride at the feeding systea located near the top of the reactor- This would §ean that a vertical reactor can operate efficiently with mininui excess elemental fluorine s beeha nt I . proven in our experiaents that only about 5X excess eleaental fluorin s i necessar e siioot th r hfo y operatio a singl f o n e vertical fluorization reactor to ensure excellent conversion of uraniua tetrafluoride r tri-uranifo . m octa-oxide e correspondinth » g figure is aroun . Evidently7X d » n thineveca s e achieveb r d wit a flaih e reactor» which calls for an excess elemental fluorine of over e 25%.saaTh e goes true wit a singlh e fluorization reactor» which requires more than 1QX excess fluorine. Less fluorine lost is therefore e claiieb anothe o t d r advantag r verticafo e l flaorizatioo reactors. . Anoun4 f Slad Productivito t an g e Reactoth f o yr s alreadha It y been aentioned before thae conversioth t f o n uraniui tetrafluonJe and tri-uraniun octa-oxide into hexafluoride in « vottu'a! rciH'ior gives sl:igc of I'.ilZ I). IttX and not any more ihnn I)X respectively.h . lil a flash c realtor of.the same diameter, slag foricd »ould amoun o 1-ÜXt t e amounTh . f o slat g foricd durine th g reaction process would unavoidably be reflected in the direct field f notao i n importanattaineda s i d an t, indicatio e operationath f o n l efficienc a reactor f o y n thii ,s respects conspicuoui t i , s enough that vertical reactors prove to be highly advantageous. Low slag foriation in a vertical reactor is credited, first of all, to the structural design of the reactor. Countercurrent flow of solid feed and elemental fluorine provides excellent conditions for bette d thorougan r h reactio e takeb o t nn place. Unconverted intermediate products would fall onto the the gas distributor at the botton and are reacted upon by freshly introduced elemental fluorine; further promotios conversioit f o n n into hexafluorid s henci e e effected f courseO . , steady operatio n extendea r fo n d perios i d another reason for low slag foriation.

146 ïitb this innovative structural design» handy to operate and easy to get steady operation, the reactor claims to have rather high n diaaetenu 0 productivity20 r a reacto r f Fo o thir . s design» 4000-5000 kg/œj . h. of uraniun hexafluoride can be attained. The achievement of higher productivity can be explained as follows. First e structurath , e l verticath desig f o n l reacton ca r conduct out reaction heat- froa 'the reactor under predetermined conditions and allows handy and steady operation. Secondly» increasing the rate of gas flow is possible with such a design. Furthermore, solid feed of appropriate particle size reacts rapidly with elemental fluorine at the bed teaperature to forn intermediate products» which would further react with elemental fluorine to give uraniuis feexafluoride» concurrently with the disoutation reactions at high températures s i notice t I d. that intermediate products react with elemental fluorine during their fall onto the bottom distributor» and sinc e particlth e e specifieth si'/. d an s c gravit f suco y h Intermediate product e mucar s h greate d highean r r thae originath n l solid feeds, e allowdar highef course O » s flo. Mo ga e rat r w s f th ,o ratee la f o s should be controlled to be greater than that required for the fluorizatio e originath f o n l solid feed t belobu , w that which would cause entrapment by the gas. An increase in the rate of gas flow withi e reactoth n r would aean a increasn e aaoun f o elementat l fluorine inside the reactor body, which in turn would bring about an increas n productiviti e e reactorth r fo y.

5. Special Features Concerning tie Fluorization of U^CL.

There are certain special probleis that should be taken care of wïien dealing wite fluorizatioth h f tri-uraniuo n « octa-oxide. Exten- sive hea f reactioo t n foned durin e fluorizatioth g f o tri-uraniun a octa-oxide must e systee conducte,th b f o a t otherwisou d e the- reactor body »ill be ruined. Furtberaore, since tri-uranium octa-oxide is a recovered by-product froi uraniui enrichment t involveI . s 5~l) of various enrichneat» Safety aeasures e nusattendeb t o durint d g its fluorization. Both of these problems have ibeen successfully r experimentsattendeou n i o t d .

Conclusion Experimental results attained prove that converting nraniuB tetrafluohde or tri-uraniaa octa-oxide into oraniui hexaflouride in a vertical fluorization reactor under pre-deterœined conditions i s highly advantageous in many respects. Low loss io eleiental fluorine» w slad higlo an g h direct yiel f aetalo d » cosparabl o thost e e attained

147 lith horizontal agitating reactors e claiœedar , . High productivity cofflparable to flaue reactors is also evidenced. Sinple structural design» bandy to operate, easy to œaintain, less floor space occupied» lot capital cost, excellent adplabilily for processing different ra» »«terial e othenr s r additional advantages clilned. Vertical flaoriza- tion reactors are'Borei advanced than horirontal agitatind an g fltttdUation reactors, and they cotparable to flaae retctors. Extended period f repealeo s d cxperiucntalio s proveha n n thae th t technique of converting uraniu» tetrafluoride or trl-uraniui octa- oxidc into uranium hexafluorid a vertica c i e l fluorizalion reactos i r »ost reliable. The reactor itself and its operation are technically proven e techniquTh . e successfullb n ca e y transplante cowercian i d l practice.

References

ZHANO IHI-HUA THE CONVERSION FROM URANIUM TETRAFLUOR lDK INTO HEXAFLUORIDE IN A VRT1CAL PLUORI7.ATION REACTOR

'CHINESE-JOURNA P NUCLEAO L R SCIENCE AND ENGINEERING* Vol.2, 1982.1

148 THE SOLVENT-CONTAINING RESIN FOR REMOVING URANIUM FROM EFFLUEN REFININF TO G PROCESS

ZHU CHANG-EN, ZHOU DE-YU Beijing Research Institute of Uranium Ore Processing, Beijing, China

Abstract

Two types of solvent-containing resin (CL—5209 and CL-5401) have been prepared by suspension polymerization of monomer presence th n i sf activ o e e Components e activ.Th e component e dialkyar s l alkyl phosphonate (for CL—5209d )an trialkyl phosphine oxide (for CL-54Ü1) respectively. Uranium in dilute nitric acid or acidic nitrate solution can be adsorbed selectively and efficiently with these resins. Losse f theso s e active component waten i se les ar r s than 4 ppiii. Because of these characters, these resins can be used to remove uranium from effluents of refining process, such as the raffinate from TBP extraction stage, the filtrate from precipitation stage, etc e concentratioTh . f uraniuno m in the raffinate treated with CL-5209 or CL-5401 column is lower than 0.05 mg/L. Ammonium carbonate solution can be used as eluant for loaded column. Water may also be used for élution of loaded CL-5209. Uranium in the eluate may be recovered by precipitation or by recycling the eluate into the feed solution of TBP extraction stage.

The chemical concentrates produced by uranium ore processing variou e plantth d san uranium-containing wastes fro e productiomth metallif no c uranium fuel elementd an s uranium dioxide ceramic fuel elements contain so many impurities that it is necessary to refine these materials. TBP extraction process is commonly used for refining. The waste liquor from this refining process consists mainlf yo the raffinate of the TBP extraction stage and the filtrate from the precipitation stage. This effluent contains about , 5-20+ H _ N 0 3 mgU/L0- d man,an 0.1-y^ differenNO N 3_ t impurities. In general , uranium in this effluent should be remove t firsta d e uraniuTh . m concentratio decreasee b n ca n d to a value of less than 0.05 mg/L by neutralizing the

149 ei'flueul with ammonium hydroxide loi lowed by udso rptioii with silica gel. The disadvantage oi' this procedure is that a highly radio-active solid uuste containing nrani inn is produced n exchangio n A . e rosin e m incapabls adsorbin' ol e g uranium from dilute nitric acid solution. Cation exchange resins have a poor selectivity lor uranium. So these ion—exchanger e unsuitablar se treatmenth e r th fo ef o t efiluent mentioned above, it is »ell known that some neutral esters of the acids of phosphorus, ul/^o pliosphlne oxides n extracca , t uranium from nitric acid solution selectively and effectively. However, the uranium concentration in the effluent is so low that solvent extraction technique« cannot be used economically for removing uranium. H.Kroebel and A.Meyer reported tiiat a solvent-containing resin was prepared by adding an extractant to a mixture of monovisiyl a«d polyvinyl compounds and by carrying out the suspension polymerization e presenc e extractantth th n i f o e . They calle e resith d n obtaine s Levextrea y thib y d wa s l resi d tooan n k examply b e a TBP-coiitaining Levextrel type resin to describe its characteristics -ti-7}. A detailed description of the structur d characteristican e Levextree th f o s l resis nwa given by H.W.Kauczor and A.Meyer X8l. Some authors have investigate e possibilitth d f applyino y e TBP-containinth g g Levextrel resi nucleao t n r fuel reprocessing {9-iiJ.A TOP—containing resin CL—TBP simila Levextreo t r l resis wa n prepare r researcou n i d h institut s abilitit d o t yan e adsorb uranium was determined. The results showed that this resin had a saturation capacity ol' 148 mgU/g resin in a solutio nitriN _ 5 Ö e gU/L cÜ. f th o n aci d s .A an d concentratiOU f nitrio R c acid uraniuan d m »ere lowered, however, the saturation capacity of the resin for uranium decreased considerably. The active component in the resin, TBP, shows a high solubility in water (about 400 mg/L), so this resin is unsuitable for the treatment of the effluent from refining process knows i t nI . that dialkyl alkyl phosphonate and trialkyl phosphine oxide exhibit a much higher ability to extract uranium from nitric acid solutions

150 than that of trialkyl phosphate. Moreover, the solubility of these compounds in water decrease considerably with the increas f theio e r molecular weights. Base n theso d e facts, two types of solvent-containing resins were developed in our research institute and named as CL-5209 and CL-5401 respectively.

1. PREPARATIO E CHARACTERISTICTH D NAN CL-520F SO D 9AN CL-5401

The so called 5209 is a dialkyl alkyl phosphonate, in which the total number of the carbon atoms in the three alkyl groups is no less than 22. 5401 is a trialkyl phosphine oxide, in which the alkyl group is Cg-Cg, mainly Cy. They both were synthesized in our research institute. An organic phase of styrène, divinylbenzene, initiator and the extractant (5209 or 5401) was added to a polyvinyl alcohol solution. The mixture "was heated with stirring for 0 a raintemperatur3 o , left ,C t thia t0 8 s f o temperature e for a period of 10 hours. After cooling, the aqueous phase was seperated off and the resulting bead polymer, i.e. solvent-containing resin, was washed with water and dried at 60°C. The properties of the solvent-containing resin are dependent on the dosages of divinj1 benzene and extractant. Whe e amounth n f divinyo t l benzeri less i o s thae nth lOfof o total amount of the monomers, considerable amount oT the extractant can swelled into the matrix of the styrène- diviiiyl benzene copolvmer, resultin n decreasini g g adsorption rate. It is evident that tiie content of the extractant in the resin shouls hig a o obtais possiblt e a hb ds a a n o s e solvent-containing resin with high capacity o mucTo . h extrtictan e resinth n i ,t however, would give riwo t e decreas e mechanicath e l stabilit e resinth f o .y The general characteristic CL-520f o s d CL-5409an 1 are shown in TA13LE 1.

151 TABL Genera. E1 l Characteristic f Solvent-containinso g resins

CL-5209 CL-5401

Active component 5209 5401 Active component content, *t/o 60 50 Particle size, mesh 20-60 20-60 Bulk density, g/mL 0.58 0.55 Specific gravity less tha1 n

. 2 UKHAV10R F CL-5UOO S y AN«) CL-540 U1UNJUN I 1 M ARSOIU'TION AND EhUJ'H/N

Ü.I, Adsorption rat« 0.5 g of sol vent—coutaini iig resin mid 50 «iL of uranium

solution (1.16 gU/L UN0_ N , 31 ) nere added int a 100-mo L ground—glass stoppere e flasd s th .shakeflus wa kd an hn vigorously. Samples were take t differena n t contact timf o e 10, 30, 60 and 120 win., respectively. The remaining uranium concentration in these samples were analysed. Thereby the uranium adsorption capacity of the resins at different contact time coul calculatede b d e adsorptioTh . n curvef o s CL-520 d CL-5409an uraniur 1fo showe ur mF1G.n i n y b 1 plottin e uraniuth g m adsorption capacit e resith f no y ugainst e contacth t time. F1U.1 indicates tha e resint th bot' soi h have a high adsorption rate for uranium.

2.2. Relationship between uranium adsorption capacity of the resin and uranium concentration in aqueous solutions with different concentration f nitrio s c acid 20 mL of uranium solution (1 gU/L, 1 JSI, 0.5 ^J or 0,1 N^ HNOg) and a certain amount of resin were added into an 100-m f ground-glasLo s stoppered flask. Al'ter o shakintw r lo g hours,samples were taken f î om the .solution to determine the remaining uranium value uraniue . th Tuu s i tm concentration i n

152 o a a -H a. n rt « u t. fl iO

4 U 00 80 100 120 Contact time, min F1G.1. Adsorption curves of sol vent—containing resins the aqueous solution in equilibrium. Thereby the urunium adsorption capacit e resith f no y coul e calculatedb d . FIGs.ei, 3 and 4 show the relationship between the urunium adsorption capacit uraniue f CL-520o yth d m9an concentration in the aqueous solution ( 1 N., 0.5 _N and 0.1 _N HNOß) in equilibrium respectively. Some results with CL-54Üd 1an CL-TBP are also given in FlGs.2 and 4 for comparison. FIG.2 indicates that CL-TBP cannot adsorb uranium effectively from an 1 ÎJ HNOy solution. On the contrary, CL—5209 and CL-5401 exhibit a high ability to adsorb uranium in suc a solutionh .

00 * CL-5401 x 50 £ a L-5ÜÜ9 a! 40 0. m r} 0> u u c bo 30 / 0 \ O D / -p M Oi a 2/ 0 . CL-TbP L, / A O ffl O T3 -

FIG. 2. Relationship between adsorption capacit e resith f nyo and equilibrium concentration of uranium in 1 ÎJ HNOo solution

153 J-iü.3 indicates that CL-520 n adsor9ca b uranium effectivel ILNO. yN froy 5 solution0. m . Flu.4 indicates that CL-520 o longen s i 9 r effective lor adsorption of uranium from 0.1 ]S HNOß solution. However, CL-540 s stil1i gooa l d adsorben uraniur fo t n suci m a h dilute nitric acid solution.

faU

50 u a ai H 40 a, n ö v y u M JO L-5209 20 o n •a 10k

0 2 1. 0 1. 8 Q. 6 0. 4 0. 2 0. 0 Equilibrium concentrâtjon of urnnium in solution, gU/L

FIG.3. Relationship between adsorption capacit e resith f no y and equilibrium concentration of uranium in 0.5 ]S[ solution

50 o a •i H 40 O. 91 d S) * CL-5401 o t. C bu JO o "^,

20 o n JL-5209 -a 10

0 2 1. 0 1. 8 0. 6 0. 4 Ü. 2 0. 0 Equilibrium concentratio f ur/iniuo n m in solution, gU/L.

F1G.4. Relationship between adsorption capacit e resith f no y and equilibrium concentration ol uranium in 0.1 N^ HN03 solution

154 2.3. Elution of uranium from loaded resin The uranium-loaded CL-5209 could be eluted with water. 8 bed volumes of eluant were needed to complete elution at a flow rate of 1-1.5 bed volumes per hour, in this case, tailing was somewhat observed. When a 5/"o ammonium carbonate solutio s use o elutwa n t d e uranium-loade th e d CL-5209o ,n tailin s foundwa g . Wate s ineffectivwa r r elutinfo e g uranium from CL—5401. A 5fo ammonium carbonate solution was used for this purpose volumed be 5 .f eluan o s t were neede o complett d e th e elution.

2.4. Treatmen uranium—bearinf o t g effluent

The effluent (200 n^U/L HN0. N 3,2 ) from refining process was treated with a column loaded with 20 Lg CL—54Ü1. f efflueno m s passe 2 A tota1 wa t f do l through this column. After this treatmen e uraniuth t m concentratioe th i ii n raffinate was less than 0.05 uig/L. \ 5% ammonium carbonate solutio s use s wa neluanta d . Uraniue eluatth n i em obtained could be recovered by recycling the eluate into the feed P extractioTB solutio e th f o n stag r precipitateo e s a d uranium carbonat y boilinb e e eluateth g . After filtratior o n décantation, uranium carbonate so obtained was recycled to P extractioTB n stage.

. 3 LOSSEACTIVE TH F SEO COMPONENT

CL-5209 resi s contacnwa t with water thoroughly. Then 5209 concentration in the aqueous phase was analysed by gas- liquid chromatography. The result indicated that the loss of 5209 in water was 4 mg/L. Put 0.5 g CL-5401 into 10 L of water and made them contact each other thoroughly. Then the uranium adsorption capacity of the resin was determined with uranium saturation method. The result showed that the uranium adsorption capacity of the resin so treated was only 6.7J& less than the value of the untreated resin. According to this result the los 540f so waten 1i r coul calculatee b d mg/L7 1. s .a d

155 e characteristicBaseth n o d s described abovee w , suggest that CL-520 d CL-540an 9 1 shoul e applicablb d e th o t e removal of uranium from effluents of refining process.

REFERENCES

KBOEBELjIl. , et al., üer. Offen. 2,162,951 (1973). X2) KftOEBEL,Jl , Ger. l a t . e Offen , . . 2,244,300 (1974). X3Ï KfiOEBELjH. , et al . , üer. Offen. 2,244,323 (1974). £4) KBUEBEL,Jl. , et al . , ürit. 1,407,257 (l97ö). . 3,9ÜO,76KBOEBEL,R.US , . l a 2 t (1976)e , . KJlüEBELjH., et al . , Proc. Int. Solvent Extraction Conf. Lyon (1974) 2095, X7l KHOEHEl,,ll t u.e , 1, liiat. . Chem. Eng. Symp. Serie« Ko, Hvdrometallurg.42 y (l97ö. 24 ) X8X KAUCZOIl,H.W , Hjdrometa. l a t e , . l lurg (1978^ . 3 / Ü5 ) OUHSENFEL1J,W. , et al . , Kerntechnik JJ3 (1976) 258. OC1ISENFELD,W , Heaktortag.. l a t e , . , XFachvort1 r. (1977) 381. ALFOIlUjC.E. , et al . , lleport 1978,IIFF-2770, CONF-780452-1.

156 WASTE MAN/ ^EMEN REFININE TH N D TI GAN CONVERSION OF URANIUM IN CANADA

A.W. ASHBROOK Eldorado Resources Ltd, Ottawa, Ontario, Canada

Abstract

Eldorado Resources Limited has operated uranium refining facilities at Port Hope. Ontario, since 1942. It was not. however, until 1955 that solvent extraction was used to produce nuclear grade uranium trioxide n 195I .e productio8th f o n uranium dioxide was commenced to provide fuel for the domestic CANDU reactors. Conversion facilities were added in 1970 to produce uranium hexafluorid r exportfo e . In 1975 a decisio s mad wa w no builrefinin t ene a d d an g conversion facility in the Port Hope area. The outcome was, however, that a new refinery was built at Blind River in northern Ontari om nort k (som f Por0 o hw 50 et ne Hopea d )an uranium hexafluoride plan te existinbuilth t a t g Port Hope facility e BlinTh . d River refinery came on-strea n 1983i m d an , the Port Hope conversion plant in 1984. Today n Canadai , , uranium yellowcak s refinei e t Blina d d River e uraniuanth d m trioxide produce s shippem souti k d o 0 t h 50 d Port Hope s convertewheri t i e o uraniut d m dioxid d uraniuan e m hexafluoride. Process wastes from the Port Hope facility have been since 1956 and continue to be buried in a licenced waste management site m frok som m6 1 ePor t Hope. o wastThern s i e management site availabl e Blinth do t eRive re plannedrefineryon s i .r no , Becausincreasinthe of e g nee betteto d r manage process wastes, and to reduce their production. Eldorado embarked on a major program to reduce the amount of waste produced, to reduce the amount requiring burialo maximist d an , e recycl d reusan e f o e these materials. This paper addresses the approach, the successes or otherwise, tha n tthii Eldoradsd ha endeavor s ha o . Sinc s inceptionit e , the program has resulted in achieving more than an 85 percent reduction in the wastes requiring burial, through the recycle of materials to other industries and by instituting process modifications which limit waste production. Initiativeo t s reduce still furthe e neer buriath rfo d e beinar l g actively pursued.

157 1. INTRODUCTION Process wastes arising from the refining and conversion of uranium yellowcak o uraniut e m hexafluorid d uraniuan e m dioxide fuels e pasthaveth n ,i , been o wastburietw t ea d management sites nea o Port r t Hopee firsTh . tt Welcome a sit s i e , some 2 km from Port Hope, and was used between 1948 and 1955. The second site is at Port Granby. some 16 km west of Port Hope. This site has been used since 1956. and is still in use today. Both sites are licenced, and are maintained and operated in accordance with the requirements of the Atomic Energy Control Board. Ovee las . th Eldoradryear5 t so r o s develope ha o a majod r progra o bott m h reduc e amounth e f wasto t e produces it n i d operation o recyclt waste d th s an mucss a possiblee a ef o h o T . date e prograth , s achieveha m r dcen pe ove t5 8 reductior n i n e wasteth s requiring burial y recyclinb , g waste o othet s r industry and by instituting a number of process modifications which limit waste production.

TABLE 1

TYPICAL CHARACTERISTICS OF RAFFINATE

Constituent______Concentration (q/L) Uranium 2 Thorium 6 Nitrate (as N) 11 Ammonia (as N) 4 Phosphate (as P) 17 Sulphuric acid 450

Silica (as SiO2> 19 5 0. Organics Iron 2 Aluminum 15 Copper 0.1 Density 1600 0 pCi/15 L Radium-226

158 2. WASTE GENERATION 2 . l Raffinate e majoTh r waste produce y Eldoradb d o results froe refininth m g of yellowcake to nuclear grade 003. The refining process consist f digestino s e yellowcakth g n concentratei e d nitric acid purifyinan d e resultanth g t solutio y solvenb n t extraction (SX), using tributyl phosphat a kerosen n i e e solution. Uranium is selectively transferred to the organic phase in the extraction step while the impurities in the yellowcake remain in the aqueous phase. This is heated with sulphuric acid to e nitrith driv f cof e acid whic s recoverei h d re-usedan d . The concentrated waste, called raffinate, is an inevitable waste product of uranium refining. It is produced at a rate of about 500 kg per 1000 kg of uranium refined. A typical analysis of currently produced raffinate is given in Table 1.

Since 1955. when 003 production began at Port Hope. Eldorado s useha a numbed f raffinato r e disposal methods. Initiallya , neutralized and filtered solution was discharged into Lake Ontario. This practice cease s replace wa n 196 i dd 8an y b d disposal of neutralized raffinate at a licenced waste management site. More recently e buriath , f liquio l d wasts wa e halted and a dry limed raffinate was buried at the waste management site or stored in drums at the refinery. All these three disposal methods had major shortcomings. Consequently. development work was carried out on the recycle of raffinate to a uraniu e recover me th residua milth r f fo lo y l e uraniuus d an m e sulphurioth f e milc th aci ln i processd . This recycld ha e the added advantage of consolidating the waste from milling and refinin e siteon t .a g

2 .2 Calcium Fluoride Most of the 1103 produced in the refining process is converted into uranium hexafluoride (UFg) e feeth .d materiar fo l uranium enrichment plants e uraniu. th r cen Aboupe f o t8 m 7 t refined by Eldorado is converted to UFß. Since the Canadian nuclear program uses natural uranium g productioUF l al , s i n exported. Figur 1 showe a ssimplifie d flowsheee th f o t conversion process currently employe Port a d t Hope.

To Atmosphere t Llm« r——*——t l -—«—-, Calcium I r J Scrubbers J- -I Filler "-• Fiuonde

I——T——J Sludg. I i.———KOH—— ...——J

UOj. Reductio FluonnaliO1 n1 n KIuonnationl Filler I Cold UFo t „ | [ , toUO I to UFb f Traps t t H, HF

FIGUR : 1 CONVERSIOE N OPERATIONS

159 Pure UO is first reduced to UC>2 with hydrogen. The UC>2

o aqueout s thei d fe ns reactor3 s s wherreactei t i e d with aqueous hydrogen fluorid o product e e UF4. Reactor off-gas, containing uraniu d hydrogean m n fluoride s scrubbei , n packei d d tower columns with a potassium hydroxide solution. UF from

the hydrofluorination reactors is dried and reacted wit4 h fluorine to produce UF6. The gaseous UF6 is filtered before being coole d condensean d n col i s solia dd6 tapsUF d . Off-gases are collected and scrubbed with KOH. Finally, the s meltei d transferreUF an dr 13-tonn o - 9 o t ed shipping

6 cylinders.

All gases leaving the UFg conversion process are scrubbed with KOH solutions to reduce HF and other gaseous and particulate constituent o levelt s s below emission guidelines. H solutionKO e Th s froe scrubberth m e reactear s d with limo t e convert potassium fluoride to KOH and to precipitate the fluoride as insoluble calcium fluoride. Filters remove the calcium fluoride, producing a sludge of about 35% moisture. This calcium fluoride e majosludgth s r i ewast e product from s conversioUF e th n process s producei t a ratI .f t o a ed r eacfo hg f uraniuk o 100abou g 0 k G 50 tm converte o UFgt d A . typical composition of the sludge is given in Table 2.

TABLE 2

TYPICAL COMPOSITIO F CALCIUNO M FLUORIDE SLUDGE

Constituent______Concentratio weighty b % ( n ) HO 35 £> CaF 26 KOH 11

CaC03 5

Ca(OH)2 23 Uranium 0.02

Sinc g productioeUF n bega n 1970i ne calciu,th m fluoride sludge has been transported to a licenced waste management site r buriafo n trenchesi l . 2. 3 Ammonium Nitrate Eldorado also converts refined uranium into ceramic grade uranium dioxide (UO) powder e powdeTh .s shippe i r o t d

Canadian fabricator o product 2 s e fue n CANDi l e pelletUus r fo s reactors e uraniu. th r cen Aboupe f o t m2 2 trefine y Eldoradb d o is converteo t d Refined uranium is converted to ceramic IK>2 by precipitating ammonium diuranate (ADU) from a solution of uranyl nitrate with aqueous ammonium hydroxide at about 80°C. The ADU is

160 separated froe ammoniuth m m nitrate solutio y filtratiob n r o n by a centrifuge system. This ammonium nitrate solution is the major waste product from UC>2 conversion e ammoniuTh . m diuranat s firsi e t drie d thean d n reduce o UC>t n drotari 2 y kilns e wastTh . e ammonium nitrate solutio s concentratei n y b d evaporation. About 800 kg of ammonium nitrate solution are produce r eacfo dh 100f uraniuo 0g k m converte o UC>2t dA . typical chemical analysi e solutioth f o ss give i n n Tabli n . 3 e

TABLE 3

TYPICAL CHARACTERISTICS OF AMMONIUM NITRATE SOLUTION

Constituent______Concentration Nitrogen 255 g/L pH 8.3 Uranium 0.00L g/ 3 pCi/1 L Radium-226 Density 1250 g/L

When UÜ2 conversion first began in 1962, and until 1974, the ammonium nitrate waste solutio s disposewa n f wite o d th h raffinate froe solventh m t extraction refining process. Betwee ne fal 197f th 1977o l d 4e ammoniuan .th m nitrate solution was disposed of at a licenced waste management site, separately froe raffinateth m . 2.4 Garbage In addition to process wastes (raffinate, calcium fluoride and ammonium nitrate) e refininth , d conversioan g f uraniuo n m produces miscellaneous garbage. This consistf o s non-combustible wastes whic e contaminatear h r havo e d th e potential for low levels of radioactive contamination.

All garbag s currentli e y burie a licence t a d d waste management site. The amount of garbage buried annually is about 0 tonnes50 . Combustible garbage is reduced to ash in an incinerator which s installewa t Pora d t Hop n 1979i e h productio.As s aboui n t 5 tonne 2 waste s buriei th sh t e a das annually l al d an , management site. 2.5 Scrap Metal Refinin d conversioan g n operation t Pora s t Hope have resulted in substantial quantities of scrap metal. This scrap metal consist f carbono s steel, stainless steel, copped an r

161 aluminum e for e materiath f pipingo Th mn .i s i ,l sheetind an g fabricated items. Past disposal practice for scrap metal was burial with other contaminated garbage licenceth t a e d waste management area. In 1980, however. Eldorado develope a proposad l which would allow scrap metal to be recycled through conventional channels Stringent requirement e enforce e recyclinar s th n o d g program, namely: o visibln - e uranium e scraexistth pn o smetal . l scraal - p a radiatiometa s ha l n leve f leso l s than 0 uRer hou50 n pe contactmo r d an ; e scrath - p materia s transportei l d directly from Eldorado e milth l o t facility where remelting occurs.

3. WASTE MANAGEMENT PROGRAMS 3.1 Waste Minimization Eldorado's primary objective with respec o wastt t e management is to reduce the amount of wastes which require burial at a waste management site. Significant progres s beeha s n made since 197 n achievini 5 g this goal, despite expansioe th f o n facilities which will almost triple the company's processing capacity. The systematic decrease in wastes buried during the period 1977-1985 is shown in Figure 2.

HaSILS fiUBJULaLJafiUUEE

12.000

10,000

8,000

6,000

2,000 On 1975 1976 1977 1978 1979 1S80 1981 1982 1983 1981) 1935

FIGUR2 E

Ammonium nitrate burial ceased after 1977. when it was sold as a fertiliser in the local area.

162 Raffinate recycl o uraniut e m mill s beguwa s n 197i n 8 wita h trial program d continuean , d thereafter. For the last four years a program to recycle calcium fluoride wast o steet e l mills a fluxin s a , g agent s beeha , n underway. Problems have been encountered with respect to the amount of uranium in the waste (<200 ppm). Thus while the federal regulatory body (AECB) considers <200 ppm uranium to be recyclabl a stee o t el mill e provinciath , l regulatory agency consider o highto s.e b thi o Procest s s modification e beinar s g develope o reduct d e uraniuth e m conten o allot t w recycla s a e flux material. If this can be achieved, the only requirement for waste management at Port Hope will be for non-combustible garbage and incinerator ash. Currently, the calcium fluoride is sent for burial at a licenced waste management site. e overalTh l progres n reducini s g waste quantitie s showi s n i n Figure 3 where the total amount of wastes buried is expressed as a proportion of the uranium processed at the refinery. In 1977 wastf . o e Por s burieg oveth K wa et t 2 ra Granbd y waste management sitr eacfo eh kilogra f uraniuo m m processed. This s beeha n reduce o abou t d5 kg/k0. t . U gContinue d operatiof o n recycling and minimization programs could further reduce this to 0.05 kg/kg U.

WASTES BURIED. PORT HOPF. AS « F1IMCTIO F UMHU1KHO RFFINFH 2.1

2.0

1.6

1.2

0,1

1975 1976 1977 1978 1979 Ï980198 0 19 1981 3 1981 1985

FIGURE 3

The successful reduction in the amount of wastes requiring buria s beeha l n achieve y vigoroub d s pursuio wasttw ef o t management strategies, namely: - develop new processes which minimize waste production; and - identify uses for process wastes which allow their recycling to other industries.

163 3.2 Process Modifications Development of new processes which minimize or eliminate waste generatio e preferreth s i n d waste management strategy. c processinUF w ne e gTh method, calle e "wet-wayth d " process, uses aqueous hydrofluoric acid to produce UF4 from UC>2 in place origina th f gaseou o en i lF H sprocess . This provider fo s a mord alsan eo F efficienH allow f o se procesus t s off-gaseo t s be scrubbed with water instead of KOH solution. The dilute HF solution, which results from water scrubbing, can be recycled e "wet-waytth o " e originaprocessth n (I l. process, water scrubbing was not practicable because there was no use for dilute HF solutions.) Some KOH scrubbing is still necessary wite "wet-wayth h " process s t becausi bu ,F H e e th mos f o t removed during the water pre-scrub, the volume of calcium fluoride is reduced by at least 75 per cent to less than 100 kg per 100 g 0UFgk . compare e previouth o t d s "dry-way" process. Since conversion of 1103 to UFg first began at Port Hope in 1970. fluoride emissions have been controlled by scrubbing process off-gases with KOH e scrubbin.Th g process resultn i s the nee o dispost d a calciu f o e m fluoride sludge. Figur4 e shows the increasing amount of sludge produced in 1979-82 as a functio e correspondinth f o n g productionUF g . Over this period n increasa ther s wa en sludg i e e production above what migh e expecteb t d becaus f increaseo e g productionUF d n I . fact, the increased calcium fluoride sludge generation is a result of improvements in air emission control. Figure 5 shows the relationship between the amount of sludge and ar arbitrary air quality index ovee perioth r d from 197 o present5t . (The higher the air quality index the greater the fluoride emissions. e improvemenTh ) r qualitai n n i recenti y t years ha s e expens th n beeincreasa t a f no e n calciui e m fluoride sludge generation.

CALCIU&f LlDfiUE PfflffllCÜ) JS A EUHCLL 1.0 -

0.8 -

0.1

0.2

1978 1979 1980 1981 1982 1983 1984 1985

FIGUR4 E

164 08-

06- FIGURE 5: 04 RELATIONSHIP BETWEEN WASTE 02 PRODUCTION AND AIR QUALITY

3.3 Waste Recyclin o Othet g r Industries Where it is not possible to eliminate a particular waste, recyclin o othet g r industries avoid e nee o burth st d y wastt a e a licenced waste management site. Three wastes - ammonium nitrate, raffinate and scrap metal - are currently being recycled. The substantial increase in the amount of wastes being recycled is show e recyclinn FigurTh i n . 6 e g program began wite th h f ammoniuo loca e us l m nitrat s fertilisea e n 1978i r e .Th progra s bee ha o msuccessfus n l thae demanth t d exceede th s supply. Eldorado's fertiliser contains less uranium and radium than most commercial fertilisers.

ilLi BLLIULD 111 ata imusi«i[s

,yy; n;a 19/9 1983 $ Ha i9&t FIGUR6 E

The recycling of raffinate to uranium mills in Ontario began wito testw ht program n 197i s d 19798an . Currentlyl al , raffinate produce t Blina d d Rive s recyclei r o uraniut d m mills in Elliot Lake. 3.4 Summary e amoun Oveth e pas f wast. th ro tyear5 t so er o sfro m Eldorado's refinin d conversioan g n operations requiring waste management (bxirial s beeha ) n reduce r y centalmosb dpe .0 8 t Once the objective of recycling calcium fluoride is in place, over 95 per cent of what were previously considered wastes will have become by-products, used commerciall n othei y r operations.

165 THE REGULATION OF URANIUM REFINERIES AND CONVERSION FACILITIES IN CANADA

J.P. DIDYK Directorat f Fueeo l Cycle and Materials Regulation, Atomic Energy Control Board, Ottawa, Ontario, Canada

Abstract

e nucleaTh r regulatory proces t appliei s a s o uraniust m refineried an s conversion facilitie n Canadi s s reviewedai . e earlth yn I 1980s, Eldorado Resources Limited propose o construct d d operatan t e new facilities for refining yellowcake and for the production of uranium hexafluoride (UF^). These projects were subject to regulation by the Atomic Energy Control Board descriptioA . e Atomith f no c Energy Control Board's comprehensive licensing process coverin l stagegal f sitingo s , construction, operatio d eventuaan n l decommissionin f nucleago r facilities wa s tracest i i s a d applied to the Eldorado projects. The Atomic Energy Control Board's concern with occupational healt d safetyan h , with public healt d safetan h d witan yh protection of the environment so far as it affects public health and safety is emphasized.

Some regulatory difficulties encountered during the project's development which led to opening up of the licensing process to public input and closer coordination of regulatory activities with other provincial and federal regulatory agencie e describedsar .

e Board'Th s regulatory operational compliance progra U refinerie r mfo d san conversion facilities is summarized.

1. INTRODUCTION n passinI Atomie th g c Energy Control Act n 1946 I Parliamen1e th , f Canado t a declared atomi f nationaco energe b o t ly interes d thereforan t e undee th r exclusive jurisdiction of the federal government. The Atomic Energy Control Board (AECB) was created to administer the Act which provides for the control d supervisioan developmente th f no , applicatiof atomio e cus energyd nan . Under the provision e Act e Boarth ,th f o sd issue e Atomith d c Energy Control Regulations^ which determine the authorization and supervision regime

167 p i c i

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s_ i CO x: O

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5-mu£jOZJ u_i9 II Ii o oO« 1 C3 ïEa l or Q- ,—t Ï f O K n c o u lift * * g* Si ** s OO S • «K !_ - * » II i' 0) fi ü! W o « • H E o

1 |Cs« s Ï I •; - *• s R«Ïs 1 1 n Sg i i I h ^- E E II Ii £| CB )

1 l

168 applicabl l nucleaal o t er facilities e Boar Th s chose.ha d o issut n e only general, skeletal regulations; specific regulatory requirement e appliear s d through the licensing process.

. 2 STRUCTUR E AECTH BF O E n organizatioa e AEC Th s i B n consistin a Boar f f fivo go d e membera d an s supporting technical and administrative staff of approximately 265 persons. In the following text, the "Board" means the five-member Board and the "AECB" means the organization and its staff. (See Figure 1).

Various Division AECB'e th f so s Directorat f Fueo e l Cycl d Materialan e s Regulatio e responsiblar n r regulatinfo e g mine d uraniuan s m mills, refineries and conversion plants, fuel fabrication plants, heavy water plants, radioactive waste management facilities, transportation and packaging, and the use of radioisotopes . The Fuel and Heavy Water Plant Division (FWD) has responsibility r uraniufo m refiner d conversioan y n plant licensin d compliancan g e activities. Most licensing assessment and review work and facility inspections are carried out by project officers of the FWD who are mainly "generalists". They are complemented by staff specialists in radiation protection, quality assurance and a variety of engineering disciplines.

Not involved in licensing but reporting separately to the Board are two advisory groups, the Advisory Committee on Radiological Protection and the Advisory Committe Nuclean o e r Safety.

3. SAFETY PRINCIPLES AND OBJECTIVES OveAECB'e e coursth th r f so e existence safety principle d objectivean s s have been developed that underlie its regulations and licensing practices. The Board's Advisory Committee on Nuclear Safety has produced a document , which was endorsed by the Board, on the safety objectives for nuclear activities.

The underlying concept is that the primary responsibility for achieving a high standard of safety resides with the licensee. The regulatory agency in turn is responsible for providing the objectives for nuclear safety and the guidelines r theifo r applicatio r auditin fo s wel a ns a l g industry's performance. This approac he regulator leadth o t s y body havin o determint g w effectivelho e e th y safety objectives are being achieved.

. LICENSIN4 G PROCESS The licensing process is the means by which the AECB gains assurance that a nuclear facility wil e sitedb l , designed, constructed, commissione d operatean d d n complianci e with safety criteri d requirementan a s establishe AECBe th .y b d

169 This assurance is achieved by establishing communication with the applicant at an early stage in the project and maintaining surveillance over all safety-related activities in each phase of the project's develot lient from the initial conceptual e facilitdesigth f o n y througs maturit o t eh operatione Th . inteno identift s i t y point f contentioo s s earla t a a stagyn s possibla e o t e avoid situations wher e applicanth e t will fin t economicalli d y difficul o makt t e changes required by AECB staff.

1 Licensin4. g Stages For all nuclear facilities, the Board has established distinct administrative steps for the implementation of its authority. (Sec Figure 2). These consist of site acceptance which includes a public information program, approval to construc a licenc d an to operate t e e applicanTh . t oust convinc e Boari th e i- d detailed written submissions thae planth t t will meee healthth t , safetd an y security requirements of the Regulations. Details for these subi i ssior.s arc normally available in the fore of guidelines.

3 CONSTFUCIICN APPROVAL

E CLOSURSP 5' E

) I APPROVA STAGEN I L S

) 2 BURDE APPLICANE PROOF TH O N N O F T

INSPECTION AT ALL STAGES

OINT R1VIEW BY REGULATORY AGENCIES

17C Althoug e nucleath h r industr s subjeci y o federat t l jurisdiction througe th h Atomic Energy Control Act, it has to be realized that a nuclear facility operating withi geographicana l locatio a certai n i n n provinc f Canadao e s ha , some impact on that province and therefore provincial regulatory agencies do have a legitimate concern with regard to the operation of the facility.

In recognition of this joint interest, the Board has established a joint regulatory process. This means thae Board th a leat s da , agency, invitel al s regulatory agencies, federal and provincial, whose area of responsibility could e propose e impacteth b y b dn o dnuclea r facilit o participatt y a consultativ n i e e regulatory process. This process ensures that legitimate concerny an f o s agency, federal or provincial, are considered in the regulatory process and are reflecte e licenca conditioe for th f th o m n n i di e r requiremento n . Figur3 e outlines a Joint Regulatory Process chart as it might apply to a Site Assessment situation.

Submissio f Slteno j Applicatio o AECBt n j

AECB Licensing Applicant to Division Address Concerns

Health & Welfare Provincial AECB Security Canada Labour & Safeguards

Seller and Pressure Vessel]

Provincial Health

Provincial Environment

_L Site Assessment

Site Accepted by IRC

AECB Decision n Sito e Approval

Fig.3. Joint Regulatory Process Chart.

171 In many cases, the applicant may be required by an environmental agency to prepar n environmentaa e l impact statemen a forma d submio t an t lt i thearin r o g review process e imoacTh . t statement normally include e typ f informatioth so e n required by the AECB for site acceptance and can be submitted for that purpose. If a hearing is required by another agency the AECB withholds any site acceptance until the results of the hearing are available and evaluated.

4.1.1 Site Acceptance e basiTh c objective e Sitth e t a Acceptancs eo establis t stag e ar ee th h conceptual design of the facility and to determine whether it is possible to design, construct, and operate the facility on the proposed site to meet the safety objective d requirementan s s establishe AECB e e primarth Th .y b d y documentation required is a Site Evaluation Report providing a summary description of the proposed facility and its impact on the environment including information on land use, present and predicted population, principal source d movemenan s f watero t , water usage, meteorological conditions, seismolog d locaan y l geology.

During thi se licensin phasth f o e g process e applicanth , s requirei t o t d announce publicly his intentions to construct the facility and to hold public information meetings at which the public can express its views and questio e applicant'th n s representatives.

2 ConstructioA.1. n Approval Prio o grantint r a Constructiog n Approva AECe th lB mus e assureb t d thae th t desig s suci n h thaAECe th tB safety principle d requirementan s t s me wil e b l d thae planan th t t wil e builb l o appropriatt t e standards o n orded I .o t r s thisnecessari t i , y thae desigth t e sufficientlb n y advance enablo t d e safety analyse e performeb o t s d theian d r results assessed.

Construction will onl e authorizeb y d oncdesige th ed safet an n y analysis programs have progresse e pointh o t t de judgmen thatAECBe th th n o i ,n ,f o t further 'significant' design changes wil e requiredb l .

4.1,3 Operating Licence Before issuing an Operating Licence the AECB must be assured, primarily, that the plant, as built, conforms to the design submitted and approved, and thae plan th r operatiot fo s e satisfactoryar n e requirementTh . s include submission of a Final Safety Report, completion of a previously approved commissioning program, and approval of operating policies and principles.

172 . 5 AECB LICENSIN F ELDORADO G O RESOURCES LIMITED FACILITIES FOR uo3 AND UF6 PRODUCTION

Backgroun1 . 5 d Eldorado Resources Limited a federa, l Crown corporation s beeha , n minind an g refining first radium and then uranium for 50 years. At present the Blind River refinery, in Blind River, Ontario, processes yellowcake concentrates from uranium mines in Canada and other countries to produce UO^ • In Port Hope, Ontario, the UOß is further processed to uranium dioxide (UÛ2 ) fuel for CANDU reactor d als uraniuo an st o m hexafluoride (UFß r export)fo .

In light of a growing world-wide demand for UF^ in the mid 1970s, Eldorado proposed to build a second uranium refinery in Ontario with a capacity of 9,000 tonnes per year of uranium as UF^. Eldorado began the process of selecting a site for a new Ontario uranium refinery in 1975. By 1976, a long list of potential site d beeha s n narrowe o fourt d .

2 Environmenta5. l Review Processes e fouOth fr Ontario sites studied e sit Port th ,a e t m froGranbk m0 (2 Pory t Hope) was chosen by Eldorado for development of a refinery. Over the following yea a rdetaile d Environmental Impact Statement (EIS)- s preparewa 5 d an d submitted to a panel set up by the federal government's Environmental Assessment and Review Process (EARP). Following technica d publian l c EISe revieth , f o w the Port Granby refinery proposal was the subject of two rounds of public hearings e AtomiTh . c Energy Control Board participate e hearingsth n i d , providing technical analysis on Eldorado's proposal and explaining the Board's regulatory procese panel th n earlI o .t s y 197 e pane8th l concluded thae th t refiner ye refineritselth d an fy processes coul e environmentallb d y acceptable n appropriata n o e a numbesit f i ef condition o r s were mete PorTh .t Granby site, however, was found to be unacceptable for a variety of reasons related to air quality, waste management, land use and social impacts.

Following the rejection of the Port Granby proposal, Eldorado identified potential sitePore th n eact i s f Hopo h e (Hope Township), Sudbur d Blinan y d River areas of Ontario. The Hope Township site ranked highest in Eldorado's consideration. Following environmental hearings on all the sites, the Panel's report s issue^wa Februaryn i d , 1979, concluding thal threal t e sites coule b d acceptable for the project. In July, 1979, it was announced that the Federal Cabinet had concurred with Eldorado's selection of Hope Township as the preferre e Ontaridth sitr fo oe refinery. Site preparation work bega n earli n y• s suspendewa 198t bu 0 d wheFederae th n l Cabine a newl f y(o t elected government)

173 reviewe e earlieth d r decisio d determinean n d thae refinerth t y shoul e locateb d d t Blina d River. Eldorado then proposed thaUO-e th tj production facilite b y sited at Blind River but that the UF^ conversion facility be sited on the old Port Hope Plant property.

Eldorado publicly announced its proposals and the site applications for AECB 7 approval of the Blind River refinery and the Port Hope UF6 conversion facility** were filed on September and October, 1980, respectively. With respect to the subject of environmental assessment, Eldorado chose not to submit w proposa ne e EAR e th P th o t processl e BoarTh .d staff indicated thaa t specific environmental assessment process (non public hearing process) would be necessary and that this would be part of the Board's joint regulatory process.

The Blind River site although changed slightly from the original site proposed, presendidnot a probletAECB the genera the , to m l area having been extensively reviewe y regulatorb d y agencie d accepte e EARan s th P y b Paneld e PorTh .t Hope site, however, had not been subjected to any environmental assessment process and further, it did not provide for a "buffer zone", an area surrounding a nuclear facility whic s unde i he contro th ra licensee f o l .

LLETTER-OF-INTENT

Public Environmental Hearing; AECB Required Process! Format Public Meeting Format

AECB Evaluation Formatiof no S oEI f IRC

Participation in Site Application Hearings to AECB

Panel Accepts Review and Assessment Proposed Site of Site Application

Site Application to AECB

Acceptance of Site Application ______bv IRC______

Preparatiof no Board Member Document \( AECB Decision o n Site Approval Fig.4. Site Approval Licensing Process Chart,

174 Figure 4 shov the difference in the two approaches (public environmental hearin s AECv g B public meeting format with joint regulatory review).

5.3 Licensing of the Blind River Refinery IP. September I960, Eldorado applied formall e AECr approvath fo B o t y o sitt l e the Blind River refinery on a 253 hectare site in the town of Blind River, Ontario. Board staff considered Eldorado's change froœ the original proposal.

5.3.1 Site Assessment The regulatory reviews of the Eldorado Site Application' and other documents e Interagencpresenteth e AECd th an B o t d y Review Committee, identified a nussber of deficiencies in the reports and requests for further clarification were submitted to Eldorado. Eldorado's responses to these requests were evaluate d considerean d d acceptabl e AECth By b estaf d IRCan f . The AECB staff recommended to the Menbers of the Board, i r. January, 19S1, that site approval should be given.

5.3.1.1 Public Intervention Followin e government'th g s decision that Eldorado establis a uraniuh m refinery at Blind River, Ontario, some members cf the public, from the ccw~ of Blind River and surrounding area, claimed that the eranges in Eldorado's plan-ing warranted new public hearings. Conversely the Board staff believed that the change in site location wa r= not sufficiently significan o requirt t ew publithane a t c hearin e carrieb g d out.

Ir. November, 1980 e Boar.th d receive a submissiod n froe Blinth m d River d Districan t Concerned Citizens Association reviewing Eldorado's site application. Although a number of their concerns had merit and vere considered by Board staff in its evaluation, no new technical concerns were identified which would alte e Boarth r d staff positions it t A . January, 1981 meeting, the Board agreed to receive representatives from two groups. These groups were strongly opposed to nuclear energy in general and to the establishment of the Blind River refinery in particular.

After hearing the groups' presentations, the Board decided to defer a decisio n sito n e approva o allot l w staf n opportunita f o studt y y last minute submissions and to discuss the concerns in person with all groups e Blinth n di River area e AECTh .B attende a seried f meetingo s s with various public groups including local government elected officials. The AECB made presentations on the Board's mandate, the refinery licensing

175 review wated procedurean re r issue th qualityai d f o san s , radioactive waste d radiatioan s n effects s wela ,s respondina l o questiont g d an s listenin o publit g c concerns. Following these meetings e AECth ,B staff conclude informatiow d ne tha o n t n bearin n mattero g f healtho s , safety, securit e environmenth r o y d beeha tn presented which would altes it r original recommendation that the site be approved for the purposes intended.

In mid-February, 1981, the AECB announced that it had approved Eldorado's application to site a uranium trioxide refinery near Blind River, Ontario.

Some details on the public participation and input during the course of e Blinth d River refinery site evaluation proces s i presentes d hero t e provid n indicatioa e e extenth f o whict o n te AECB'th h s licensing process has become a more open one. In 1970, when an UF^ Plant in Port Hops licensewa e o operatt d e almos o publin t c informatio s madwa ne available on the AECB approvals and no public participation in the licensing process was contemplated.

5.3.2 Construction Authorization and Operating Licence n FebruaryI , 1981, Eldorado applie r Constructiofo d n Authorizatioo t n build the refinery. Information was presented to the AECB on facility design and layout, process description, safety programs, accident analysis, emergency procedures and quality control programs for design, procurement and construction. Eldorado's application and supporting information was reviewe e AECth y membeBy b b d staf d an r f agencie e Interagencth f o s y Review Committee. (See Figure 5).

It was noted that the refinery was designed to recycle as many wastes stream s possibla s o minimizt d an e numbe th e f dischargo r e points. This design approach combined with monitorin d controan g l programs proposey b d Eldorado would enable detection and corrective action to be taken to limit the effects of upset situations to the environment.

A detailed accident analysis indicated tha o credibln t e scenarios were identified which did not have acceptable design features included to mitigate their consequences.

176 Meeting with Applicant

Submission of Construction Application to the AECB

Revie Assessmend wan t of Construction Appllcatior

Construction Application Approved by IRC

Approva f Desigo l n Criteria and Design Control Program

AECB Decision no Construction Application

Fig.5. Construction Approval Licensing Process.

Construction Authorizatio e constructioth r fo e nuraniu th f o nm trioxide refinery n Blini d River, Ontario s grantee Boarth wa , n July i y db d , 1981.

n AugustI , 1983, constructio d commissioninan n e refinerth f o g y having been completed e AECth , B granted Eldorad n operatina o g licence. Leadin o thit p su g approval, the AECB Btaff had reviewed all the necessary commissioning reports on l majoal r safet d environmentaan y l control system d foun e systeman sth d s wern i e satisfactory conditio r start-upfo n wells A . , radiation protection procedures for the plant were reviewed and some changes made as requested by the AECB staff. A Derived Release Limit (DRL) document was prepared by Eldorado and reviewe e AECth By b d staff e planTh t. design allow e planth so operatt t t a e less than 1% of the DRLs.

4 Licensin5. e Porth tf Facilitieo gHopg UF e s n OctoberI , I960, Eldorado applied formall e AEC r th approvafo Bo t y o expant l d the uranium hexafluoride conversion facilities to include a new plant with an annual capacit f 9,00o y 0 metric ton f uraniumo se tow th f Poro nn i t, Hope adjacent to its existing Port Hope facility. An Interagency Review Committee, consisting of the AECB, Health and Welfare Canada, Environment Canada and the Ontario Ministr e Environmentth f o y s formewa ,o reviet d w Eldorado's application.

177 5.4.1 Site Assessment A document "Applicatio e Atomith o ct n Energy Control Boar r Sitfo de Approval

- Port Hope UF Facilities Expansion, October 1980", was submitted to 6 n suppore siti th C e f IR o approvate AEC d th an B l application e documenTh . t 8 e AEC s baseth Bwa n o requirementd d guidelinean s s specifie Atomie th n ci d Energy Control Board licensing documents.

Some of the significant issues identified are presentee' below.

e regulatorTh . 1 y agencies agree de propose th tha f i t d sits acceptedwa e , future approvals for the operation of the new UF^ Plant would be contingent upon the existence of a clearly defined waste management plan.

Eldorado propose w facilitieo designe t d e th n o reduct s e quantitth e f o y waste a minimum o t sr exampleFo .w proces ne ,a s incorporated inte th o design of the new UF^ facility allows recycle and recovery of hydrofluori n eight-folca acid an d d reductio n CaFi n 2 generated. Eldorado d alsha o made significant progres e recoverth n d recyclini s an y f wasto g e materials.

e a larglac Th f o ke . d 2 "buffean C IR r f concer e o zone th s o wa "t n therefore, the conceptual design of the facility was reviewed to ensure that appropriate design measures were propose o compensatt d e smalth lr fo e buffer zone availabl o ensurt d an e thae publith t ce expose b woul t o no dt d unacceptable levels of radiation and hazardous substances during normal and upset processing conditions.

Eldorado proposed to incorporate a number of technical features into the plant design which would give an increased measure of confidence in its ability to contain releases as follows: l effluenal - s scrubberga t d dusan st collectors wil e installeb l d inside the structure of the operating plant, - secondary scrubbers will be installed to handle upset or emergency situations, F unloadinH l al - g from railcars n enclosewila e carrien b li d t an d ou d ventilated building e processinth o o transfet N . F H g f areao r s will take place without secondary containment, - all liquid UFg handling (filling of shipping containers and sampling operations) will take place in autoclaves and inside the processing buildings,

178 - emergency electric power maintain e operatio th sy emissio ke f o n n control systems.

Eldorado had designed additional environmental and engineering features into the proposed UFt facility to provide an adequate margin of in-plant safety systems, without reliance on the "buffer zone".

5.4.1.1 Public Information Eldorado satisfactorily met the AECB's requirements for conducting a public information program. This program culminated with the holding of two public meetings in Port Hope in November and December, 1980. In general, the Port Hope public responded favourably to the idea of the Eldorado facilities being expanded in their town.

In July, 1981 the AECB approved the Port Hope site based on a recommendation from the AECB staff. The Board required Eldorado to carr t somou ye further studiee atmospherith n o s c dispersion modelling information presented in its application for site approval, as well as studies of groundwater flow on the site, as conditions of the approval.

5.4.2 Construction Authorization In December, 1981, Eldorado submitted an application10 to the AECB requesting approva o construct l e proposeth t d facility e AECTh .B stafd an f C assesseIR e e informatioth th d n presente y Eldoradb d o assurt o e th e reviewers that the design of the proposed facility would meet the safety and environmental requirements of the AECB and other regulatory agencies.

n AprilI , 1982 Eldorad s grantewa o d partial approva o proceet l d wite th h plant construction. Eldorado was directed not to proceed with work on the HF unloading and storage area until further review on these facilities was carried out.

As well as the review of Eldorado's construction application ^ by the AEC d IRC AECan e BEnvironmend ,th an B t Canada contracted with engineering consultants to review Eldorado's proposed methods of bulk handling and storage of HP and NH^ and to review the submitted accident analysis for possible omission f significano s t events.

n MayI , 1982, after e informatioconsideratioth f o l al n f mado n e available e Boarth d, tit stafo f recommended that Construction Authorizatioe b n granted for all those remaining plant components which were not covered in

179 the partial construction approval grante Apriln i d , 1982 e AEC.Th B agreed o grant t this approval.

5.4.3 Technical Considerations of the Regulatory Review

5.4.3.1 Siting Considerations A number of issues were identified in the review of Eldorado's Siting Application which required satisfactory completion or further studies to be carried out in order to obtain additional design information to allow a complet r fo e evaluatio f Eldorado'o n s proposa o construct l e th t facility e followinTh . g summarize more s th some f o significane t issues :

(a) The review of the atmospheric dispersion modelling used in the Site Application® led to the conclusion that the modelling may not be appropriate to the Port Hope site. Eldorado was requested to carry out validation tests and a sensitivity analysis to show that e atmospherith c dispersion modelling information presentee th n i d Site Application was valid for the Pore Hope area.

Eldorado performe e requesteth d d word Boaran k d staff concluded that the information contained in the study was adequate to satisfy the requirement of model validation and sensitivity analysis of the atmospheric dispersion modelling used.

o (b) The Site Application did not contain sufficient information to substantiate the claims that the groundwater flow from the proposed plant site was not in a southwest direction towards the Town of Port Hope's Water Treatment facility which contains a number of underground water storage chambers.

Eldorad s requestewa o o conduct d a furthet r stud o determint y e th e groundwater flow direction.

The results of the study were used to determine whether additional containment facilities were require e plan th e typ tn th o de sit d an e f contingenco y plan d monitorinan s g program e developeb o t s r fo d plant operation.

5.4.3.2 Construction Review e EldoradTh o Construction Applicatio s reviewewa n d with particular emphasis placed on the design criteria, engineering design and proposed

180 operation of those pieces of equipment and systems which have the greatest potential to cause significant releases of radioactive and toxic non-radioactive substance e workplacth o t sd publian e c environment. The plant safety systems were assessed to ensure that adequate design criteria were used in their design to mitigate the effect f normao s d abnormaan l l plant operations. Also e Qualitth , y Assurance programe designth r fo ,s procuremen d constructioan t n phases of the project were assessed to ensure that the necessary controls were being implemented.

The following is a brief summary of the major items reviewed:

e gaseouTh ) a s effluent treatmen^ facilitUF w ne yte systeth r fo m consists of a series of large potassium hydroxide scrubbers and a number of dust collection systems. The scrubbing systems, designed for high efficiency fluoride removal, are arranged such that they e operateb n ca n severai d l serie d parallean s l configurationo t s allow for plant upsets and to permit on-line preventative maintenance programs to be conducted.

In order to provide additional containment in the event of an accident F storagH e th , e building 2 Cel? , l maintenance ared an a g producUF e th t handling areas have separate, independent ventilation systems.

Eldorad d proposeha o n abatemena d t system which shoul e ablb do t e adequately control air emissions.

^ producUF e Th t handlin) b g operation e facilitth f e o s ar y located in areas physically segregated from the other plant processes. The UFc transport cylinders are filled in an area with n emergencow s it y ventilatio^ nUF l samplin systet al ho d f an mo g cylinder dons i sn autoclaves i e .

F unloadinH e e desigth Th d storag f ) o nan gc e facilit majoa s rwa y review item for Board staff. Eldorado had proposed to build an HF facility which incorporates a structure to totally enclose two storage tanks, a below grade emergency storage tank and the rail car unloading station. This structure is ventilated and any HF emissions are scrubbed before release to the environment.

181 5.4.4 Operating Licence By May, 1984 Eldorado had successfully commissioned all the utilities, major process plante areath d safetf an o ,s y system d requestean s d authorization o proceet d with start-u e plan th produco 't f to p e UF^ A positiv« e recommendatio e operatin th e Boar d th s mad an dwa no t ge licenc s approvedwa e . An operating licence containing nineteen conditions of operation was issued

to 6 , 1984EldoradoPlanUF 28 Porn w operato y i ,tt ne Ma t e n Hopeo ,th e .

5.4.4.1 Technical Considerations ) Boar(a d staff conducte a desigd nprocese audith f o ts safety systems and considered them acceptable.

) Durin(b e constructioth g n phase, audit f qualito s y assurance th n o e s a distributio& 2 liqui? , HF dd liqui an ng storag UF d d handlingan e system wer e AECeth B carrie y b staf d Ministrt an fou d f Consumeo y d an r Commercial Relations (MCCR) staffs. All of these systems were jointly approve d accepte e MCC an dd AEC th an R y B b dstaff .

(c) AECB staff reviewed commissioning data and documentation related to l safetal y systems A detaile. d revie f environmentao w l monitoring procedure s conducteswa d accepte Ontarie an d th y b do Ministre th f o y Environmen d Environmenan t t Canada staffs.

(d) Eldorado prepared detailed operational quality assurance manuals and procedure o controt s d maintaian l n minimum standardl al n o s safety-related aspects of plant operations. In addition, a program had been set-u r in-servicfo p e inspectio minimizo t n e probabilitth e f o y failure of piping and equipment.

) Detaile(e d emergency procedures have been prepare Eldorady b d d an o reviewed by Board staff and found to be acceptable.

A derive ) (f d release limit documen s preparewa t d includean d d datn o a the plant emissions and discharges. This was reviewed by Board staff and is considered acceptable. Eldorado was obliged, as a condition of licence, to maintain the airborne uranium emissions from the total Port weightee Hopth f eo dfacilitm DRL'ssu e o lesth t y, f sbaseo tha% d10 n on weekly averages. A gamma component was later added to the DRL control equation.

182 ) Boar(g d staff reviewe e operatoth d r training plane prograd th an tr fo m considered it acceptable, including the training and operating procedure for the anhydrous HF acid storage area.

basie th f thio sn o s s comprehensivIwa t e revie d assessmenan w t thae th t AECB staff recommende e dpermitteb theL ER t o commenct d e operatiof o n e planth o product t e UF^-

6. AECB Licensing Compliance and Surveillance Compliance and surveillance of the licensee operations is carried out to determine whether the requirements of the Atomic Energy Control regulations and the conditions specifie e facilitth n i d y licenc e beinar e g met. Issued an s items relatin o complianct g e reviewear e d assesse an e assigned th y b d d project officer with assistance froa specialists in other divisions within the AECB and outside agencies. Staff from provincial agencies may be, in some cases, appointed AECB inspectors for purposes of inspecting the licensees' facilities and for auditing the licensees' monitoring results- These inspections are done on behalf of these agencies themselves although reference may be made to the AECB licence. The AECB staff exercises a senior auditing function in cases where other agencies are involved.

Compliance reviews, assessment d recommendationan s y AECb s B staf d outsidan f e agencie e submitte ar s AECe r consideratioth fo B o t d n takini n g regulatory action and as factors during licence renewal periods.

References

1. Atomic Energy Control Act 1946, as amended 1954. RSC 1970 c A-19. 2. AEC Regulations CRC 1978 c 365, as amended SOR/78-58, SOR/79-422. A Propose . 3 d Statemen n Safeto t y Objective Nuclear fo s r Activitie n Canadai s , ACNS-2, June 1981, AECB INFO-0055. 4. AECB Licensing Document No. 35, Fuel Facility Licensing, Part I - Regulatory Polic Licensind an y g Procedures, February, 1985. - AECB Licensing Document No. 35A, Guide To The Licensing of Uranium Refineries and Chemical Conversion Facilities, September, 1978. 5. Environmental Impact Assessment, The Port Granby Project, Eldorado Nuclear Limited, September 1977. 6. Report of the Environmental Assessment Panel, Eldorado uranium Hexafluoride Refinery, Ontario, Federal Environmental Assessment Review Office, Government of Canada, February 1979.

Î83 7. Environmental Tnpact Statement for a Uranium Hexafluoride Refinery, Blind River, Eldorado Nuclear Limited, September 1978, preparey b d Jame . MacLareF s n Limited. - Application to the Atomic Energy Control Board for Site Approval, Blind River Refinery, Eldorado Nuclear Limited, September 1980. . 8 Applicatio e Atomith o ct n Energy Control Boar r Sitfo de Approval, Port Hope UF^ Facilities Expansion, Eldorado Nuclear Limited, October 1980. . 9 Applicatio e Atomith o t cn Energy Control Boar r Constructiofo d n Approval, Blind River Refinery, Eldorado Nuclear Limited, February 1981. 10. Application to the Atomic Energy Control Board for Construction Approval, Port Hop ^ FacilitieeUF s Expansion, Eldorado Nuclear Limited, November 1981.

184 CONVERSION OF URANIUM ORE CONCENTRATES AND REPROCESSED URANIU NUCLEAO MT R FUEL INTERMEDIATES AT BNFL SPRINGFIELDS Par : UraniuA t concentratee mor s

H.PAGE Springfields Works, British Nuclear Fuels pic, Preston, Lancashire, United Kingdom

Abstract

Uranium processin s bee ha ge Springfield nth carriet a t ou d s Work 0 f yearsBNF4 o s r fo L. During that time some 85,00U t 0 equivalent of uranium ore concentrate and in excess of 15,000 t U equivalent of uranium recovered from spent fuel reprocessing have been converte o nucleat d r fue r nucleao l r fuel intermediates at Springfields. This paper describes how the conversion facilities have evolved in response to changes in factors such as market demand, feed and product specifications, industrial hygien d environmentaan e l constraints, regulatory practices etc. Comments on the adaptation which wile necessarb l o deat y l with anticipated changee th n i s nuclear fuel cycl e providedar e .

1. INTRODUCTION

From its Headquarters at Risley, near Warrington, in the north west of England, British Nuclear Fuels pic (BNFL) co-ordinates the activities of its three operating divisions which between them provide a comprehensive range of nuclear fuel cycle services e activitie Th e thre.th f eo s divisions e summarisear s followsa d : e manufacturTh . i f nucleao e e Springfieldr th fue t a l s Works of BNFL near Preston in Lancashire is the responsibility of Fuel Division. e enrichmenTh . ii f uraniuo te Capenhursth t ma t Works of BNFL near Chester is the responsibility of Enrichment Division.

185 e reprocessinTh . in f irradiateo g e dSellafiel th fue t a l d Work f BNFo s n Cumbrii L e responsibilitth s i a f Reprocessino y g Operations Division which is also responsible for operating two Magnox type nuclear power stations, one at Chapelcross and the other at Sellafield.

The aim of this part of the paper (Part A) is to describe current technology for converting uranium ore concentrates (UOC o nucleat ) r fuel intermediate t Sprmgfielda s o t d an s comment on recent or proposed changes in this technology. The conversion of reprocessed uranium at Sprmgfields is dealt with m Part B.

conversioC UO In t ordepu n o t technologr y inte th o context of Sprmgfields site activities as a whole it is useful to summarise the main site activities which are:

1. Conversion of UOC to natural uranium tetrafluoride (UF ), 2. Conversion of Magnox reactor depleted uranium (ViDb) o reactot r depleted UF,. 3. Conversion of a. Natural UF, to uranium hexafluonde (UF,) for 4 o enrichment at overseas enrichment plants or at Capenhurst b. Reactor depleted UF, to MDL UF, for enrichment t Capenhursta .

4. Reduction of a. Natural UF. to uranium metal for fabrication of Magnox fuel , Italiaelement d UK Japanes an nr fo s e reactors. . b Tails deplete o tailt sF U depleted d uranium metal for non-nuclear applications. . 5 Conversio f enricheo n o ceramit , UF d c grade uranium dioxide (UO-,) for fuel fabrication plants at Sprmgfields and overseas. . 6 Fabricatio f oxido n e fuel assemblie r Advancefo s s Ga d Cooled Reactors (AGR) and Light Water Reactors (PWR and BWR). 7. Manufacture of cladding components for Magnox and AGR fuel elements. 8. Recovery of uranium values in residues recycled from uranium conversion and fuel manufacturing activities.

186 2. UOC CONVERSION PROCESSES. RECENT AND PROPOSED DEVELOPMENTS

Brief descriptions of the principal conversion processes s existhe w a e provideno yar t t recenorden i pu d o t rt developments into appropriate context.

1 Conversioo 2. Purt C e UO Urany f o n l Nitrate The uranium purification process at Springfields is based on continuous dissolution, filtration and solvent extraction ia nnitri c acid medium using counter curren x mixebo t r settler contactor a 20 d % an sTBP/O K solvent e existinTh . g technology has evolved over a period of 30 or more years e replacementth y b , refurbishmen r stretchino t f equipmeno g t originally installed to meet the demands of the first UK civil reactor programme (Magnox). Since that time some 85,00U t 0 of uraniu e concentratemor s representin a gwid e variet f Yelloo y w Cake type d originan s s have been processed through the purification plants. The flexibility e procesth of s with respec o accommodatiot t e width f eo n rang n Yelloi e w Cake qualit s stili s beed l ha y an sucn s a h to provide little if any incentive for introducing fundamental changes.

2.2 Conversion of Pure Uranyl Nitrate to Uranium Trioxide In all essential respects the Springfields process for producing natural uranium trioxide (UO, s )similai o that r t which was later adopte t Sellafiela d r producinfo d depleteU MD g d UO-. A descriptio e Springfieldth f o n s process follows.

\ e physicaTh d chemicaan l l propertie e uraniuth f o s m trioxid P productioU ee th fee o t dn process have a crucial bearing on the performance of that process, and some comment n theso s e propertie d theian s r origi s relevanti n .

UO- is prepared from pure uranyl nitrate by a two- stage continuous process comprising climbing film evaporation followed by fluidised bed denitration. The essential features of this process are summarised below.

187 The product of the purification stage, pure uranyl nitrate s pumpei ,e tubth e firso t df th fousid o tf o er long- tube natural circulation evaporation units in series. The liquor level in each of the four evaporator stages is maintained at approximately /o of the tube height by automatic control of the liquor flow into each stage. Discharge of concentrated uranyl nitrate product froe lasth mt stage, which operates under vacuum (<~350 mm Hg abs), is automatically controlled so as to maintain a concentration of approximately 110% w/v 'U' using boiling point elevation to signal product strength. Variatio f evaporatoo n r throughpu s achievei t y variatiob d n of the steam pressure applied to the first evaporation stage over the range 3-5 Kgm/cm2.

Steam condensate froe firsth m t evaporation stagd an e e condenseth d vapour froe othemth r three stage e usear s d for pre-heating the feed liquor before the mixed condensâtes are recycled to the filtration and purification stages. The thermal conservation resulting from the four-stage evaporation and condensate/feed heat exchange is such that the amount of water evaporated is triple the quantity of steam use r evaporationfo d .

The molten product of the evaporator is then denitrated ia fluidisen d reactobe d o givt r e uranium trioxide th n i e form of a free flowing granular powder.

e demtratioTh n reacto a vertica s i r l cylinder wita h conical base through which protrud fluidisin1 2 e r nozzlesai g . e reacto , fluidise Th t airUO ho rf n s ,o i chargeo d d be a , maintained at 300°C by electrical heating elements in tubes through the cone base and in jackets around the walls of the cylinder. Molten uranyl nitrate is pumped into the a spra n fluidisea locatevi gu y d dbe dabov e internath e l heating element belo3 e surfacd e fluidisewth th an s f o e d ! bed. The UO., particulates formed within the bed are removed continuousl e overfloth y b y w froe surfacmth d pneumaticallan e y transporte o storagt d e hopper t actina lifa po a t s vi a sg seal betwee e reactoth n d storagan r e hopper e fluidisinTh . g

188 air and gaseous reaction products, steam and oxides of nitrogen, are routed via sintered stainless steel filters to the nitric acid recovery plant which consist a pre-absorptio f o s n condense a serie d an rs arrangemen x absorptiosi f o t n towers stacked with stoneware rings. Water is fed to the last absorption towe a rat t ea r calculate o product d a efina l acid strength of 40% w/w H NO., which is blended with purchased acid for use at the dissolution stage. The final traces of oxides of nitrogen are removed from the acid absorption plant exhaus y scrubbinb t a towe v n i gw/ r % irrigate10 a y b d solutio f caustio n c soda.

The reliability and availability of the principal items of process plant in the UN to UO., sequence, most of which are constructed of AISI grade 321 stainless steel, has been such that all of these items are still in regular use after more tha year0 2 n f operationo s . Over that time productioth e n capacity of the denitration units has been stretched considerably in uprating exercises involving increase of bed height and heat transfer surfac o allot er increaswfo f electricao e l power e higheth t A r . ratingKw e 0 th s80 ratin o t w g K iro0 m30 heat flux is such that precise and extensive monitoring and contro f heao l t transfer surface temperatur s requirei e d to avoid damage to the product or overheating of the reactor barrel.

The evaporation/demtration uprating w exercisno s ha e been completed to the stage at which the total Springfields capacity is 11,000 t U /year with an option to increase to 12,50 /yeaL t 0 r when justifie y demandb d .

The product of the denitration stage, of which upwards of s 100,00beeha nU t mad0 o datet a efree-flowin s i , g powder with properties typical of those listed in the following table:

189 Typical Properties of DO-, Produced by Fluidised Bed Denitration

Crystalline form Tf or Type III oxide Appearance Aggregate f onioo s n skin type particles Mean particle diameter 130 microns Tap density 4.5 g/cc Angle of repose 40° Specific surface area m2 0. 2 /o gt 1 0. Sodium content 10 ppm 'U' basis NO content 0.5% w/w Moisture content w 0.05w/ %

2.3 The Rotary Kiln Uranium Tetrafluoride (UF ) Production Process For approximatel year0 3 y s fluidise d technologbe d y was used at the reduction and hydrofluorination stages of uranium tetrafluoride manufacture, both natura d depletedan l , t Springfieldsa . Followin e successfuth g l introductiof o n rotary kiln technology in 1978 it was decided that the less economi d moran c e hazardous fluidise d planbe d t shoule b d phased out and replaced by a second kiln plant which is now under constructio d schedulean n r commissioninfo d g n 1987i .

2.3.1 Hydratio f Uraniuo n m Trioxide. e volumTh e expansion which occur s, particl wheUO a n e is subsequently converted to UF. can create the situation where hydrofluorinatio th e rat th f eo e n reactio s limitei n d e diffusioth y b d f frogasee o F UO,an nreactiomth U o t s> , - n site unles e precursoth s , particlUO rs sufficienha e t innate open pore volum o accommodatt e e expansionth e e th n I . e higth cas hf o edensity surfacw lo , e , particleareUO a s prepared by fluid bed denitration, the required open pore volume is easily obtained by reaction of the UO, with water under conditions which lead to rapid and complete formation e uraniuoth f m trioxide dihydrat a significan d an e t expansion in particle volume. Subsequent decompositio e dihydratth f o n e then yields a particle with a lattice expanded sufficiently

190 to accommodate the volume change in the conversion of

UO3 to UF4.

UO, produced in the denitrator is transported pneumatically to one of a pair of feed hoppers from whence it is fed to the hydrator at a controlled rate co-currently with the stoichiometric quantit f water o ye hydrato Th . a jackete s i r d stainless steel trough agitate n interruptea y b d d screw conveyor. Precise control of residence times and temperatures in specified section e hydratoth f o s r ensure e formatioth se requireth f o n d hydrate which discharges inte feeth o d e nexhoppeth r t fo r y free-flowindr staga s a e g powder with properties typical of those e followinlisteth n i d g table.

Typical Propertie f Uraniuo s m Tnoxide Dihydrate

Mean particle diameter 200 microns Density 2.7 to 3.0 g/cc Angle of repose 65° Specifi 6 m o 2t c/ g4 surface area (after decompositio t 450°Ca n )

2.3.2 The Conversion of UO, to L F.. Early in the study of the dryway reduction and hydrofluorination reaction it became evident that temperatures high enough to cause sintering or solid surface modification could readil e attaineb y t localisea d de reactin th site f o s g solid. This effect occurs during the reduction reaction and e initiaith n l stagee hydrofluorinatioth f o s n reactionn I . either case the result can be incomplete conversion at the hydrofluorination stage due to what is frequently termed deactivation. Factors which hav n importana e t bearinn o g the deactn ation process are the rates of heat release and heat removal from the reacting particle. The former is a functio f factoro n s which influenc e rat th f ereaction o e , le temperature, reactant concentrations, surface properties, etc, whereas the latter is dependant on heat transfer by conduction withi e particly convectiob th n d an ed radiatioan n n froe particlth m e withi e reactinth n g mass. Interactiof o n these variables call r verfo s y careful contro f reactioo l n

191 conditions to ensure reproducible production of high grade UF-, a level of control which has been achieved in the rotary kiln form of contactor.

2.3.3 Common Feature e Rotarth f o sy Kilns. Each rotary kiln system can be considered as comprising:

powdeA . i r feed off-gaan d s filtration hopper rotatinA . ii g kiln barrel iii. A powder discharge hopper

e powdeTh r feed hopper also contain e offgath s s filter clusters for removal of particulates from the exhaust before s dischargit e condenseth o t e d scrubbean r r systemn A . automatic valve and fan sited in the exhaust gas line downstream e scrubbeoth f r control mba5 e 1 pressur. th g s^ re kil t th a n n i e

The kiln barrels are supported at each end on riding rings restin n rolleo g e driverb bearingn t speeda nca d an s in the range 1 to 8 rpm by one of a pair of electric motors, or, in an emergency, by a diesel powered motor.

Axial and radial movement of powder within the kiln a functio s i f rotationao n l speed, kiln angle, volume throughput, the properties of the powder and the number, size and shape of the internal flights and dam rings. The influence of these and other variables such as reactant composition and temperatur e optimisear e o ensurt d e productio a produc f o n t suitable in all respects for downstream processing. Control of the reaction temperature profile, which is of vital significance in this regard s achievei , a combinatio y b d f electricao n l heating, force r coolinai d d comprehensivan g d precisan e e monitorin f elemento g , kiln wald reactanan l t temperatures.

Seal assemblies designed to prevent egress of the kiln content e locatee powdear sth t a d r feed dischargan d e ends where leak detection instrumentatio s alsi n o sited. Continuous measurement of kiln alignment and seal condition is carried out to maximise seal life.

192 The content e powdeth f o s r discharge hoppere ar s periodically transported pneumatically on N-, to the feed hopper e nexth t e a processsstag th f o e .

Kiln bed thickness by diametrical gamma scanning, and residence time distributions, by radioactive tracer injections, have been determined on both kilns for a variety of operating conditions. This typ f informatioo e s beeha n n helpfun i l designing a dynamic simulation model which takes account e majoth of r process variable d considersan s n particulari , , reaction rates, powder transport mechanisms and all aspects of heat transfer and kiln temperature control.

2.3.4 Reductio f Uraniuo n m Trioxide. Metered quantitie f hydroge, hydrato s UO e d ar e an n contacted counter-currently in a stainless steel rotary kiln y precisb d an e controe temperaturth f o l d concentratioan e n of reactants complete conversion to UO,, : > achieved without Ct los f activityo s .

The hydrate feed rate is regulated by a screw feeder at the hydrate hopper outlet which maintains the hopper contents at a constant level to establish a positive seal between the kiln and the hydrator. The gas space in the hydrate hopper is continuously monitored to detect and respond to hydrogen levels about a preset value.

Exhaust gas from the reduction kiln, containing steam, excess hydrogen, nitrogen and traces of nitrogen oxides s dischargei o atmosphert d a condenser a vi e , scrubbed an r flame arrestor.

The typica 7 produc e reductioU0 lth f o t n na kils ha n Cj oxyge nuraniu: m rati f 2.0a specifio d 2an c surface area o3 mf 2/g.

2.3.5 Hydrofluorinatio f Uraniuo n m Dioxide. Metered quantitie f liquio s d anhydrous hydrofluoric acid are vaporised at "•>20°C and fed into an inconel rotary

193 kiln wher s contactei t i e d counter-currently with metered

quantities of UO2 to produce high grade UF (>98% UF4) at a high AHF usage efficiency (^95%). To prevent thermal damage to the basic particle leading to deactivation, incomplete conversio n extremi r o n e cases, sinterin e bulth kf o gmaterial , careful profilin e reactanth f o g t temperatures froe powdeth m r e powdeth feo t d r discharg ee kil endth s essentiali n f o s .

The hydrofluorination kiln exhaust gas containing steam, nitroge d excess routei an n a F wate H o st d r cooled monel condense e condensatth d an r e) collecte (10-15HF w %dw/ in rubber lined mild steel tanks before on site blending for resale to the trade. The incondensibles are scrubbed with caustic soda solution before discharg o atmospheret e .

e materialTh f constructioo s e maith nr fo itemn f equipmeno s t e rotarth at y kiln hydrofluorination stag e milar e d steel, Inconel, Mone d rubbean l r lined mild steel. Polyvinylidene fluoride pipewor d valvean ke use r ar diluts fo d e hydrofluoric acid handling in the liquid phase. By comparison with the fluid bed route the elimination of the need for HF recycling and lower temperatures and pressures involved in the kiln route has virtually eliminated the corrosion of nickel based alloys which is a prominent feature of the older route to

UF4'

2.4 Conversion of Uranium Tetrafluoride to Uranium Hexa fluoride Uranium hexafluoride (UF,), both natura d Magnoan l x reactor depleted, has been produced at Springfields for

more than 20 years by the reaction of UF. with F? in fluidised , productiod reactorsUF be o Tw n. unit e currentlar s y available: The first, commissioned in 1968, has a capacity of 3,000 t U per year and a second larger unit, commissioned in 1974 a capacit s r ha year o pe datef 6,50 o yT U . ,t 0 more than 60,000 t U as UF, has been produced, most of which on behal f overseao f s utilities e provisioTh .a thir f o nd unit, r conversiofo f uraniuo n m arising froe reprocessinmth g of irradiated oxide fuel is discussed in Part B of this paper.

194 2.4.1 Fluorine Production. Description e Springfieldth f o s s facilitie r fluorinfo s e productio e givea numbear n n i n f previouo r s publications [Refs 1,2]. Fluorin s producei e e electrolysith y b d f o s e fuseth AH n di F salt KF2H t 85-90°a F n milCi d steel ceUs using amorphous carbon anode d watean s r cooled mild steel cathodes. The most recent increment of fluorine capacity was provided by the introduction of 12 KVa cells to augment the 5 KVa cells originally installed.

Because the economics of UF, production are largely determine e costth f fluorino y s b d e generation considerable development effor s beeha tn focusse n thio d s area. Short term development work has been aimed at improving the reliability of cell components and has led to the introduction of fine grain structure porous carbon anodes with improved compatibility with molten fluoride systems. Longer term development wor s concernei k d with optimisatio f electrodo n e properties and cell geometries to minimise inefficient current usage. The prime target is the reduction of anode over voltage which is uniquely high for commercial electrolytic systems.

2.4.2 UF, Production. D UF is fed at a controlled rate into an inert fluidised f calciuo d be m fluoride which serve o dispenst s e rapidly the highly exothermic reaction heat and so prevent the possibilit e formatioe sinterinth th d f o yan f nfeego F U d of undesirable intermediate reaction products. The bed s maintainei t ^450°a d a mone n Ci l reactor fluidised with a mixtur f nitrogeo e d fluorinan n d operatinan e a negativ t a g e pressure e fluorinTh . e reacts instantaneouslF yU wite th h o fort m gaseous uranium hexafluoride, whic s firsi h t filtered to remove entrained solids and then condensed. The incondensible gases containing nitrogen and unused fluorine are recycled, usin a reciprocating g compressor e reacto e th bas th f o eo t r, where fresh fluorin s introducei e a rat t ea d equivaleno t t its usage. Control of reaction temperature within the desired range is achieved by a combination of electric muff heating r forceo r coolingai d .

195 Hexafluorid s removei e d froe condensemth y liquefactiob r n f intd gravite appropriatof an th o n ru y e transit cylinder from where a gassing back operation is performed to eliminate light gas contamination. A bank of four condensers is provided o operatt s coolina e r liquefyino g g unit turn i s n usina g fluorocarbon liqui t -40°a d C whe n primaro n r baco y p u k cooling dutie d fluorocarboan s n vapou t 90°a rr liquefactioCfo n of UF,. o

In recent years attention at the UF, production stage has been centred on issues concerning containment of HF . releasesUF r o . Modifications have. beeUF ne madth t a e o b filling station o tha s l sfillin al td samplinan g g operations cae carrieb n t remotelou d d operatoran y e locatear s n i d a specially constructed control corridor physically isolated froe fillinth m g station otheo N .r operation e permittear s d e fillinith n g station area whilst cylinder fillinn progressi s i g .

In the event of a release of UF, within the confines b of the filling station a hex release shutdown and alarm procedure would be activated either by an operator or automatically , leaUF k a detector y b e automatiTh . c shutdow d alaran n m procedure shuts off the filling station ventilation system, close e cylindeth s e condenser th valv d an e f r of liqui n ru d valv d sound an ePlanx He s t alarme Controth t a s l Paned an l e Workth s Emergency Control Centre whic s quickli h y able o mustet r trained resource o deat s l , leakswitUF h . b

As a final measure, both the UF, Filling Station and HF Stock Room wile coupleb l a spra o t d y towere b whic n ca h brought into use in an emergency.

2.5 Reduction of Uranium Tetrafluoride to Uranium Metal e continuinTh g requiremen r uraniufo t m metae th l r fuefo l UK Magnox reactor programm s meanha e t that uranium metal s remaineha n importana d t nuclear fuel intermediatt a e Springfields tha w since sitete star th No therth ef . o t e a reviva s i f intereso l n uraniui t m metal productioa s a n consequence of the atomic vapour laser isotope separation

196 programmes being pursued in the USA, France and elsewhere s appropriati t o includt e e some comment e developmenth n o s t and current status of the uranium metal production stage at Springfields.

2.5.1 Development of the UF. Reduction Process. In common with mcst other countries at the time, the UK first produced massive uranium metal billets by reduction of loose charge d calciu f an mixtureo s F mU chipf o s r o s turnings

UF4 + 2Ca -> 2CaF2 + U + 134 Kcal.

For economic reasons magnesiu s themwa n consideres a d a reducing agent despite the unfavourable heat of reaction

2 Kcal8 + > MgF - .U g + 2 2M + 4 UF

The loose charge method, suitable calciuth r fo me reduction process s unsuitablwa , r magnesiufo e m reduction since th e temperature immediately following reduction would be less thae slath ng melting poin ^ 1450°C ( e operatint th t a ) g scale proposed and slag metal separation would not be possible, It was therefore necessary to select a process and equipment which would allow for increasing the heat content of the charge before onse f reactiono t programmA . f developmeno e t work in the early 1950s which addressed features such as the following:

i. charge density, which should be as high as possible o ensurt e maximum thermal conductivity, facilitating maximum preheatin d minimuan g m heat loss durin e reactioth g n

. ii charge geometry, which shoul e designeb d o t d avoid premature local ignition before adequate preheating e wholoth f e charge

iii. excess magnesium requirements

iv. choice of reactor containment

197 e introductioth o t e firsd th le f t o nplan t scale production process in 1954. This process forms the basis of the magnesium reduction plant in use today for production of both natural d tailan s depleted uranium billets dato T e. approximately s beeha 40,00 nU producet 0 t Springfieldsa d .

2.5.2 4 ReductioDescriptioUF e th nf o nProcess . Pelleting of the UF./Mg Mixture Pneumatically conveyed or drummed UF. and drummed magnesium raspings are charged to separate hoppers on the pelleting press. From these hoppers the reactants are fed directl o theit y r respective weighing machines where approximately 3 kg UF. and 0.5 kg magnesium are automatically and precisely weighed out. In accordance with a predetermined programme, the contents of each weighing machin e dispensear e x rotatinsi f do intge on omixin g buckets located around the periphery of a horizontal wheel which s keyei d int a overtica l shafte continuouslTh . y rotating bucket a time indexe t ar sa e ° unti e completel60 d th l y mixed contents of each bucket are discharged into the die n hydraulito 0 hoppe20 empta e c f th preso r yd bucketan s s reture fillinth o t gn statio pellew o begit n ne ta n mixing cycle.

From the die hopper the UF./magnesium mixture is e hydraulith feo t d c press which produces cylindrical pellets

approximately 5" diameter x 3^/4" high (125mm x 85 mm) at a pressure of about 10 tons/cm2 and at a rate of 4-5 f eaco d h pelleten pelle r e minutepe th st t cyclA e pres. th e s die is sprayed with lubricant to ensure easy removal of the subsequent pellet.

Reduction of UF^Mg Pellets Pellets containing pure magnesium swarf are blended with pellets containing Magnox (a magnesium alloy) swarf arisings from the Component Manufacturing Plant, in an appropriate rati o givot e desire th e d reduction batch size, generally about 350 kg U.

198 The pellet e thear s n loaded inte internath o l reactor assembly which consists essentially of a graphite lined stainless steel catchpot supportin a gspigotte d graphite liner. Pellets of UF./Mg are layered around the inside wall of the liner and the charge built up by the addition of liners and pellets until the complete charge has been assembled. The assembl s thei y n sealed int oa stainles s steel reactor and pressure tested.

After pressure testing the reactor is charged to a preheated electric furnace and the lid is connected to the pressure relied servican f e linesr approximatelFo . ye pre th 75 f %-o ignition heat soaking period the reactor is evacuated to remove moisture which would otherwise interfere with the effectiveness of the reduction. For the remainder of the heat soak period e reactoth s paddei r d wit a pressurh o 7 t argo 0. s f ga no e r abovba 0 e 1. atmospheri o t c [Refs 3,4]. After approximately 100 minutes heat soaking the charge fires with a noticeable pressure increase. The furnace is then switched off and the reactor assembly allowed to cool for one hour before disconnection and removal to the cooling bay. Twenty four hours later the reactor is broken down to yield a billet which proceeds to the uranium casting plant, a magnesium fluoride slag whic s crushei h d recyclean d r recoverfo d f uraniuo y m values and reactor components such as liners and catch pots whic e recyclear h d withi e reductioth n n plant.

The combination of high quality UF. feed «1% of both

UO2 and UO2F2) and pelleted reactants yields a high quality billet product wit a hmeta l recovery efficienc f 97o yr %o greater.

While the main activities in the reduction plant are concerned with uranium billet productio e Magnoth r fo nx nuclear reactor programme recently introduced variations on the basic reduction process theme include:

. i uranium alloy productio y co-reductiob n n with the appropriate alloying compound

199 . ii adjustmen e changth o t n feeti e d material from

fluidised bed UF4 to rotary kiln U F .

2.6 Alternative . RouteUF o t s i. Although all the major Yellow Cake converters are still using traditional dry route conversion technology at e fueth lf o cycle frone d expensivth th een t e chemicals

(HNO3.AHF) and the high temperature energy intensive steps involve e sufficienar d t incentiv o revivt e e interest in more direct wet routes to UF. incorporating electrolytic reduction.

Some recent development work at Springfields has concentrated attentio e electrolytith n o n c reductio f uraniuo n m (vi) to uranium (iv) in both sulphate and mixed fluoride/sulphate systems. Highly efficient reductio e achieveb n ca nn boti d h systems although the flexibility of the sulphate system is e limitewelth l y b established solubilitiew lo d f uraniuo s m (iv) sulphate. The definition of the conditions required to achieve high solubilit f uraniuo y m e fluorid(ivth n )i e sulphate system offers prospect r efficienfo s d economian t t rout. we c UF e processes. This developmen s potentiaha t r applicationfo l s t onl no n conventionai y l Yellow Cake conversio t alsbu no e conversiofoth r f uraniuo n m arising froe reprocessinmth g of irradiated oxide fuels.

ii. A renewal of interest in processes for reduction of o satisftailt . e sincreasin UF th ydeplete o t , UF gd demand for depleted uranium for laser enrichment ana for non-nuclear uses s prompteha , a dreviva f developmeno l e t th wor n o k fluidised bed route from UF, to UF . The early results from this work look very promising.

3. RADIOACTIVE DISCHARGES AND ENVIRONMENTAL MONITORING 1 Liqui3. d Radioactive Effluent e radioactivitVirtuallth f o l e liquial th y n di y effluent discharged from the Springfields site stems from purification stage raffinate whic s neutralisei h d with lime slurry before

200 discharge inte tidath o l estuar e Riveth f ro y Ribble some two miles from the site [Ref 5]. Control of this discharge is subject to the provisions of the Radioactive Substances Acn accordanci t s i (1960 d an ) e wit a hCertificat f Authorisatioo e n granted e jointlDepartmenth y b y f Environmeno t e th d an t Ministry of Agriculture, Fisheries and Food. The Authorisation imposes limitation d conditionan s s relatin e methodth o t g s of disposal and quantities permitted to be discharged, the sample e take b d analysed e o recordan t ns th e b d o an t s, kept n ordeI . o demonstratt r e compliance wite th h Authorisation the combined site effluent is continuously sampled and analysed for total alpha and total beta.

3.1.1 Liquid Radioactive Effluent Discharges. In the most recent 12 months period reported [Ref 5] e bet th e alph ath d q (21.an a ) TB discharg 6Ci 8 0. s wa e discharge was 152 TBq (4,100 Ci). Both the alpha and beta activity discharges are from the non-uranium members of the decay chain and represented 6% and 34% of the respective authorised limits.

3.1.2 Environmental Impact. In outline, BNFL's policy on radioactive effluent management is to operate well within statutory requirements and to keep any radiation exposure as low as reasonably achievable (ALARA). An important consequence of this policy is that discharges e limitear o thae s committed th t d dose equivalena o t t representative member of the critical group should be no greater than 0.5 mSv, related to each year of operation. e SpringfieldIth n s case criticath e l grou e peoplar p e living in house boats mooree Ribbith n i de e estuarymosth r t Fo . recent 12 month period for which a report is available analysis of silt samples showed that the annual dose to the critical group attributabl o Springfieldt e s discharge s 0.00wa s 3 mSv, predominantly from Pa.

3.1.3 Raffinate Treatment Proposals. The interest shown in recent years by potential licensees e BNFoth f L conversion proces s prompteha s d development

201 oa raffinatf e treatmen n situationi t procese us r sfo s where e prevailinth g regulation e planth r t o slocatio n would call for removal of activity before discharge of raffinate.

Evaluation of the candidate processes, initially from a theoretical stand point and then by a programme of laboratory scale experiment e conclusioth o t d le sn tha a viablt e process for purification plant raffinate decontamination would comprise:

. i selective solvent extractio f thoriuo n m isotopes and immobilisation of the radioactive extract as an active solid residue for eventual disposal.

and ii. treatment of the radioactively decontaminated raifinate o immobilist e heavth e y metal a non-radioactiv s a s e solid residue for disposal a L a municipal site.

For a notional 10,000 t U per year UF, conversion plant feeC usindUO modea g l which reflect e wholth s e rangf o e impurities likely to be encountered in the future it can be predicted tha a raffinatt e treatmen te lineplanth f o sn o t that outlined, would generat r a yearadioactive pe f abou o t r 5 5 t e sludge.

A pilot plant for progressing the next stage of the raffinate treatment development programme wile broughb l t into operation in 1987.

3.2 Airborne Radioactive Effluent Elimination of particulate uranium from the gaseous discharge of UOC conversion plants is effected by a combination of primary filter r cycloneo s s backey absolutb p u d e filters. All such discharge a e stackappropriatvi th f e o sar s e height which are sampled on a continuous or regular programmed basi d whic an se regulate ar h n accordanci d e wit a hCertificat e of Authorisation. The alpha activity discharges consists almost entirel f uraniue o accompaniey ar d man n effectivela y b d y equal beta activity from uranium daughters.

202 3.2.1 Airborne Discharges. e airbornTh e discharge froe e Springfieldwholth mth f o e s site for the most recently reported 12 month period was 0,001 TBq (0.03 Ci).

^ • ' Solid Radioactive Effluent Solid radioactive effluent from Springfields consists of inactive or lightly contaminated and mainly incombustible wastes such as incinerator ash, building rubble, glassware scrap metal and ore processing residues. This effluent is disposed of in accordance with a Certificate of Authorisation granted jointly by the Department of Environment and the Ministr f Agriculturo y e Fisherie d Foodan s .

e mosIth n t recently reporte mont2 1 d h perio e dischargth d e amounte n 11,65i o 0.0 t d f soli q o 4t 0TB d effluent.

SUMMARY Springfields Works, which was the first factory to operate in the UK atomic energy programme will, in a few months time, be celebrating the 40th anniversary of its start-up.

Since that tim n additioi e o meetint n g approximately 15% of current world demand for UF, conversion; b

. i Springfield s produceha s ^ millio3 d n Magnox fuel millio» » elementfueI fueI R R 1 1 ll AG AG n ;ntpinspinsd d an an sn energ ;a ; y equivalent roughly equal to 600 million tons of coal

. ii Springfield technologR ID s y either directlr o y under licenc s produceha e a dsignifican t proportioe th f o n ceramic grade uranium dioxide currently used in nuclear fuel fabrication throughou e worldth t .

BNFL believes thae experiencth t e gainee pasth tn i d coupled with the results of the ongoing in house research and development provides a good technology base for dealing wite problemth h d opportunitiean s s which e facehavb o dt e in the future.

203 REFERENCES

[1] ROGAN, H. "Production scale processes and plants in the United Kingdom - The conversion of uranium ore concentrates to nuclear grade uranium hexafluori.de and to enriched uranium dioxide". IAEA Study Group Meetinn o g the Facilities and Technology needed for Nuclear Fuel Manufacture. August 1972.

[2] PAGE, H. "United Kingdom experience of production of uranium fluorides". Proceedings of an IAEA Advisory Group Meeting, Paris, June 1979.

] [3 WILLIAMS, A.E PatenK ."U t Specificatio 780,974o N n , 5 March 1954.

(4 ]Paten K U DUXBURY, t al Specificatio t e , D. , 933,436o N n , 22 January 1959.

[5J BNFL. Annual Repor n Radioactivo t e Discharged an s Monitorin Environmene th f o g t 1984.

204 PARTn REFININ IRRADIATEF GO D URANIUM MATERIALS CONVERSION OF URANIUM ORE CONCENTRATES AND REPROCESSED URANIUM TO NUCLEAR FUEL INTERMEDIATES AT BNFL SPRINGFIELDS Par : ReprocesseB t d uranium

. PAGH E Springfields Works, British Nuclear Fuels pic, Preston, Lancashire, United Kingdom

Abstract

Som O eU hav 25,00s a e U bee 0t n recovered from irradiated •J uranium metal fuel (Magnox) since reprocessing,e begath t a n Sellafield Works of BNFL more than 30 years ago. More than 15,00 U equivalen 0t f thio t s reactor depleted uranium have been converte o uraniut d m hexafluorid t Springfielda e s Works.

This paper summarises past experienc f convertino e g reprocessed uranium to UF, and the changes which will be encountered in the o future.

1. INTRODUCTION

hav_ SomUO ee bees 25,00a nU t recovere0 d from irradiated uranium metal fuel (Magnox) since reprocessing began at the Sellafield Works of BNFL more than 30 years ago. More than 15,000 t U equivalent of this reactor depleted uranium, at U levels of less than 0.711% have been converted to uranium hexafluoride at Springfields Works d subsequentlan y enriche e BNFth t La d diffusion plants r Urenco o Centrifuge plan t Capenhursta t . This source of enriche fueR d luraniu AG mad e th em f o account% 75 r fo s to date and its reactor performance has been equal to that of fuel derived from non-irradiated uranium.

e earlIth n y 1990s uranium originating froe mth reprocessing of irradiated oxide fuel (AGR and LWR) will emerge from the THORP reprocessing plant at Sellafield.

207 In order to return this uranium to the nuclear fuel cycle plants will be provided at Springfields Works which will:

. a conver e uraniu, suitablth t UF r feedino mt fo e g o diffusiot r centrifugo n e plants

and b. convert the product of the enrichment plants into finished AGR or PWR fuel assemblies.

This paper summarises past experience of converting reprocesse e changeth d d an suraniu , whicUF o mht wile b l encountered in the future.

. 2 CONVERSIO F URANIUO N M RECOVERED FROE MTH REPROCESSIN F IRRADIATEGO D MAGNOX FUEL

2.1 Process Description and Operating Experience The process plant used for converting, to UF,, pure uranyl nitrate generated by the reprocessing of irradiated Magnox fuel has traditionally been similar to that used for converting pure uranyl nitrate derived from uraniue mor concentrates. This process plan s beeha t n fully described in previous paper d consistean s ] [Ref 2 1, ds essentially e followinoth f g principal stages:

. i continuous evaporatio f puro n e uranyl nitrate in multi-effect climbing film evaporators

a. continuous denitration of molten uranyl nitrate hexahydrate in a single stage fluidised bed to produce uranium trioxide

iii. batch wise reductio d hydrofluorinatioan n n singli n e *> stage fluidised bed o product s e uranium tetrafluoride

. iv continuous fluorinatio f uraniuo n m tetrafluoride in a single stage fluidised bed to produce uranium hexafluoride.

208 In the early years of conversion of reprocessed Magnox depleted uranium (MDU e Springfield e th fee )th o t d s conversion sequence consisted of MDU pure uranyl nitrate at approximately 250 g U /l and with " U concentrations of 0.4 to 0.6%. This liquor was transported from Sellafield to Springfield n stainlesi s s steel road tanker d thean sn > U- F processinN U e th f o g e Fluidisee lineth on n i feso t d d Bed Plant. However, shortle yth aftef e o star th p r u t 2nd Uranium Plant at Springfields, evaporation and denitration units, identica o thost l e develope t Springfielda d s were installed at SeUafield. Since 1964 all MDU U0_ has been produce e reprocessinth t a d g site and, after appropriate under cover storage in 50 gallon mild steel drums, is transporte y roab do Springfield t d r conversiofo s o UF,t n .

Following the successful introduction of the first Rotary Kiln UF Plant at Springfields a second such plant is now under construction to replace the fluidised bed reduction and hydrofluorination stages which have recently been phased out for reasons of economy and safety. Future MDU processing campaigns will therefore incorporate rotary kiln technology instead of fluidised bed technology in the long established , conversioUF UO o t , n sequence.

Whilst the majority of the MDU UCL produced at SeUafield has been enriched at the Capenhurst diffusion plants for subsequent processing into AGR fuel, smaller quantities at lower enrichments were diverted to stock prior to the recent closure of the diffusion plants. This stock is currently e URENCth o t bein d O fe gcentrifug e plant s wila s l arisings of UF, from future MDU conversion campaigns.

Although most of the BNFL experience of conversion of MDU relates to eventual use of the uranium in AGR fuel a small quantity of pseudo-natural uranium has been produced from blends of reprocessed MDU and urai.ium derived from the reprocessing of irradiated oxide fuel. Blending of the two components as uranyl nitrate was carried out at Springfield e pseudo-naturath d an s l liquo s thewa nr converted

209 o uraniut m e conventionabilletth a vi so metat N U ll sequence described previously unforeseeo N . n problems were encountered in Magnox fuel manufacture nor in the subsequent second irradiation e pseudo-naturacyclth f o e l uranium metal.

^ •^ Health Physic d Safetan s y Considerations 2.2.1 Origins of the Radiological Hazards. In the Springfields process sequence from UO, to e UFmaith , n radiological hazard arises froe possibilitth m y of uranium dust inhalation, a hazard which can be exacerbated e presencth y b f traco e e quantitie f transuranio s c elements when MD Us beini g converted. These elements, although presen n insignificana n i t t amoun y massb t , coul y virtub d e of their high specific activity, constitute a significant proportion e Deriveoth f r ConcentratioAi d n (DAC e activitth f o )n air i y .

To ensure that the proportion of the DAC taken up by the transuranic elements is kept within levels which do not constitute a major problem, it is the practice to define a limiting transuranic specification which is related to the uranium.

e presencTh f residuao e l fission product e MDth Un i s can also lea o unacceptablt d e external radiation levels unless controlled. Experience has shown that radiation levels millire5 1 e o surfacr t houth mpe p t o u a f rfdrum o e r o s reaction vessels can be tolerated without significant exposure of personnel. This leve s beeha l n prescribe e limitinth s a d g radiation level acceptable in the Springfields conversion sequence -

The main radionuclides of interest to the plants in the UO, to UF, sequence when converting MDU are:

Transuranic elements Plutonium. Neptunium Residual, fission products Ruthenium

The concentration of Th, the decay product of deriveU 23MD 2 dn Ui fro, e reprocessinmth e relativelth f o g y

210 low burn-up Magnox fue s beloi l e levewth t whica l h Health Physics problems became significant.

2.2.2 Health Physics arid Safety Experience.

1 StageUF UO o st 3 No significant air contamination or external exposure dose rate problems were encountered at the fluidised bed . stageUF UO o st , wher l powdeal e r transpor d reactinan t g systems were completely enclosed. For the same reasons no difficultie e anticipatear s e rotarth t o a dt y ~ kilUO n UF. stages during future MDU conversion campaigns.

UF^ to UF& Stage Virtuall e tracth l e al yplutoniu d rutheniuman m present in the original feed UO, is retained on the calcium fluoride particulate solids which are from time to time removed f T om the fluidised bed system in the form of reactor residues or reactor filter residues. An important consequence of , reactiothiUF so t n featur. a veract UF s s a ysthai e th t effective decontamination stag r thesfo e e nuclides.

During MDU conversion campaigns solid residuee ar s removed from the fluidised bed reactor system in accordance with well developed techniques, operated under Healtd an h Safety Department supervision, involving temporary containment, pressure suit respiratory and contamination protection, and contro f radiatioo l n exposure e residueTh . e sealear s d in small mild steel drums and, after appropriate storage 238 to allow for the equilibriation of the short lived U daughters which are also removed at the UF. to UF, stage, the residues are transported to a Company disposal site.

General Comments The concentrations of transuranic alpha emitters in uranium recovered from the reprocessing operation are sufficiently low as to have only limited radiological impact on the conversion stages at Springfields.

211 Much the same comment applies to the principal fission 99 product contaminants, Te a long half-life soft beta emitter and Ru a short half-life gamma emitter (via its daughters).

In the case of MDU feed material interim storage of the UO, product of the reprocessing plant and appropriate campaign scheduling effectively eliminated the radiological impact of Ru on the UO, to UF, conversion stages.

2.3 Product Quality More than 15,00 f MDo t 0s beeU ha n processed through e conversioth n stage t Springfieldsa s f chieO .f concern was the residual fission product concentration and the transuranic alpha activity.

A number of campaigns have been specially carried out at Springfields converting UO, with very high levels of transuranic alpha activity, essentially due to neptunium. It was found that most of the neptunium content could be remove d preventean d d from enterin n enrichmena g t plant. The primary mode of neptunium removal was again shown to be adsorption on calcium fluoride in the fluidised bed reactor system. Residual neptunium was removed in special traps and further decontamination took place during the vaporisatio , transinUF froe t mth cylinder s when feeding the enrichment plant. These experiments have demonstrated e specificatioth n requirement r uraniufo s m productf o s reprocessing plants for conversion to UF, for supply to enrichment plants.

3. CONVERSION OF URANIUM RECOVERED FROM THE REPROCESSING OF IRRADIATED OXIDE FUEL (AGR.LWR)

3.1 Health Physics and Safety Considerations In addition to the naturally occurring isotopes of uranium, , otheU r uraniud an U m Uisotope, e producear s d in nuclear fuel durin d aftean g r irradiation. These ar e principally 232U, 233d 236 f Uwhican Uo e greateshth 232s Uha t potentia r e impacHealtfo th l n o ht Physics aspect e conversionth f o s .

212 As previously mentione e resultanth d levelw lo tf o s e relativelth n i U burn-uw lo y p Magnox fuel meant that the Health Physics implications of MDU conversion were associated principally with residual transuranic alpha emitters such as plutonium and neptunium and residual fission products such as ruthenium.

However BNFL are currently planning to construct at Springfields a new dedicated plant for the conversion of the oxide reprocessed uranium (ORP) which will be produced e Thermath n i , l OxidUO s a e Reprocessing Plant (THORP) at Sellafield. It is intended that the Springfields conversion plant will be available coincident with the commissioning of THORP in the early 1990s. It is envisaged that shortly afterwards a comprehensive fuel manufacturing facility will be provided at Springfields to produce AGR and PWR fuel from ORP UF, enriched in URENCO centrifuge plants.

3.1.1 Characteristic UraniumP OR f o s . e implicationSomth f o e epropose th r fo s d conversion plan r handlinfo t uraniuP OR g m froe reprocessinmth f o g oxidR eLW fuee discussehighed ar lan R r AG burd p u n belo y comparinwb e principath g l differences betweeP OR n uranium and non-irradiated or natural uranium (NIU).

232 .. ___Uranium This isotope which does not exist in natural uranium can be formed by a number of routes (Fig 1) and it can e seeb e principan th tha l al t l uranium isotope e potentiaar s l sources of U following irradiation. The U concentration in ORP uranium depends on several factors including:

e elapseTh d. i time x betweevaporisatiohe e th n n stage of fuel manufactur e starth f irradiatioo d t an e n

e fueTh l bur . ii p u n

e coolinTh iii. g time before reprocessing „, <. .. f 232,, 234., 235 , 236 iv. The concentrations of U, U, UT1 and UT7 before irradiation.

2!3 .. / ^ ————— —— —— -..,....—. L U

n , 2n n, Y

230 7 77 V A7^ b Th " u n,T s 237,, n,r

232 231 236 237 Th Th Np Np n, 2n n, 2n

231 236 Pa Pu n,r

FIGUR l E Production Route r 232fo sU

Typical concentrations of U in ORP uranium are 232 a few parts per billion on a total uranium basis. U has a 72 year half-life and is the parent of a complex decay 228 chai e immediatn Th (Fi . ga half-lif2) es daughteha h e T r 9 yearo1. f s whic s muci h h greater tha e halth n f livef o s subsequent daughters, these being of the order of minutes 090 h therefor r T dayso . e control e concentrationth s f o s all daughters as a function of time from the point when 228 Th is separated from its parent by chemical or vaporisation 228 h witT processesh tim f e o buile Th p frou d. m separation from the parent U is illustrated by the table:

Time since separation % of Equilibrium

1 month 3 1 year 33 4 years 83 10 years 100

214 71.1 yr

228Th 1.91 yr I 224 RD a 3.66 day I 220., Rn

55.c se 6

216Po 0.15 se

212Pb

10.64 hr 60.55 min 0.305 us

361

3.07 min {!>, V Stable

FIGURE 2 Decay of 232U

T T Q At equilibrium the activity of Th is equal to the 232 e activitieth e ar e subsequenf eacth s o s a f o hU activit f to y daughter s concentratio it a nobl s s i sga e n althougR n s a h and that of subsequent daughters is affected by the containment e materias chemicaoth it f d an l l e formannuaTh . l limit 232 s therefori o f U intak f eo e more restrictive thar fo n U, U and U due to the presence of the short lived daughters whic e bonar h e seekers with long body retention times.

232 208 daughtersU l emitT e th On sf , o ever y penetrating 2.6 MeV gamma radiation which represents an additional, independent radiological hazard r exampleFo . e ingrowtth , h of this daughter reaches a maximum after a decade of storage,

215 t whica h poin e gammth t a dose raturaniuP eOR froe mmth ordero tw f magnitude o sb n ca e greater than tha f nono t - irradiated uranium. Since .-.eparation froe gammmth a emitting daughter n occu a numbeca s t a r e sequencf stepo th r n i s e UO, to finished fuel the increase in gamma dose rate is influence e timth e y intervab d l between these steps.

234 , Turanium This isotope constitutes 0.005 f non-irradiate%o d uranium an s concentratei d d during each enrichment process. Typical concentrations in ORP are 0.02%, increasing to 0.13% after further enrichment. The main cause for concern is the 3 fol2- d ris n specifii e c alpha activity whicf particulao s i h r significance becaus e potentiath f o e l consequences with respect to airborne activity.

235 TUraniu1 m The mean discharge U concentration from oxide reactor s expectei s e aboub o t e drangt th 0.9 d e %an abou t this mean is expected to be 0.6 - 1.2.

236., Uranium This isotope does not occur naturally and is produced from U by neutron capture. It functions mainly as a neutron poison, requiring extra enrichmen d thuan ts imposing an economic penalty on the ORP fuel cycle.

Transuranic Alpha Emitters Thes e producear e d during irradiatio d arisan n e from neutron capture and decay chains. They comprise essentially the residual traces of neptunium and plutonium after the reprocessing operatio d botan nh isotope e presenar s n i t the ORP uranium at sufficiently low concentrations to have limited impact.

Fission Products Thes e principallar e e traceyth f rutheniuo s d technetiuman m remaining in the ORP uranium after reprocessing. Ru is a short half life (1 Year) isotope which increases the

216 99 gamma activit e uraniuth f o y m (vi s adaughter)it c T . a lon s i g half life soft beta emitter.

3.1.2 Health Physics ConversioAspectP OR f o s n Fiant Design. Detailed study of the Health Physics implications of introducing ORP uranium into the Springfields conversion process led to the conclusion that the consequences of the increased U concentrations would be:

i. a more restrictive airborne contamination limit due to the alpha emitting daughters n increasea . ii V d d Me gamm an 6 2. ae dosth o et rate du e gamma emitting Tl daughter.

To ensure thae externath t l plus internal radiation dose uptake satisfied statutory requirements and was consistent with BNFL policy guidelines it would be necessary to make extensive use of shielding and remote handling equipment in the conversion process. Backfitting of existing equipment wa d followins an rule t ou da fulg l scale plants trialwa t i , decided to provide a dedicated plant for the conversion of ORP UO-j to UF,. 3 D

The choice of the fluidised bed route for the , meanproductioUF P s thaOR e majorit f th to n f o y e non-uranith c impurities wile removeb l y adsorptiob d n on the calcium fluoride bed material. All of the residual plutoniu d uraniuman m decay producte d th mos f an so t rutheniu d neptuniuman m wile immobiliseb l e fluidiseth n i d d d residuebe s which will require special handling arrangements for their removal, storage and eventual disposal. Similar arrangements will be provided to deal with the residues generated in the transit cylinder washing facility and those arising from neptunium and plutonium traps.

217 3 . 2 ORP Uranium Conversion. BNFL's Intent ions 3.2.1 Project Assessment. o principatw BNF d ha L l object n mini s d when evaluating the recycling of ORP uranium to finished AGR or PWR assemblies:

i. to ensure that the radiation dose uptake was consistent with BNFL policy guidelines and within statutory limits o ensurt . eii tha e economid th tan c cas r recyclfo e f o e ORP uranium was sufficiently attractive to persuade utilities o adopt t recyclin s sooa g s possiblea n .

Consideratio e Healtth f o nh Physics implicationf o s recycling indicated that y comparisob , n with non-irradiated uranium extra costs would be involved because of the additional plant complexity associated with the extra shielding, containment d handlinan g constraints. Nevertheless, despite substantial processing cost premiums compared with non-irradiated uranium, there is a strong economic incentive to recycle.

Early in the project evaluation exercise consideration was given to the possibility of blending ORP uranium with non-irradiated uraniu t msoma e e overalstagth n i e l sequence, pure uranyl nitrate to finished fuel. However as this would require all the downstream plant to accommodate ORP derived feedstoc e concepth k f blendino t s rejectewa g n favoui d r of the segregation concept. Segregation permits ORP uranium to be confined to the minimum volume of processing equipment, s particularli d an y well e suiteURENCth o t d O centrifuge enrichment technology which can accommodate modules designed specifically to handle reprocessed uranium.

It is now BNFL's intention to construct a new segregated conversion facility at Springfields which will come on stream coincident wit e commissioninth h f THORo g t a PSellafiel n i d e earlth y 1990ss envisagei t I . d that shortly afterward a fues l manufacturing facility designed to process ORP uranium from enriched UF, through to finished AGR and PWR fuel will be commissioned.

218 3.2.2 ORP Uranium Conversion Technology. The reference design of the conversion facility at Springfield e followinth s ha s g characteristics:

i. The 650 t U per year unit will be broadly compatible wite averagth h e outpu f THORo t d wile broughb an Pl n lino t e e samth t ea time.

e conversioTh . ii n unit will accept feed UO., containing 235 a maximum U content of 1.3%. This specification will embrace mor f eTHORo tha% 95 nP s arriveproducwa d dan t at by balancing the cost of separative work lost by blending agains e increaseth t d uni t e geometricallcostth f o s y more 235 restricted plants able to accommodate higher U assays.

e planTh iiit. wile designeb l o limit de averag th t e annual effective dose equivalen e summatioth e (i t f externao n l and internal radiation doses) received by plant personnel to not more than 0.5 rem (5 mSv) per year.

Uranyl Nitrate Evaporation and Denitration The evaporatio d denitratioan n e purth f e o nurany l nitrate produce e uraniuth n i d m purification sectio f THORo n P wile carriet b Sellafiela l t ou d n plani d t incorporated into e THORth P complex.

235 After any necessary U assay adjustment the uranyl nitrate liquo d intr fe oa geometricall wil e b l y safe single stage continuous evaporator. The concentrated uranyl nitrate evaporator product, at ~ 1000g U/I, will then be continuously denitrate t 300°a d a geometricall n Ci y safe electrically heated fluidise d reactor be de UO Th _. produc t e denitratiooth f n stage a fre, e flowing powder, wile b l precooled during transport from the denitrator to storage hoppers and will then be drummed off for transport to Springfields.

, UF o t O ConversioU P OR f o n conversioP OR e Th n plant will diffe, rUF froe mth o plant described briefly in the previous paper in that:

219 i. all process equipment such as hoppers, reactors, condensers, cold traps, scrubbers etc will be of geometrically safe design

. ii there wile extensivb l f shieldingo e us e , containment d remotan e handling equipment appropriat e increaseth o t e d radiological hazards previously discussed particulaA . r example of this increased hazard when processing high uraniuP OR Dur e p problemu nth wile b f lhandlinm o g calcium fluoride reactor residues in which will be concentrated the residual transuranic alpha emitters and fission products together wite decath h y daughter e uraniuth f o s m isotopes notably U

appropriateln A iii. y shielded dedicated facility will be provided for the periodic washing of UF, cylinders in the ORP transport cycle. The washing facility will also contain equipment for immobilising the gamma emitting solid residues, mainly U daughters, removed from the cylinders during each washing cycle.

3.3 Uranium Recycle and Effluent Disposal 3.3.1 Enriched Uranium Residues Recovery Plant, EURRP, Enriched uranium containing residues of many types arise froe diversth m e rang f processino e g operations carried out Springfieldsa t . These residues have hitherto been recovered in a mass controlled, batch operated plant. A new plant has recently been installed and commissioned base n saf o dy shap b e e equipmentplanw ne t e whicTh . h will also be available for treatment of enriched ORP recycles is to be used to process residues in the 1 to 5% enrichment s improveha rang d an en lin i d e storage facilitie o meet s t current standards of uranium accounting. Where possible the recovery processes are continuous. Another key feature recoverw ne e oth yf plan s bee s thaha i t n t i t designed o givt r econtaminatio ai verw lo y n l leveloperatoal n i s r areas d this beean ha s, n achieve y puttinb d g virtually all the equipment into cubicles.

220 The processes use e essentiallar d y simila o thost r e e existinuseth n i d g plant, standar t processewe d s being used for dirty residues and dry processes for clean residues. The principal end product is pure uranium dioxide.

The essentia l e recoverstepth n i s y proces e prear s - treatmen f impuro t e residues followe y dissolutiob d n i n nitric acid. The impure uranyl nitrate is purified by solvent extraction using pulsed sieve plate columnsr extractiofo e on , n and strip, and a second for back-wash.

Ammonium diuranate (ADU s precipitatei ) d froe mth pure UN solution and filtered off. The ADU is then calcined

to produce U_00 which can then be reduced to U00. Uncontam- J O 6 inated clean residues can also be integrated into the overall recovery proces y meanb s f oxidatioo s r oxidation/reductioo n n processes involving belt furnaces.

This residue recovery plana majo s i t r additioo t n the facilities at Springfields. The plant overall has dimensions of 60 metres by 53 metres. It is capable of dealing with f 40-7enricheo U t 0 d residue r yeape s r dependine th n o g purit d enrichmenan y e feeth df o tstreams .

3.3.2 Enriched ORP Residues Recycle. Solid residue fueP slOR fabricatiofroe mth n processes, suc s rejeca h t pellets, grinding sludges c wile scheduleet b ,l d for recovery in EURRP at the optimum time with respect o ingrowtt daughterU e subsequen u th f d o h an s t returf o n e recovere th fueP luraniuOR P manufacturinn OR da o mt g campaign.

LiquiP OR 3.3.d 3 Effluent. s intendei t I d thae EURRth t P raffinate stream arising from treatment of ORP residues will be decontaminated on a campaign basia dedicate n i s d e facilitlineth f thoso n s o y e proposed for the UOC purification plant raffinate described in e previouth s paper. After such treatmen e liquith t d effluent wile suitablb l r dischargfo e e locath lo t eestuary . SoliP 3.3.OR d4 Effluent. The ORP solid effluent, at the forecast scale of operations will amount to less than 50 t bulk per year of Intermediate Level Waste. These arisings will be disposed of at one of e controlleth d NIREX sites.

. 4 SUMMARY

BNFL has considerable experience of recycling reprocessed uranium fuelintw ne o. Shortly afte e THORth r P reprocessing plant comes into operatio e earlth n yi n 1990s BNFL intendo t s be able to offer a comprehensive recycle service comprising conversion, enrichmen d fuean tl fabricatio uraniumP OR r o n .

REFERENCES

[1] ROGAN, H. "Production scale processes and plants in the United Kingdo e conversioTh m- f uraniuno e mor concentrate o nucleat s r grade uranium hexafluorido t d an e uranium dioxide". IAEA Study Group Meetine th n o g Facilitie d Technologan s y neede r Nucleafo d r Fuel Manufacture, August 1972.

[2] PAGE . H "Unite, d Kingdom experienc f productioo e f o n uranium fluorides". Proceeding n IAEa f Ao s Advisory Group Meeting, Paris, June 1979.

222 NITROX PROCESS: A PROCESS DEVELOPED BY COMURHEX OF CONTINUOUS DENITRATION

. ROMANR O Malvési Plant, COMURHEX, Narbonne, France

Abstract

COMURHEX(subsidiar f PECHINEYo y developpes ha ) d since 198 2procesa obtaio st n sinterable oxydes by direct denitration of uranyl nitrate. This process can be used to produce mixed sinterable powder of UO2/PuO2. Some experiments were mad proved ean e feasi th d - Nitroe bilitth f xo y process. The paper resumes the advantages of the Nitrox proces gived san s some experimental details.

Recyclin e energth o e plutoniuyt g th lin f o e m originating from spent fuels reprocessing requiree th s development of processes for the preparation of

composite U02/PuO powders of characteristics i

appropriat manufacture th o t e f sintereo e - d pellets. Additionally, and as always where plutonium is being handled, the various powder preparation operations have to be designed to curtail as far as possible the formation of gaseous effluents or contaminated liguids, the decontamination of which is seen as expensive. Again e "sensitiveth , f plutoniuo e on " s materialsi m , the use, possessio d shipmenan n f whico te subjec ar h t to extremely severe regulations intende ensurto d e complete control over utilisation. Based on these considerations, the Comurhex company (a member of the Pechiney Group) has developed the Nitrox process methow ne f manufactura ,o d f composito e e U02/ Pu02 powders by direct thermal denitration of a uranyl and plutonium nitrates solution.

E "LIGHu FUEL P TH - I MIXE / N TI SU DWATER " REACTOR FUEL CYCLE Reprocessin e spenth g t fuels from light water reactors will mak possiblt i e o recyclt e e weakly enriched uranium (ca. 0.9 %U 235 d plutonium)an .

223 The plutonium produced, if recycled in its entirety, could be used to produce some 10 to 15% of fuel requirements. It may be estimated that in 1995 and beyond some 2 000 metric tons per annum (tpa) of "light water" spent fuels coul e reprocesseb d d worldwide. Under these conditions, reprocessing would yield the plutonium potentially sufficien manufacture o th t r 0 fo t20 f o e 300 tpa of composite fuels. The recycling stakes are therefore substantial and have attracte e interese fuee th firmdth th ln i cyclsf o t e industry.

IMANUFACTUR- I COMPOSITF O E E FUEL On leaving the reprocessing stage, the recyclable uraniu e ford plutoniuth f nitrian o mn i e c ar maci d solutions. Two principal types of process have therefore been developed with a view to the production from nitrate solution mixee th df so UO2/Pu0 2 powder suitablr fo e e manufacturth e sintereth f o e d pellets. e firsTh t route correspond e separatth o t s e preparation of the two ÜO2 and PuO2 powders, of characteristics such tha n mixino t g they will yiel mixee th d d UO2/PuO2 powder read r pelletinfo y d sinteringan g . Mixin s veri g y frequentl a two-stagy e procesn a ; s initial "master" powder is prepared at high PuO2 concentration and the U02/PuO2 mix then adjusted by addition of UO2 powder. In the second route, the mixed powder is prepared directly by coprecipitation of the uranium and plutonium salts fromixee th m d nitrates solution. Thes proceso tw e s series exhibit certain shortcoming: s The production of mixed powders by mixing PuO2 and UO2 demands very precise adjustment of the particle sizes of the two powdered oxides. This typ f proceso e s involves grinding operations which generate polluting aerosols. Account also has to be taken, when these aerosols deposi wallsn o t f wha,o t happens to the americium 241, which exhibits very substantial alpha-activity. Processes employing coprecipitation more often than t entaino l aqueous effluents which require b o t e decontaminated.

224 After production of the powder by one or other process, the operation f pelletingo s , sinterin loadind an g f o g rods can be carried out in an environment technologi- cally suited to the handling of plutonium.

NITROE TH Il - lX PROCESS This continuous process, developed between 1982 and 1985 by the Comurhex company, basically consists of preparing the mixed U02/PuO2 powder by direct thermal denitration of a suitably adjusted solution of uranyl and plutonium nitrates. e maiTh n process stage e representear s d belo: w

Uranium nitrate soin. Plutoniu{ ] m nitrate soin. f Pu/U mixin d adjustmenan g t [Concentration by evaporation] ^Crystallisatio y coolinb n g {Dehydration and denitration under vacuum j [. Calcination ————i————. • Reduction | UO2/Pu02 powder ready foruse I

After mixin e nitratth g e solution e choseth n i ns ratio of Pu/( U+ Pu) , thee takear y n concentratioa dow o t n n O gramf somo 20 1 eU s-fu métal/litre P - . The concentrated solution is then crystallised by cooling. The nitrate powder so obtained is dehydrated and then denitrated at reduced pressure. The mixture of oxides obtained can then be calcined and reduced. The characteristics of the U02/PU02 powder obtained depend on temperature and oressure values in the course e dehydrationth f o , denitratio d calcinatioan n n operations and on the duration of the heating steps. Main features of the process are described here after : The main step in the process is that of denitration and dehydration. This operation is conducted using a

225 combinatio f pressuro n d temperaturan e e calculateo t d obviate melting of the nitrates at the denitration stage (cf. Fig. 1) . Operating condition l stime al mus e belot b se a t th w equilibrium curve. n examplA s givea piloi e f o nt operation under which dehydration/denitration operations were carried out in three stages. Main experimental date tabulatear a d below :

A - Dehydration and denitration

Temp. Duration UO3 Ü03 + 6% Pu (Pu02) steps ( °C ) (hrs ) 20 1.5 260 1.5 400 1.5

Specific surface of oxides 32 ml/g 40 ml/g O/M ratio 3.079 2.955 i i î

Calcinatio- B reductiod an n n

UO2 UO PuO- 2 2 Calcination 1 hr - 600° C525°- r h C 1 Reduction 1 hr - 600°C 2 hr - 700°C O/M 2.102 2.073 BET surface 6.2 ml/g 6.5 ml/g Density 1.7 g/cc 2.4 g/cc

C - Sintering

UO2 Ü0 Pu02- 2 Pelletising pressure 3.5 3.5 Crude density 49% 52% Sintered density 96% 94%

226 penmentai graph btaine y ATGb d .

e materiath n i 0 lH2 %

HUH

Fig. l. IV - PRINCIPAL ADVANTAGES OF THE PROCESS e procesTh s presents several substantial advantages, viz : Competitive operating costs Good control of environmental factors Operating safety Eas f operatioo e n Flexibilit d adaptabilitan y e variouth o t ys feedstocks employed and finished products sought. Good final product qualit d reproducibilityan y .

Competitive operating and capital costs e procesTh s flowshee e powdet th dowr sinterin o fo rt n g is straightforward and involves few essential equip - ment. All the plant and equipment employed is normally availabl n industri e d necessitatean y s small special adaptation. The process eliminates or reduces the impact of a number of costly operations existing with competing processes, such as : adjustment of U/Pu concentrations ; particle size reduction ; Pu inventory questions ; the problems of treating liquid effluents and of aerosol contamination.

Additionally, the process comprises a single step of productio f sinterablo n e powder, leading directlo t y the mixed UO2/Pu02 powder ; it therefore eliminates the independent stages of production of U02 and Pu02 powders whic e generallar h y indispensabl n processei e s employin e UQ2/PuOth g 2 powder mix.

Good control of environmental factors e procesTh s generate o aqueoun s s effluents sinct i e necessitates neither filtratio r washingno n . r doe No t generati s y plutoniuman e r uranium-bearino - g aerosols since it entails no grinding or micronising, in contras a numbe o t tf othe o r r processes. e oxideTh f nitrogeo s d watean n r vapour producee th y b d denitration operation are recombined as concentrated nitric acid and can therefore be recycled.

228 Operating safety a) As regards problems of criticality, all the initia procese l th par f o ts (evaporatio d crystalan n - lisation) can be carried out under conditions of favourable geometry. The dehydration/denitration furnace can be controlled y masb r geometryso e remainde th ,e proces th f o rs being carried out in the absence of water. Also worth noting is that the process does not involve transit via a "master" mix of high Pu content and that as from the nitrate solution stage the uranium ana plutoniu e concentration e mixeth b n n i d ca m s required e finisheith n d fuel. u inventorP e Th s eas i ) yo establishb t y , beiny b g control of concentration in nitrate solution. c) The problems of shipment of the Pu feedstock can be very much simplified sinc e Nitroth e x proces n takca s e the uranium and plutonium in separate or premixed nitrate solutions. Mixin n takca g e plac s froe a e th m reprocessing plant, thereby avoidin e problemth g s associate s owndit wit.n e shipmeno th hu P f o t

d) Also deserving of mention is that in this process the plutoniu n remaica m n solutio i ne converte b d an n d to oxide only at the stage of production of the mixed powder n thiI .s way s chemica,it l purificatioo t n remove americium can be carried out just prior to use.

Eas f operatioo e n ) a Adjustmen f U/Po t u concentration e carrieb n ca sd out in a single step at the nitrate solution stage. b) The process can be employed to obtain sinterable powders of substantial Pu content, up to 25% and even beyond. ) Trialc s carrie t havou d e shown thae parameterth t s yieldin e sinterablth g e composite powde e easo ar rt y control. What is involved is maintaining a heating programme by adjustment of temperature and pressure, these being easier to control than chemical parameters (pH etc.) or physical parameters (particle size distributions obtained by grinding operations) which it is essential to set under certain rival processes. d) The powder obtained before pelleting and sintering exhibits good fluidity and, moreover, high density (of the order of 2.5 to 2.8 kg/dm3), thus facilitating the pelleting operation.

229 e) Compacting is carried out at relatively low pressures (of the order of 3 t/cm2). f) Sintering is effected at the usual temperatures. g) The pellets obtained have a good mechanical strength.

Flexibility and adaptability to various feedstocks employe d finishean d d products sought s regardA s feedstocks s possibli t i , n thii e s process to employ uranium and plutonium nitrates either mixed or separate.

As for the sinterable powder, this can be prepared so as to meet various sintering procedures. e solubilitTh n concentratei y d nitrie c th aci f o d sintered pellets obtaine s excelleni d solutio; t n efficiencie f betteo s r than 99.95% have been obtained on non-irradiated pellets. Also, the composite material, at every stage of the production process, is soluble in nitric acid, thus guaranteeing easy recyclin f productioo g n rejectd an s further enhancing the advantages of initial mixing of o thliquitw e d nitrates.

Reproducibility It should be noted that the reproducibility of the characteristic e pelletth f s excellento si s y virtu,b e e simplicitth f o f the,proceso y f productioo s e th f o n sinterable powder and the ease with which the parameters determining these characteristice b n ca s controlled.

V - CONCLUSION e NitroTh x process constitue n importansa d originaan t l step forward in the preparation of composite powders. Economically well placed t resolvei , delicate th s e problems of plutonium-bearing aerosols or effluents. s flexibilitIt y make t possibli s o envisage t e us s it e jus e reprocesssins easila tth t a y g plant staget a s a s any other site. The Nitrox process thus plays a part in completing the fue ln ensurini cycl d an e g safet n handlini y e th g plutoniferous materials arising from reprocessing.

230 RECENT DEVELOPMENTS IN THE PURIFICATION OF URANIUM RECOVERED FROM IRRADIATED MATERIALS

P. DE REGGE, G. COLLARD, A. DANIELS, . HUYSD . SANNEL , N Nuclear Chemistry Department, Studiecentrum voor Kernenergie (SCK/CEN), Mol, Belgium

Abstract Experience is reported on the recovery of hiqhlv enriched uraniun fron tarqets irradiate e productioth r fo d f isotopeo n r fo s medica w lmateria ra use e Th .l consist f uranateo s n filteo s r tex- tures, paper tissues and plastic bags. The uranium is leached out by nitric acid ; organic material is destroyed by hydrogen perox- e filtereth d idan e d solutio d intfe a oPurex-typs i n e extraction cycle. Uranium is extracted with tributylphosphate dissolved in dodecane e extractioTh . n cycle using mixer-settlers consistf o s eight decontamination stages, eight scrubbing stages and twelve stripping stages. The decontaminated uranium is further handled outside the shielded facility in a qlove-box. Using tenoyltrifluoroacetone as an extractant the zirconium, niobium and plutonium quantities still present with the uranium are removed until the final product specification e reachedar s e uraniuTh . s finalli m y precipitated from a buffered ammonium acetate solution in the form of uranium peroxide, which is filtered off and calcined at 850 °C to U_00. J O e finaTh l product meet e specificationth s r recyclinfo s g inte th o targets with respecs residuait o t l radioactivity (less than 2600 s plutoniuBq/g)it o t , m content (les s tota it tha l o 1 ppmt n d an ) impurities wit a boroh n equivalen f leso t s tha 1 ppmn . Presently about 2440 g of uranium have been treated in five campaigns. The chemical yield of the wet chemical operations is aroun 6 percen9 d t including samplin r producfo g t characterisations and analyses.

. 1 INTRODUCTION

e productioFoth r f "Moo n , 131133d IXan e radioisotopes, highly enriched uranium targets are irradiated in different reactors by the Belgian National Institute for Radioelements (IRE), The targets typically consist of 4.2 g uranium metal with an 235U % cla 3 9 d enrichmeninto t f aluminiuoo 9 8 aboug f a o 0 t 3 t n i m tubular form e targetTh . e reprocessear s d shortly after their

231 irradiatio y dissolutiob E IR t mixtura a nn i n f sodiuo e m hydroxide and sodium nitrate. This operation provokes the liberation of xenon in the gaseous form and the dissolution of iodine and molybdenum as I~ and MoO^2" respectively [ l] . The uranium and most of e otheth r fission product a e precipitatlef ar s a t f hydroxideo e s r uranateo f undefineo s d composition e precipitatTh . s filterei e d and washed on a glass-fiber filter. The filters are then collected into a polyethylene bag and some paper tissues used for cleaning the dissolution and filtration equipment are added. After treat- f uraniuo men g f aboue o materiat 0 th m 14 t s storei la stain n i d - less steel containe d transportean r o SCK/CEt d N Mol.w Ovefe a r year f operatioo s f highlo ng k yabou 5 enriche4. t d uranius ha m been accumulated and its recovery and recycling into the process were considered. This decisio s basewa nn economicao d l reasons a s n inventoro wel s a ld supplan y y considerations e procedureTh . s used to recycle and purify this material and the experience thereby obtained are reported in this paper.

2. CHARACTERISTICS OF RAW AND FINISHED PRODUCTS

The raw material obtained from IRE in a stainless steel container typically consist f sodiuo g f abouo s m 0 16 uranatt e witn a h 235U content of 130 g, packed together with 35 g of glass fiber filters, about 80 g of paper tissues and 20 g of polyethylene Decay head radiolysian t s during storage usually transfore th m whole content into a brown-coloured adherent residue whose transfer from the container is only possible with spoons and scrapers. The characteristic d producten e th , f Uo s e well-define0 ,ar : d 0 0 o o - combined total impurities should not exceed 2500 ppm with a limit of 500 ppm for any element and specific lower limits for elements with high boron equivalenc; e

e betd gammth an a - a emitting impurities shoul t exceeno d d 260q 0B r gra pe f uraniuo m ; m e plutoniuth - m content shoul d e spontaneoulesb dan sm thapp 1 n s o fissiont transplutoniu e du s m elements shoul t exceeno d d 1500 min"1 .

3. PROCESS FLOW SHEET

Taking into accoun e characteristicth t e feeth d f o materias l and the associated radioactivity levels a first extraction cycle e Purebaseth n xo d proces a shielde d carrien i an s t ou dd cels wa l considered. Because higth hf o eenrichmen e e uraniuth th d f o tan m

232 restricted quantities of material involved, small mixer-settler batteries of critically safe geometry were chosen to form the heart of the extraction system. The capacity of the nixer-settler batteries being 500 ml hour"1 e aqueouth d o organit an s c phase flo we strippinratith n i o g (HC) section being set at 1.43, an organic phase flow rate (HCP) of 200 ml h"1 is obtained [2]. This flow ratio is the determining factor because a single organic phase passes all three of the batteries. e decontaminatioTh n batter s operatei y a hig t ha d loading factor n dodecâni P d tolerateTS an eusin% 0 3 g a concentratios g 0 8 f o n e organith n I"i 1 c n aqueoua phas d an es phas4 o et acidit5 3. f o y N. This leads to an active feed flow rate (HAF) of 50 ml h"1. In practice the feed concentration (HAF) is lower at about 230 g l"1 e uraniuorganith d an cn phas g l" e s loaderesultin8 1i e.5 Th o t d g throughput of the installation is then 12 to 13 g h"1. The scrubbing flow rate (HAS) was calculated from the condition that the organic to total aqueous flow rate should exceed 1.85 (for 2.5 l/kg U in HAS). This results in a scrubbing flow rate of 40 nl o removt N n acidite 5 a e use 1. t ar a df 1 o y Actuall. h" 1 l h" m 0 5 y zirconiu d niobiuan m m fission products A .spli t scrubbing phase wito differentw h t aciditie s beeha s n t bee considereno n s ha t bu d implemented e characteristicTh . f thio s s flow shee e presentear t d in tabl. I e

TABLE I : Flow sheet characteristics for the decontamination cycle

Stream 1 2 3 4 5 6 7 8

Code HAF HAS HAX HAW HAP HCX HCP HCW

Flow (mlh'i} 50 40 (50) 200 90 200 285 285 200

U cone) (M . 1.35 - - - 0.34 - 0.237 -

U cone, (gl"1 ) 320 - - -- 80 - 56 -

H+(M) 4 3(1.5) - 3.6(2.7) 0.05 0.01 0.01 -

TBP (vol.?) - - 30 - 30 - - 30

Operatio f thio n s process requires additionnai equipment sucs a h dosing pumps and storage tanks of the appropriate capacities. They are shown on figure 1. The combined aqueous stream leaving the decontamination battery (HAW s continuousli ) y evaporatede th , distillate being low-level aqueous waste. Similarly the uranium product solution is continuously concentrated to reach a final volume of about 2.5 liter. This solution containing the decontaminated uraniu s transferrei m d froe shieldeth m d facility inta o

233 O! O-

(T3

X 3

er (O

OJ o>

+J o fö S- -(-> X LJÜ

CT)

234 r furthefo glov x bo re purification, oarticularly from zirconium and plutonium. Therefore batch extractions with snail volumef o s xylene containing 0.5 M tenoyl trifl uoroacetone are carried out. r thiFo s operatio e finad als th n an vien i of l o w precipitation, packin d shipmen e an uraniug th f e solutioo tth n s subdividei n n i d quantities containin . AfteU ge phas g th abour 0 e 13 tseparatio n d whean n satisfactory decontaminatio s achievedi n e uraniuth , s i m precipitated as uranium peroxide in a buffered ammonium acetate solution. The precipitate is washed, filtered and calcined to

. 4 EQUIPMENT

4.1. Hot cell equipment All equipment for the dissolution, the clarification and the first extraction cycle is mounted on a working surface of 200 x 65 a cel n i l m shieldec f leado e manipulationdm Th c .wit 5 1 h e ar s carried out with three pairs of masterslave manipulators on 3 working positions. An overhead crane of 250 kg load capacity and a service area behin workine th d g tabl e alsar e o provided. e dissolveTh r equipment consist a dissolutio f o s n vesse d offan l - s treatmenga t limite a condense o t d d sodiuan r m hydroxide washing bottles connected to a vacuum pump. The extraction equipment consists of three sets of mixer settlers wit e extractio 8 stagehth r fo sd scrubbin an n g operation2 1 d an s e strippinth stage r fo s g operation. Storage tank e providear s d r feefo d solutions, reagent d effluent an se extractio th f o s n operation o continuouTw . s evaporator e providee concentraar sth r fo d - tio f higo n h level linuie uraniuth r d fo mwast d producan e t solution. Liquid transpor e storagth o tt e fro d tankan m s doni sy b e vacuum or air lifts. Only one pump is installed into the shielded r celsupplyinfo l e higth g h active feed solutioe firsth o tt n mixer settler battery.

4.2. Glov x equipmenbo e t The glove box equipment consists of ordinary laboratory scale instrumentation for the mixing, batch extraction, precipitation and filtration operations. Final conversio o IL0 t ns carrie i t ou d 0 j o in a furnace using silica crucibles.

5. SEQUENCE OF OPERATIONS

The sequence of operations for the recovery of the uranium is e followingiveth n i n g paragraphs.

235 5.1. High active feed preparation

w materialra e Th s obtaine e storedar fron E stainlesi dIR m s steel containers and have the composition given in section 2. figur 2 eshow a blocs k diagrae feeth d f preparatioo m n procedure. e uraniuth Mos f o s trecoverei m y leachinb w dmateria ra e th gl with 9 M nitric acid. Insoluble material, glass fibers and plastic are retained in a basket, transferred onto a 2.2 u, filter paper and thoroughly washed. After drying this residue is transferred back e initiath o t l stainless steel container w contain.fe Testa n o -s ers have shown that the residue contains less than 2 % of the uranium ; systematic gannaspectrometric measurements of the residuee beinar s g carrieo verift t you d their uraniun content using 1!+1+Ce-Pr as an indicator of fissile material.

U TARGET . FP . Pu PAPER « POLYETHYLENE I HNO 5 1, , g 0 20 - 0 « 9 M

DISSOLUTION IN NITRIC ACID HNOj 9M

UOj SOLUTION HLTRATO D WASHINNAN F GO E RESIDUATH L CAKE WASHING BOTTLE CONOENSOR UOj u P » f f

PARTIAL EVAPORATION

30V. _L 0,5 I DESTRUCTIO F ORGANICNO S AT BOILING TEMPERATURE

FLTRATtON - WASHING

HjO

FEED ADJUSTEMENT

HAF FEED

Fig. Bloc2 .F preparatioHA k e diagrath f o nm froe uraniuth m m target

236 o destroT e organith y c materia n solutioni l , hydrogen peroxids i e added and the mixture is boiled. The solution is then filtered a second tim n enucleopor^ throug4 0. a h e filter r criticalitFo . y reason e quantitieth s e streateb thaa singln n ca ti d e operation are limited to 550 g 235y which corresponds to the combined solution obtained from four to five raw naterial containers. The final uraniun concentrations and nitric acid molarity are then adjusted y evaporatiob M 4 o t r additio o n5 3. d f respectivelo nan M 9 0. o t y concentrated nitric acid. Typicall 5 lite2. y f feeo r d solutios i n then obtaine e d higstoreth an d h n i dactiv e feed tank. This feed preparatio ne criticaphasth s i ee proce e lwholth th ste f n -o ei p durd particularlan e e leachinth y d filtratioan g n phase e relaar s - tively slow e characteristicTh . e feeth d f o ssolutio e givear nn i n tabl. II e

TABLCharacteristic: I I E Procese th f o ss Solutions Typical value f curreno s t runs

HAF HCP Decontaminated Final Uranium Feed Solution Product Solution Solution

Volume (ml) 2100 2475 2475 U g r1 259 214 214 Pu mg T1 25 14 1 < 1 i^Cl- q B e 1.19 lu12 n.d. n.d. llfLPr Bq T1 1.19 1012 n.d. n.d. 137Cs Bq l""1 1.94 1010 n.d. n.d. 1 106T q RB u 5.62 1010 3.3« 10 6 -2.06 10 0 125Sb Bq l'1 4.09 10 7 n.d. n.d. 1 T 95 q ZB r n.d. 2.76 110 ~ 3.00 Iff« 95Nb Bq T1 n.d. 1.15 106 1.16 10 5 1 15T "*Eq B u n.d. 2.53 105 2.53 105

t determinedn.dno . .

5.2. Solvent extraction purification

The solvent extraction operation is carried out on a continuous basis usually lasting 50 hours for the actual production, precede d followedan d 8 hour r respectively b fo s y battery equili- bratio d clean-outan n e reagentd througTh .fe e ar hs dosing punps from the supplies outside the shielded cell and the process is operated according to the flow sheet described in section 3 and show e n figurhigi nTh h. 3 eactiv e waste solutio s evaporatei n t a d . Radiolysi 1 h" l m abou s0 product10 t d organian s c remnante ar s gradually destroyed whe d inte nboilinfe th othee ar yo gn acid an d foamin r violeno q t reactions have been observed e volum.Th e reduction factor is limited because of the presence of significant

237 STRIP SOLUTION SCRUB SOLUTION HAF TBP/OODECAHE HNOj VH M HNOj 1.5 M U/Pu/FP HNO, 3,5-4.5 M (<:) 6 C 280 ml IT1 50 ml . h'1 50-55 ml h'1 200 ml h"1

a 0 50 1 o SCRUBBING i ml h" 0 a 0 UJLJ EXTRACTION i STRIPPING « . a t DISTILLATE DISTILLATE U.W LLW

' ——— EVAPORATION 1———— EVAPORATION

HCW (CONCENTRATE] CONCENTRATE ORGANIC 1 U P HC 1 HLW WASTE

Fig. 3. Flow sheet of the liquid-liquid extraction process for the gross purificatio f uraniuo n n

quantitie e f feesodiuo sth dn i msolutio n which eventually cris- tallises as sodium nitrate. Therefore regular transfers to the HAW storag e carrieo avoit ar e t dou d cloggine systemth f o g. Radio- lysis and formation of gas locks has been observed in the feed lines and an outqassinq device has been installed between the HAF fee e pumpdth tand . an k e higTh h organic phase loading avoids crud formatio e phasth t ea n boundaries and suppresses fission oroduct coextraction [3]. No particular problems have been observed in this section. To favour the decontamination fron zirconium-niobiun in the scrubbing battery, it is operated at 54 °C. The uranium product solution is continuously evaporated and concentrated to reach a concentration of about l H of uraniuin. Initially considerable foamin s observewa g n thii d s evaporatos a r radiolysis products accumulate and acidity increases. The evaporator is now initially loûded with a small quantity of concentrated nitric acid and organic material is now destroyed immediately in the boiling acid. Thereby foaming has been completely avoided. To reduce the fire risk s during the extraction campaigns the shielded cel s maintainei l d under nitrogen atmosphere e charac.Th - teristics of the product uranium solution are given in table II. Since no particular attention has been paid to plutonium until this stage, it is codecontaminated and coextracted with the uranium to a large extent.

238 5.3. Zirconiu d plutoniuan m m removal

e solutioTh s obtainea n d froe evaporatoth m r doet meeno st yet the radiochenical and chemical purity specifications but is sufficiently decontaminated to be handled without shielding in a glove-box. Considerable purification is still obtained at the precipitation stag t zirconiubu e d plutoniuan m m will coorecipitate quantitatively. Therefor a batce h extraction procedur s appliei e d o reac t e desireth h d specification r thosfo s e elements prioo t r the precipitation. The flowsheet of this operation is shown in figur. 4 e

A TT l m 0 10 x 3 3 BATCH EXTRACTION WITH TTA

HYDROXYLAHINE

NjNOj

VALENCY ADJUSTMENF O T Y OXCO-REDUCTIOB N CYCLES AND PH AQJUSTEMENT TO 0

.2-x 59 ml TTA

2 BATCH EXTRACTION WITH TTA

8 BUFFERIN2. = PH T GA

PRECIPITATION OF U04

FtTRATION/WASHWG DRYING

CALCINATION AT ISO • C JL

Fig. 4. Block diagram of the final purification of U

The extraction is carried out with 0.5 M tenoyltrifluoroacetone (TTA) dissolved in xylene. Three extractions are carried out at a volume ratio aqueous/organi a batc n f i abouo hc 0 1 extractiot n installation to remove essentially all of the zirconium. The plutonium valency is then carefully adjusted to PiA* using hydroxyl-

239 aminé-nitrate and sodium nitrite ; the acidity is reduced to less than l N and plutonium is quantitatively extracted v/ith 2 batch extractions (TTA/xylene) in a volume ratio of 20 to 1. The characteristics of the solution after this extraction stage e alsar o presente n tabli d. II e

5.4. Uranium precipitation as peroxide

Precipitatio e uraniuth f s peroxido na m s beeha en selected because e excellenth f o t removal fron solubl d hydrolysablan e e cations that can be achieved in this procedure. Different operational nodes have been reporte e literaturth n i d e [4][5]. Taking into account the characteristics of the solution and the settling and filtration of the precipitate, the following procedure has been developed. The uranyl nitrate solution is buffered with ammoniu s reachedmi acetat 8 2. .f eo untiH p a l Usuall n w orange-rea gramfe f o a sy d precipitat e formear e t a d this stage containin a smalg l quantit 5 weigh3 f o uraniuo t y t 0 (3 m e solutioTh . s filtereX) i n d w uraniuprecipitatean no d s i m y b d the slow addition of hydrogen peroxide until about 1.2 ml/g U are reached. During this operation the pH decreases again and reaches a final value around 1.2 e brighTh . t yellow precipitat s allowei e d o settlt d washean e d several time y décantatiob s n usin a washing g liquor containing ammonium nitrate, ammonium acetate buffed an r excess hydrogen peroxide. The precipitate is then easily filtered on a 7 |j.m filter paper. A final washing with water removes the washing liquor salts e precipitatTh . s driei e t moderata d e temperature and transferred into a silica beaker. It is then converted C durin° 0 6 hours1 g85 t a .y heatinb r ai 0 n i U g o t 3 8 5.5. Chemical yield and material balance

The quantities of uranium in a single stainless steel container are not well defined in spite of its high enrichment and the associated value. On the average, the total quantity in a numbe f containero r s corresponds roughle declareth o t y d valuey b s IRE. The first measurement concerning the chemical yield of the proces e clarifie th s mad i sn i eF solutionHA d . Samplin r confo g - centration and free acid determinations before and after the decontamination cycle consume 4 to 6 g of uranium. The high active waste solution typically contains less than 3 g of uranium and the other waste streams contain négligeable quantitie s showa s n i n table III. e chemicaTh le decontaminatioyielth f o d ne shieldecyclth n i e d cels thui l s around 98. f uraniuo . Abou5 % e g th s los2 t i m o t t TTA/xylene solution and 3 to 4 g into the orange-red precipitate.

240 Wast: TABL I eII E Stream Characteristics

Organic Stream Distillate Distillate Phase HAW HAF U product HCP

) l l urn(m Vo e 6850 14 500 11 050 2800

i l- g m U 2.7 10.4 97 < 1300

1 l" g Pm u 0.0021 0.0006 0.035 < 1

""Ce Bq T1 1.50 107 8.4 106 2.6 106 1.33 1012

l""Pr Bq T1 1.57 10 0 8.4 106 2.6 10« 1.38 1012

137Cs Bq T1 6.5 105 5.4 105 n.d. 3.9 101° 106 1 Ru Beir 2.46 10 4 3.9 105 9.2 106 7.0 101°

125 Sb Bq T1 7.6 IQ5 8.9 icy* 1.18 105 n.d.

95Zr Bq T1 n.d. n.d. 6.7 105 n.d.

95Nb Bq l"1 n.d. n.d. 6.3 105 n.d.

n.dt determinedno . .

Filtrates and washing liquors contain typically less than 0.2 g of uranium. Samplin e finath lf o gproduc t further g consume 4 o t 3 s so that the net chemical yield of the operation is around 96 %. The overall material balance is hampered by the uncertainty on the raw materials and the inaccuracy of the volume measurements in some storage e thankshieldeth n i s d cell.

. 6 CHARACTERISTIC D PRODUCEN E TTH F O S

The characteristics of the end products in the different campaign t identica no o smalt e e ar ls du lprocè s adjustments carried out as a result of earlier experience. Particularly the con- e scrubbinditionth r fo s g stage e plutoniuth , m e removath d an l peroxide precipitation have been modified with respect to their initial values and yielded considerable improvement of the end product purity e characteristicTh . w achieveno s thee sa e ar yar d summarised below.

6.1. Product stoechiometr enrichmend an y t

The uraniun content of the converted oxide is typically within 0.3 % of the theoretical value of 0.8466 for a product with

241 this isotopic composition. The 235U content is around 92 weight percent e 23lTh *l. i conten f 0.7o t 0 weight percen s extremeli t y e procesth usefu r sfo l analyses before decontamination because they can be based on alphaspectronetric measurements of 231+U.

6.2. Radiochemical purity

Presently a radiochemical purity better than the specified 2600 Bq/g L) is achieved. The earlier batches however showed residual activities of 95Nb (up to 3200 Bq/g), 106Ru (up to 500 Bq/q), I25sb (up to 1200 Bq/g), 137Cs and i^Ce (both up to 3900 Bq/g) an do 110 t 151 p *E0 (u Bq/g)u , Plutonium concentrationm pp 0 6 o t p u s e earlieith n r batches havw beeno e n lowere o lest d s tha 1 ppmn . Transolutonium element e undetectablar s e finath n li e producte .Th thorium conten s variabli t d range an e0 ppm 6 so .t fro 0 1 m

6.3. Chemical purity

e chemicaTh l purity fulfill l specificational s d onlan sa y few elements exceeding 10 ppm are detected by spark source moss spectrometry. Most important in the earlier batches was zirconiun t currenbu m pp t batche0 50 o st rang0 40 ate belo 0 ppm1 w . Similarly the residual phosphate content has been lowered from 80 ppm to about 5 ppm. Chlorine ranges from 5 to 70 ppn and, rather unexpectedly, sulfur is systematically present in quantities aroun 0 ppm 15 e origi d.o Th t frof thi o 0 n5 ms sulfur contamination t cleaino st thia r s momen t coulbu t d come from radiolysif o s TTA.

e totaTh l boron equivalenc l chemicaal f o y l contaminants doet no s exceed 1 ppm boron equivalent whereas the recommended specification is set at 2.5 ppm. An exampl a specificatio f o e n sheet accompanying each product batc s givei h n tabli n. IV e

242 TABL : Typica V ProducI Ed En l t Specifications

Uranium batch UT4-020

Atom percent 231* U/U 0.706 Weight U,0 : 151.4„ o 6 JO 92.199 Weight % U : 84.54 g

236 u/u 0.690 Uranium : 128.044 g

238UAJ 6.405 Atomic Weigh : 235.23t 7

Impurities

Radioactive Chemical / 1 1 —— - — -- Bq/g U U y / |Jy (1 a = ± 50 %) 95Zr < 185 Li < 0.002 95Nb < 1110 B 0.083 Ma 0.61 5 31 < 1Q6Ru Al 2.0

l37Cs < 85 Si 0.22 P 1.6 0 37 UtCe-P < r Ça 1.1 Ti 0.030 15"Eu < 1130 V < 0.001 Cr 0.079 Mn 0,013 Fe 0.89

Co 0.072 Ni 0.56 Cu 0.006 Actinides Zn 0.096

Ag 0.047 Pu < 1 Sn 2.4 Ba < 0.010 Th 1.1 Pb < 0.10

S 53 Cl 5.8 K 18 Zr 2.6

7. CONCLUSION

The experience has shown that highly enriched uranium can be recycled and converted from irradiated targets into nuclear grade base material with relatively simple e singlmeanson y B e. Purex- based extraction cycle, followed by specific chemical separation techniques for the removal of plutonium and some residual fission products, decontamination factors in the range of 105 to 106 are obtained. Further purification is achieved by the precipitation of uranium with hydrogen peroxid well-specifien i e d conditions where- n easila y b y filterable precipitat w abous obtainedi no e o tt p U . 2.5 kg of this material have been processed without particular

243 difficulties and with a chemical yield of about 95 %. About half o samplint e e lossedu d analyticath an e gf ar o s l waste. Unrecover- able quantities discarded to the process waste are below 0.5 percent.

REFERENCES

[1] SALACZ J., "Reprocessing of Irradiated 235U for the production of 95Mo, 131I,133 Xe radioisétopes", revue IRE, ^9, 3 (1985), 22.

[2] "Purex Technical Manual", HW 31000 (1955).

[3] OCHSENFELD W., BAUMGARTNER F., BAUDE U., BLEYL HJ., ERTEL D. and KOCH G., "Experience with the reprocessing of LVR, Pu recycl d FBR-fue e MILLan e th n Ii lfacility " KFK-renort 255; 8 Proc. Internat. Solvent Extraction Conf., Toronto, (September 1977).

] CURTI[4 S M.H., "Continuous Uranium Peroxide Precipitation", ISO-repor 7 (1966)20 t .

[5] GARNER E.L., MACHLAN L.A., SHIELDS VI.R., "Uranium Isotooic Standard Reference materials" NBS special publication 260-27 (1971).

244 THE REPROCESSED URANIUM CONVERSION: TEN YEARS OF OPERATION OF COMURHEX

R. FARON Pierrelatte Plant, COMURHEX, Pierrelatte, France

Abstract

This facility was in the middle of the 1960 a pilot plant to convert UF6 to sinterable U02. In the 1970 this plant after some modification was rede- signed and provided for the conversion of UNH coming from LWR to UF6. At the end of 1985, 1300 t/U has been converted and delivere o enrichment d t plants. The conversion of reprocessed uranium from light water reactors is : - slightl yo 2,2 t enriche % 5% 9 0, d - contains impurities

Durin e conversioth g n process COMURHE o eliminatt s ha X e not only chemical but radiochemical impurities otherwise the material would never meet the specification for re-enrichment. In this paper we are looking the main impurities and decontamination effect of the COMURHEX process.

- HISTORICAI L . 1 Slightly depleted uranyl nitrat . o UF6(set e1) g Fi e Since 1970, COMURHE s converteha X d about 10.00U t/ 0 of reprocessed uranyl nitrate coming from the French graphite-gas reactors. Thi s slightli s y depleted uranium and, provided some precautions are taken, e converteb n a normaca n t i di l conversion plant. 1.1. Hightly enriched uranium hexafluoride to U02. In 1967 this facilit s beeha ye n th buil r fo t "deconversion f nightlo " y enriche o product 6 UF de sinterable U02.

245 UNH

u Ml»«.y mvcMto . v M lu.tm« i ,

Fig.l. Uranium from reprocessing.

h U/ g Capacitk w fe : y Total production : 1967-1970 200 kg U Proces t proceswe : s s ) 1 (se g eFi 1.2. Slightly enriched uranium hexafluoride to U02 1969-1970 deconversion by wet process of UF6 slightly enriched 3,5 % 235 Capacity : 8 kg U/h Total productioU t/ 0 6 n 1.3. Slightly enriched uranyl nitrat 2 (se U0 g 2.3. eFi o et ) e 197 Ith e nfacilit 0th modifies e conversioi y th r fo d n of reprocessed uranyl nitrate in UF4 and erection a fluorinatio f o n plan o convero t UF6tt 4 .UF t Capacity.originally the capacity was 200 t U/Y g U/hk 0 198n ) i (3 e capacit6th arouns i y 0 t/U/40 d Y (60 kg U/h).

PROCES- 2 S DESCRIPTION UNH UF4 After acceptance of uranyl nitrate by controlling, . Chemical elements (U contained, and impurities) . Radiochemica) l y activit ß a ( y

246 . V* " T I 7. l y'e.t n f,\T.r7 Yr I kLJ

Fig.2. F . U Conversio o t H UN f o n

Fig.3. Conversio o UF,t . . UF f o n & 46

247 2.1. UNH is pumped into a precipitation tank where it reacts with ammonia solution to give ADU ammonia diuranate. This reactio s controllei n y temperaturb d H p d an e measurements.

2.2. The second step is a filtration with a vacum filter where the excess liquid is drained and ADU is fed into a mixer. Because the thixotropic property of ADU s possiblii t U sludg AD o pumt e y mean b eth p a s metering pump into the following step.

2.3. Calcination 2 ReductioU0 » - U AD n This drying, calcination and reduction is made in a rotary kiln heate y electricab d l furnace a firs n i , t part of the rotary kiln ADU is dried, the second step consist a transformatio n i ss i 3 U0 d nan int3 U0 o reduced into U02 by a countercurrent of hydrogen obtained froe crackinth m f ammonio g a temperature along this rotary kiln is from 130° C to 650° C at the exit.

2.4. Hydrofluorination From the first rotary kiln U02 is introduced to a second rotary kiln through an intermediate air tight valve, after the rotary kiln,hydrofluorination is completed into a screw. HF is introduced at the end of the screw. Through an air tight valve UF4 is collected into drums before the fluorination step.

2.5. Fluorination UF4 -> UF6 After grinding UF4 is fed into a flame reactor by meana dosin f o s g scre o controt w e ratith l o between UF4 and fluorine, in order to obtain a maximum efficiency a fluorine excess is necessary but the effi- d smalciencan l% f fluorinatio o yquantiti0 10 t no es i n f unburno t produc s collectei t d int a druo s a m 7 s filterei UF5F1 6 4 U UF y mean. b d, F9 s 2 U , F2 2 U0 f sintereo d "monel filters y porosity 0 (1 " ) unburnt e flamth material f eo reactop e to recycle ar e s rth t a d until the accumulated dust reaches a radiation level Remu 0 .o50 f After filtering the purified UF6 passes through two cold traps assembled in line and will be solidified at -40°C. Ae capacits th soo a cols a nf do y tra s reachei p e th d coolan a heatin s switched i t an gf liquiof d d secures

>48 the melting of UF6 which is drained into an appropriate transportation container.

3 COMURHEX Experience in Reprocessed Uranium 3.1 COMURHEX, in its Pierrelatte plant, has been converting reprocessed uranium from light water fuel elements since 1972. Since then, more than 40 batches have been converted representin a totag f 130o l , 0T These batches comprised fuel elements from various burnups from 15 000 to 33 000 MW/d/t. Before reprocessing these fuel elements have been stored in cooling pools years6 durin o t .3 g e initiaTh l enrichment leve U 235( l )f theso e fuels . Consequentl% wa s4 o frot 2 m e havw y e different types of reprocessed uranium : - U 235 content between 0,7 and 2 % - U 232 content between 0,03 and 0,15 ppm compared 5 23 U o t - U 236 content between 0,135 and 0,435 % compared to U235 Through these various batches, COMURHE s acquireha X a d large experienc n thii e se havw fiel d e an drecentl y made a general study about the characteristics of the different batches. This stud s prove 2 ha yconten 23 da funcU s thae i t -th t tio f boto n h burn-u d coolinan p g time before retreat- ment \ but the measurements seem to be always under the theoretical values. Example isotopiTh : e c compositio R fueLW l a elemen f o n t after a 3 years cooling for a 33 000 Mto/d/T burn-up. Isotope % in weight U 232 10-7 (100 ppb/U 235) U 233 IQ'6 (l ppm/U 235) U 234 0,02 5 23 U 0,09 4 0, 6 23 U 3.2 Moreover, reprocessed U. has a, ß, y activities.

The a activity comes in general from the transura- t essentiallbu , rriaCm n, yelementsAm plutoniu, Pu , ,Np m and neptunium 237.

349 The ß activity comes essentially fron ruthenium 105 which gives volatil fluorfdes and oxyfluorides with a complex behaviour during the fuel cycle. The y activity comes essentially from thallium 206, daughter product of the U 232 with a short life. 3.3.Influenc e variouth f o es isotopic compositio f uraniumo n . s a daughtefiliatio ha s i d 2 an r23 6 nU 23 produc - u P f o t products with a ß y activities. E specification o UDO nowt p e th , 2 give23 U sa slimi t 232/ U f ( 0,11o m U 0pp 235) . This value will changn i e e o futurincreasint th e du e g cooling period d highean s r burn-ups. Therefore, it is necessary to adapt the level of U 232 in enricher's specification. We think this value could be 3 times higher than the current level today this specification is studied by AST M Nuclear Comittee. We'll have to take into account in the fuel cycle (conversion, enrichment, fuel fabrication) of the changing in the U 232 filiation products. The first of them, thoriu 8 (perio22 m d 1,91/year) producen no a s volatil fluoride that can be filtered from UF6. n reprocessei 3 23 U - d uranium remain n extremela t a s y low concentration and far below the enrichment specifications. 6 follow23 U e increasin - th s g burn-upse th s soo A .s a n burn-up reaches 33 000 MW/d/T, concentration of ex reprocessin . reacheU g d evean sn overpassef o % 0 5 s U 235. This isotope has a high cross section and helps to generate within the reactor artificial isotopes such as 232U neptuniu d .an 7 23 m Moreover a neutroni s i t i , c poison thatR needLW n i s an overenrichment in U 235. This overenrichment is around 0,5 % of supplementary U 235 by 1 % of U 236. - Finally, U 234 interferes in radiochemical activity of reprocesse a lowed t uraniualsos a i r t d degreebu ,an m , a neutronic poison.

- DECONTAMINATIO 4 N PROCESE EFFECTH F O TS 4.1. obvious Firsi t i t s thae conversioth t e uranyth f lo n nitrate solution into UF6 does not carry out any varia- e isotopith tio f o n c assa e reasof uraniumyo th s i n t I .

250 e specificationth y wh s concernin mïnoe th g r isotopef o s e urany e th uraniusam th n i ele nitratar m e solutiod an n UF6. 4.2. The decontamination effect of the process can concern onle metallith y c impuritie e radiochemicath r o s l impurities (No comments about metallic impurities)- e firsIth n t ste f conversiono p , from uranyl nitrate solution into UF4, no purification effect can be noticed All the impurities are co-precipitated with the uranium. e chemicaAlth l l compounds have approximatel e samth y e behaviour as uranium. e seconIth n d step, that means durin e fluorinatioth g n 4 intUF o e n noticUF6th ca f e o a considerabl,ew e effect f decontaminationo . 4.3 .g activity The ratio elements must be divided into two groups :

The transuranian elements as Np, Pu which give alpha radiation e removeb n ca sd froe UF6th m . These compounds form hexafluorides at the highest valency with physical properties close to UF6, but the kinetic of fluorination t stablisno d the vere n w speciallan i ear ylo yu P r fo y absenc a fluorin f o e e atmosphere. Thee inclinear y o t d convert into a lower valency, as a solid product. So, the main part of them are found in the ash or dust whic e collectear h e e flambottoth th t f ea o md reacto r or filter. In any case, if a small amount of them can be found in the liquid UF6, they can be separated by filtering the liquid i)F6 through a 2 microns filter of sintered monel alloy, before fillin e containerth g .

4.4.ß and y activities The fission products as Ru, Tc (Zr, Nb, Cs) or the filiation products as Th (Pa ...) give beta and gamma radiations. Their behaviou s quiti r e different, according to the vapor pressure. The fluorides of ruthenium and technecium have approximately the same vapor pressure as UF6. For them, no significant deco-facto s noticedi r . For the other elements, the fluorides are not volatile and they are are collected with the dust. 4.5. Starting wit alphn a h a activite th f t 15.00yo a m dp 0 e processentrth f e outpuo y th , t materia s closi l o t e

251 150 dpm, that means a factor of decontamination of about 100e betd gammTh .an a a activitie e reduced ar sa lowe n i r proportion (factor 2 to 4). The process, as developed by COMURHEX, is able to convert and purify up to 400 t per year of reprocessed uranium, at a maximum U5 content of 2.25 % and produces an output material fullfilling today's requirement e enrichth f o s- ment plants. Though the yield of the process for a more 10 tons amount cae fixeb n t arouna d d 99, wit% 5 h respece uraniuth f o tm balance e uraniuth ,e dus f s bouno abouth i mt % o t d1 t output. Thise % uraniuoutputh e dus 1 s th i tf n o ti m tribute, we have to pay to improve the quality of the UF6 during the conversion. In short, accordin e amountth o t g , this mean n immediata s e dus% uraniu1 t a prod man - 6 recoverin o 98,UF t % 58 9 f o g duction. Concerning the dusts, it seems now that the best solution consists of a specific and appropriate treatment of the waste n ordei s o stort r e them definitivel n gooi y d conditions.

5 - URANYL NITRATE SOLUTION ACCEPTABLE BY COMURHEX 5.1. Usuall e referencth y e specification e ASTM ar st som f ,bu o e them mus e defineb t d accurately. COMURHEX's propositions n givca e some informations abou e sentencth t e "values agreed between purchaser and manufacturer" after an experience of more 1000 tons d concernin,an g onle alphath y , betd an a gamma activities. 5.2. a activity As starting point, we are considering that the contamination of the air in the workshops must be, in any case, lower than 75-7 6p Ci/m f air 3o r natura .Fo l uraniu e limith ms i t 120 pCi/m3 of air. t thige so T value e alphth , a activity mus e limiteb t d an d calculated from three groups of elements. The a activity of the transuranium elements as Pu 238, Pu 239-242 and Np 237 must not exceed 11.000 dpm. The acti- 2 mus 23 e les b tU vit sa activif o ythae th n d -46.30an m 0dp ty for Am 241 - Cm 241 - 242 and all the daughter products of U 232 underneath 25.000 dpm (Th 228, Ra 224, Rn 220, Po 216 212i B , l 208),T . e calculatioIth n e greatesth f o n t admissible concentration, e weightth f eaco s h group represen: t elemenU r fo t% 0 6 r transuraniafo 25% n 2 23 U r fo % 7 r daughte m fo plu. A % Cm d s 8 an r 2 product23 U f o s 5 . 3 .3 and y activity In orde o protect r t more efficientl e workerth y s againse th t radiations, we want to limit the total activity, due to the fission productu Ci/k 0 . 3 U g o t s r activityfo U r g gammfo /k i )U C a g y activitC i/k 7 1 d an y e uranyTh l nitrate solutions delivere a HaguL dy eb unti w no l e respectinar g these specifications. In these conditions e abl ar o assur t ee ,w e announce th e d yiel f conversioo d n wit n acceptabla h e conversion cost.

6 - CONCLUSION COMURHEX intentiow productione o built a s p i u nd nt facia y lit Pierrelatte plant with a nominal capacity of 1000/1200 tons U per yea URE: r X 2000. COMURHEX has already established a detailed engineering study for this new facility and has been able to realise a pre- feasibility study.

253 /Z PROBLEMS DUE IX .MPURITIES IN URANIUM RECOVERED FRO REPROCESSINE MTH USEF GO D LVVR FUEL, FROM THE POINT OF VIEW OF RECYCLING

. LEYSEE R Deutsche Geselischaft für Wiederaufarbeitung von Kernbrennstoffen mbH (DWK), Hannover, Federal Republic of Germany

Abstract

Specific impurities containe uraniun i d m recovered from reprocessin f usego d LWR fuel create problems at the diverse stages of the fuel cycle when it is recycle a reenrichmentvi d . These impuritie mainle ar s y fission productd an s actinides. They cause either handling problems (radiologie protection of personnel) or technical problems (such as neutron absorption by u ). 236 brieA f e impuritierevieth f wo s givei s n together wit probleme th h s they create.

Product specifications are briefly discussed, and suggestions as to desirable further purification (if it is possible) are made, among them proposed changes of specifications presently being discussed within RSTM.

It is most important, if further purification is considered necessary, to find the stage of the cycle where this can be done under the best economical conditions.

1. Introduction

Reprocessing of used LWR fuel yields two products: plutonium and recovered w uranium. The amount of this uranium is about 95 /o of the original fuel uranium fed into the reactor. Depending upon initial enrichment and burnup, its contents of U^^,. can be substantially higher than that of natural 235 uraniu U o / approximately)m 9 (0.0. o 8t thin I . s Wcase containt i , s valuable separativ e recycleb n advantagn a e ca o wor t dd a kan vi e reenrichment.

255 However, this recovered uranium contains impurities which are not - or only in extremely small trace containes- naturan i d l uranium. Thes mainle ear y fission product d actinidesan s wels a , s impuritiela s remaining froe th m dissolutio fuee th l f ndurino g reprocessing (elements containe n fuei d l cladding or provenient from process chemicals). These impurities may cause two types of problems: - handling problems mainle ,du radiologicao t y l protectio handline th f no g personnel alst ,bu o environmental problems - problems caused to reactor operation due chiefly to neutron absorbing effects.

e presenTh t paper give briea s f relevane revieth f wo t impurities which cannot be further separated by reprocessing, the main problems they cause, and suggestion desirablo t s sa e further purification whic chiefly hma y intervene during conversion to UF,. b Current product specifications ar.d proposal o changt s e thee alsar mo discussed in short.

. Specifi2 c impuritie n recoverei s d probleme uraniuth d an ms they create.

2.2.1 Activit f fissioo y n products

Current specifications compare fission product j3 and -y activity with e activitth f ageo y d natural uranium e latte.Th s naturai r l uranium whic s beehha n removo purifiet s a e o decas d y product d whican s ha h been storemont1 r t leasha fo d orden i t reaco t r n equilibriua h f mo shor. t liveh T d an deca y a P product , h sT

The activities of such aged natural uranium are (related to 1 gram of heavy metal): 4 0.6: Ci/gH8a v M (2.50 Bq/gHH1 2. ) 4 ß : 0.68 v Ci/gHM (2.52 . 10 Bq/gHM) 3 y : 0.16 v Ci/gHM (5.92 . 10 Bq/gHM)

The ratio of fission product activities in used LWR fuel to that of aged natural uranium give n indicatioa s f decontaminationo n factors required from a reprocessing plant for fission products. The following table shows some typical decontamination factors (DF threr )fo e currenR LW t burnups and cooling times of 1 and 2 years before reprocessing:

256 Burnup Cooling timyea1 er Cooling time 2years (MWd/kgHM) DP (ß) ) D(T F DF (ß) DF (Y)

30 6 2.+ 0 5.1 + 6 1.0 6 + + 6 0 3. 36 6 2.+ 1 6.8 + 6 6 + 1.1 4.0 + 6 40 2.3 +6 6 + 0 8. 6 + 1.17 4.8 + 6

Decontamination factors thus vary betwee . 8.1d n 0an 1.10 In general, modern reprocessing plants have a DF of about 10 for fission products (this is not specific for each nuclide).Thus for fuel with burnup f 4000so 0 MWd/k onld e yeaan gon y r cooling time, reprocessed uranium is within specifications concerning fission product activity. Following isotopes are worthwhile to be mentioned separately:

This isotope H contribute UN Y activit e d th an ß f o boto ye t s th h difficuls i d an separato t e withi Puree th n x process formt i s sa .A n volatilno e fluoride wilt i , l mainly caus a probleee th n i m conversion plant. However with long cooling time before reprocessing (the German WAV-plan plannes I t o reprocest d s fuel wit coolina h g time oyears)f7 problee ,th m vanishes.

- Te ) a 5 10 2 1/ v 9 9 This long lived ß-emitter shows a tendency to follow the uranium product. Due to the volatility of TcF (Re.fig.1), it is carried 6 through the conversion process and may create emission problems in the enrichment plant.

) d 8 36 = T ( 6 R10 u This isotope contribute y activit d an sf UNHß o ylargel.e th o t y However, long cooling periods strongly reduc problem.e th e e Th volatility of RuF, is lower than that of UF by a factor of about 0 1 (Re.fig.l)3 . This means tha a furthet r strong decontamination takes place during conversion .y contribut Herma t i e wasto t e e problems.

257 10

-1C» -50 50 100 150 200 250 Temperature! °C )

Fig. i: Vapour pressures of volatile fluoric! es

2.2.2 Transuranlc a activity

Transuranic isotopetotaa y b l e limite H admissiblsar UN n i d ea activit f 1500o y 0 dprn/gU (disintegration r minutspe r grape ef m o uranium). present a F U a limit n I f 150o t 0 dpm/g s admittedi U . However, this 6 value is considered to be too high at present. The reason is that especiall F yhav Pu NpF d ea simila an , r degre f volatilito e o that y t 6 6 F U (Re f o ,n contac figI . .1) t with metallic surfaces they decay 6 n volatilno e easilth eo t ytetrafluoride d thuan s s create problemo t s the enrichment plant. Ther t seemi e s that thee depositear y d mainly withi e feeth nd system presumablt ,bu y also withi centrifugese th n . This would cause an accumulation problem in the long run

1)According to a private information from Uranit.

258 Following transuranic isotopes should be mentioned separately:

- NP237(T 1/2 - 2-l°& a) As this isotope is chemically very similar to uranium, it is difficult to separate it by the Purex process. It mainly contributes to transuranic a activity in UNH and causes problems in the enrichment plant (see above).

- Plutonium

(T 1/2= PU23 88?= "7 a/PU -"^^o"4 23" 2 9= 6^ "W 6' These 3 Pu isotopes also contribute to a large extent to the a activity of UKH and cause problems in the enrichment plant (see above).

- Americiur d Curiuan a m (T : An\ = 433 a ''Am = 7.4 . 10 a/Cm = 163 d/ Cm = 18.1 a) 244 These isotopes contribute to some extent to the a activity of UNH. Furthermore, Arc.*, concentration increases with cooling time (due to ß-deca f P"^.-,)m isotopeo yC e Th - s decay relatively fast. n ordeI o fulfilt r e specificatioth l f 150o n 0 dpra/g r totafo u l transurani ca activity reprocessine th , g plant must hav commoa e n decontamination factor of about 10 for Pu, Am and Cm.

Experienc Karlsruhe th t a e e pilot reprocessin s showha g > nplan - (w t e followinth g distributio f transuranino activit* < c n »tfHi y :

% Np0 :6 % Pu4 :3 Am and Cm: 6 %. Thus efforts should concentrate ^ removamainl, th Np n f o yo l

2.2.3 Uranium isotopes

As uranium isotopes cannot be separated by chemical means, the problems they caus n onlyma masteree ca b y a judiciou y b d s choic f recyclineo g strategy. This may change in future with the availability of laser enrichment capacite , th »nic s o removhha t y e isotopes specifically.

259 Following uranium isotopes should be mentioned:

- U 232 1/2 U has a strong radiological effect due especially to strong emissio decaf no y Thiproduc. affecy sma stagel l al tT f o s 208 recyclin d particularlan g y enrichment e firsth , t decay produch T t Z/Ö being deposite solia s a dd fluoride.

This is evident from the decay series of U232"

208 22«„ 220__ «t 215, Pb 228^ R» R»i —» f Pb 60 3.1 1.91 3.61 55 0.14S 10.6 72 h Kin •In •tab!« yi yr d »ec •ec

The current specification limit for U is 0.11 pprc /U , which is very stringent and would be reached by used fuel according to present burnu d initiapan l enrichment standards afte year7 r f so cooling time (Re. fig. 2).

11. oa 31. •• > M.4 M M . M4« . M . 44 M . 41

10 a a Ccoolin5 g time)

i: o.

rt •-• CM

—i——i——i——i—n——i——i——i——i——i——i——i——i——i——i——i——i——T J4.00 M.M ».«• 40. «I 41.M 44. M 44.04 •«.» M.M 11 Burnup in MWd/k»

Amount of U-.,, particles as a function of burnup and initial U«-,, enrichment Basis 6 KW/kOrige3 « gP ncoolin 0 yearg1 timeo st 5 s

Fig. 2

260 - u a) 233 Vl 1/2 *•" " - This isotop mentiones i e specificationsn i d w lo .s Howeverit o t e ,du concentration s practicallha t i , o influencyn cyclee th n .eo

- U ) a 5 10 234 v' 1/2 "" naturan I l uranium this isotop concentratioa s eha f abouno t 0.005. 5% After burnup, it is about 0.013 %. U__ mainly is a problem in fuel 234 manufacturing, as it contributes about 80 % of the uranium a activity.

a) U236 (T 1/2 U is mainly created in the reactor through neutron absorption of 236 U . Fig. 3 shows the U concentration as function of burnup 235 236 and residual U,-,- contents causet .I o problemsn s during processing steps, but is a neutron absorber and thus must be compensated for by a highe^ U r enrichment if the material is to be recycled. For every % U in the recovered uranium, the U enrichment must be 236 235 0.15 % higher than for equivalent natural uranium fuel.

0.5 H

Parameter: Burnup (MWd/kp.)

0.3-

0.2-

0.1-

Initial enrichnent

0 1 2, 3 3.2 Residual enrichmentj /o t to :, 5 . U-,,/23 (V U Fie. 3: ._, content a functio s a s f residuao n U l enrichment for PWR fuel • — 2 236

261 . Specification3 d suggestionan s r theifo s r modificatio r furthefo d ran n purificatio f reprocesseno d uranium.

Problems concerning specifications of recovered LVR uranium have been a matte f concero r r severafo n l years n somI . e cases o stringenthes e "ar s a t to make U recycling nearly impossible in future (e. g. U ), in other cases they are too "generous" and should thus be lowered (e. g. a activity, Tc ). n generalI e currenth , t standar f modero d R reprocessinLV n g plant s suchi s , that further purification of the product uranium could only be achieved at very high costs s thui st . I importan e whethese o t tr such further purification cannot - where required - be carried out more economically at other stages of the fuel cycle.

1 Element3. d isotopesan s forming volatile fluorides

These elements are carried on into the VFf. The problem will thus be D wit e enrichee fueth hth l d manufactureran r . Nuclides forming volatile fluorides are divided into two groups: a) Nuclides specified in relation to U : Cr, Mo, Va, W, U , U : 2 23 3 23 ^.o->

US-DOE US-DOE ASTM proposals UNH specs UF specs for UF specs 6 6

Cr 3000 ppm/U 1500 ppm/U chango n e Mo 500 200 « V 500 200 «

Va 500 200 N

500 W U233 500 0.11 " 0.11 U232 0.3 - 0.5 PPm/U235

Current experience with WAK reprocessing plant shows that for all nuclides (except ^), values in UNH can be kept to the values specified for whic, e roughlF U har y halvaluee th f s specifie UNHr fo d , o

s beeha concernin nt i propose , gU d withi no fro g AST mo t M0.1 1 3 ppm/ latel, 0. o ppm/Ue lattet Th y5 ppm/ U. 0. eve ro Ut n value would allow to recycle uranium with original enrichments of 4 % and burnup up to 50.000 MWd/tHM (Re. Eig. 2). This high concentration of U232 could cause problems due to decay products. However, sublimating , froe containeUF mon othere th r o feedino ,t r g int plana o t 6

262 (enrichment, fuel manufacturing), will segregate most of the decay products which form non volatile fluorides (Re. 2.2.3). They would then create a radwaste problem with the rinsing solution for the UF, cylinders. 6 ) Nuclideb s specifie n relatioi d o , totaSi)t P n , ;l (C uranium ; Br ; :Sb total halogenides; Nb, Ru, S, Ta, Ti. No problems have been encountered so far with these elements, and there o suggestion s i chango t n currene th e t specifications.

2 Element3. s formin n volatilno g , e Cm fluoride , Ca , Cd s , (AsBi , , Be A.1, Ba , Fe, K, Li, Mg, Mn, Na, Ni, Pb, Sn, Sr, Th, Zn, Zr). No problems are known with these elements except for Zr (Re. 2.2.1).

3.3 Boron equivalent (B, Cd, Co, Eu, Gd, Li, Sm). e totaTh l Boron equivalen ppm/8 f o tU present difficultieso n s .

4 Transurani3. cactivitya . e generallTh y admitte a activit d U f 1500an o TR d s H i 0yUN dpm/gn i U 1500 dpra/gU in UF (before enrichment), tot 6 However, enricher o high to US-DOd s ,an F finU value Er th dpropose fo e d 6 to go down to 25 dpm/gU in UF . This seeras unrealistic, as it is tot 6 practically on the analytical threshold. othee th rn O hand TRU-isotoper , fo conversio F D a s snha whic s largei h r than 10, and effective filters are known to reduce them further (for exampl wool)i eN . Discussion e goinsar n withio g nseemt i ASTM d s an ,tha t a value around 200 dpm/gU in UF, could be agreed upon. tOt D Reduction activit. o H woul f UN so e uneconomical n b di y s thea , y would make necessary very costly supplementary purification systems in the reprocessing plant. 3.5 ß and Y activities of fission products

US-DOspecsH UN E. US-DOE UF6 specs

T ß Y ß (yCi/kgU) (yCi/kgU) (yCi/kgU) (wCi/kgU)

a) if not less than 75 % of total f.p. 320 1360 o t activit e du s i y Ru isotopes 32 68 b)if less than 75 % f totao l f.p. 160 680 activite du s i y to Ru isotopes

263 e onlTh y problem encountered c T her (Res i e . 2.2.1),t whicno d ha h been specified d US-DOseparatelol e Eth specificationsn i y n 197I . 8a recommendation was issued by US-DOE to limit Tc in UF- (Before 6 enrichment) to 0.4 ppm/U. It seems that conversion can remove a large part of Tc, as the volatile TcF (which is carried through into the UF , Re 6r 6 fig. 1) is rather unstable and forms to an important extent non volatile compounds. Reprocessing could admit a limit of 8 ppm/u in UNK. Reduction of Tc in UF would relieve the enrichment plant. This problem is still D being discussed within ASTM.

2) 6 Specification3. r reenrichefo s F "J d It is of interest within the cycle to consider the specifications set up by the fuel manufacturer for reenriched UF concerning recovered LWR 6 uranium. These specifications are meant for uranium reenriched to a maximum of 5 % U . 3.6.1 Elements forming volatile fluorids: a) Nuclides specified in relation to U :

RBU specs

Cr 1000 Mo 1000 V 1000 Va 1000

probleo N expectes i m d with these value compares sa F U o t d 6 specifications before enrichment, as the absolute concentration of the impurities would remain unchanged in the worst case (that is, they would all be carried into the enriched product), whereas the U concentration will raise by a factor of 3 to 5 in average. In reality, at least part of the impurities will go into the tails stream.

Nuclide) b s specifie n relatioi d o totat n l uranium. n sumI , they shoul excèdt no d0 ppm/U 30 e .

No problems should arise here, as a great part of the impurities will go into the tails stream.

2) According to a communication from RBU.

264 3.6.2 Uranium isotopes

~U232 ng/s<° :2 J This means that uranium with a U content of about 0.5 ppm/U ^*3 £ before enrichment coul usede db . w ~U234: 0-13 /'° °^ tota^- uranium. This is based upon current experience and will cause no problems. 3.6.3 Kuclides formin volatiln no g e fluorides m thesu y f totan o shoulI m t excèdlpp no d uranium0 30 e . This causeo sn problems. 3.6.4 Transuranic isotopes The a activity of these should not excède 25 Bq/gU. This is the same as 1500 dpm/gU and is equivalent to current UF, specifications before 6 enrichment. 3.6.5 Fission products ß activity shall not excède 1000 Bq/gU (leading nuclides are: Tc ,

CS 137' This is equivalent to 27 yCi/kgU, or a factor of about 2.5 times less than curren F specificationU t s before enrichment n equivalen.A t 6 decontamination must thus occur during enrichment. 3.6.6 Th

This isotope is specified to 8.4 . 10 y e firsth ts deposites i i deca t d i soli a an y s A s produca dd U f o t fluoride mainly in the enrichment plant, this means that recovered uranium should not be stored for more than a few months after enrichment before o fuegoint ln go manufacturing .

. Experienc4 e with uranium recyclinG FR n i g

So far about 200 kg of uranium recovered from reprocessing have been reenriche d recyclean d FRGn i d . They were containe 8 a fuepin 10 f l so n i d assembly loaded into the reactor of the Obrigheim power station in 1983. No major problems occurred during the processing. It must however be stated that the uranium used for this project (it came from the reactor of Gundremmingen A power station, which was shut down in 1977) had only reached burnup abouf so t 18000 MVd/tU. Fuel manufacturin s alswa go uncritical, however handling of recovered uranium with higher contents of deca_ „ U y product y mak sma necessart ei shielo t y d process e partth f so . «O-J ^

265 It also was seen that part of the impurities (mainly U decay products £, O £, and possibly to some extent fission products) went into the AUC filtering solutions, whic y caushma a radwaste e proble somo t m e extent.

In order to demonstrate the recycling of uranium at an industrial scale, a programm s beeha e n initiate o recyclt d complet8 e e fuel assembliee th n i s reacto f Keckarwestheiro r a power plant o thiT . s effect, som 3 ton1 ef so uranium recovere y reprocessinb d g use dNeckarwestheire th fue f o l a plant a t La Hague were converted to UF at the COMURHEX plant in Pierrelatte and 6 reenriched av the URENCO plant in Alraelo. Currently 4 fuel assemblies are being manufactured by RBU in Hanau; they are scheduled to be loaded into the reacto n Juli r y this year. Abou 7 tont f uraniuso m wil e recovereb l d from the current reprocessing of Neckarwestheim fuel at the WAX reprocessing facility near Karlsruhe. They are scheduled to be converted by COWRHE n sprinXi f thio g s year, reenriche URENCy b d summen i O d autuman r n and then shipped to RBU for fuel manufacturing. A second batch of 4 fuel assemblies will be loaded into Neckarwestheim reactor in 1987.

The original fuel from which the uranium was recovered had reached burnups of about 30 to 33 GWd/tU, with cooling times of 2 to 3 years before reprocessing.

5. Conclusions

Uranium recovered from reprocessing of used LVR fuel can be recycled via reenrichment. However e impuritie, th som f eo containt i s s create problems, wilt i e necessar b lo s reduco t y e them. e standardTh s reache modery b d n reprocessing plants would mak furthea e r purification of UNH solutions so costly as to seriously compromise the economic advantages of recycling. However, further purification can be performed quite effectivel t relativela d an y w costlo y othet a s r stagef so the fuel cycle, chiefle conversioth t a y n stage. Transfe f producto r s (mainly UF,) froe containeon m anotheo t r r feedino r g into facilities 6 also achieves effective purificatio r certaifo n n isotopes.

These matters will have to be investigated carefully by countries wishing to choose the reprocessing and recycling alternative for their used fuel.

266 CONVERSION OF REPROCESSED URANIUM IN JAPAN

I. YASUDA, Y. MIYAMOTO, T. MOCHIJI Ningyo-Toge Works, Power Reacto Nuclead an r r Fuel Development Corporation, Tomata-gun, Okayama-ken, Japan

Abstract In 1981 Power Reactor and \; u clear Fuel Develo- pment Corporation (PNC) constructed the Conversion Test Pacility-n (CÏF-II n ordei ) o develot r e th p conversion proces f reprocesseo s d uraniuo obtait d an nm the s c a 1 e - u p data and technical know-hows which will be practically applied the plan of a pilot or a commercial conversion plant in Japan, e facilitTh y which convert e reprocesseth s 3 UQ d to U F s has the capacity to convert 4 tons of U per year. The U 0 3 produced by the fluidized-bed denitr- e reprocessinth ato n i r g plan n Tokai t i Works, P\C, is used as a starting material. The process of the CTF-n consist f mainlo s 3 0 hydration U y , dehydration d reductionan , hydrofluorinatio F fluorinatio d an n n sec onsti . 2 The main items researched by the facility are as follows: (1) Investigating the behaviors of the fission products such as Ru and Zr and the transuranic elementp containeN d an su P dsuc s vera h y small amounte reprocesseth n i s d uranium d separaan , - ting those elements from uranium. (2) Improving the chemical reactivity of reprocessed uranium prepared by the fluidized-bed denitrator. e facilitTh s converteha y d approximatel 5 tony s ; beside0 e followingth s s have become clear: (1) The chemical reactivities in reduction, hydrofluorination a fluorinatioF d an , e improvear n d by hydration of U 0 3. (2) Both amounts of F. P and T R U containing in product Ü F s are less than these in the specification of e uraniuth m enrichment pilot plan n PNCi t .

1 . INTRODUCTION The establishment of the nuclear fuel cycle using reprocessed uranium and purified plutonium can e efficientlus y uranium resource d considerablan s y e decreaseb e quantitth d f yelloo y w cake imported into a pressin s Japan n Japai ha t i .ng nee o develot d p d accomplisan e recyclth h e utlizatio f reprocesseo n d uranium into LSR, because many L W R s are smoothly opérât ing in Japan.

267 The reprocessed uranium would be ullized in the neaR fueLW r le futurafteth s a re fabrication following conversio d enrichment R an nfue AT le th d als an s ,a o mixed with plutoni urn. s developini C PN o utlizt g e bote fuelth h s described above at CTF- n in the Ningyo-Toge Works and the Tokai Korks. The main researchs being carried out at CTF- U are as follows: ) (1 Improvemen e chemicath f o t l reactivit f UOoo y . The low reaction rates, particurally a lower hydrof l uor i nat i on rate, are given as the main characteristics of the U 0 3 produced by the fluidized bed denitrator. Therefore, we need to improve the chemical reactivity of reprocessed UOs. (2) Investigations of the behaviors and the elimination of the radioactive impurities. The reprocessed uranium is purified enough s radioactivitit d an s nearli y e samth s ya e nutural uranium. However e reprocesseth , d uranium contains the very small quantities f radioactivo e impulitiesd an P ,. F suc s a h TRÜ. It is known that these impurities accompany with uranium or remain in the facilities in conversion process. The confirmin e behaviorth ge impuritieth f o s s and the eliminating the impurities from the reprocesse e verar ys importanUO d o estat t - blish the U F s conversion process and to operate the following uranium enrichment f ac i l 111 es. In 1981e CTF- th s constracte, wa n n ordei d r to solve those problems, to obtain a scale-up data d technicaan l know-hows e manth yd datan , a regarding those described above are being obtained at present. The data will considerably contribute to the plan of a pilot or a commercial conversion plant in Japan.

. OUTLIN2 E CTF-ÏTH F O ÏE

e CTF Ï Th convertÏ - e reprocesseth s o whicUO d h is produced by the fl u i d ized-bed denitrator in the reprocessing plant in Tokai Works, PNC. The CTF-II has the capacity to convert 10 mois U/hr(4 tons U/y) s converteanha d d approximatel 5 tony U sincs e th e hot running e y processprocese CTF-Th dr th s .i f no s , which consists of mainly hydration, dehydration and reduction, hydrofluorination and F 2 fluorination sections. Figure 1 shows the flow sheet of tne conversion process, and in the outline of the conversion process every section as follows will be descrived.

268 Hydroduorlnauon

—*• To -it* ih scrubbcf

i Muormoitoo t " chemical trap NoF chomical A'jO? r~J—• Second colp ,d»o ra p chomical trop

Fust cold tiap r nlXjIH KO i sauLiLxir j (luormaJor

Uranturn enrichment pîanl

! Ash recatvor

UFjCyhnd«

FLOW SHEET OF REPROCESSED URANIUM CONVERSION PROCESS

F I G. 1

(1) Grinding and classification s chargei e fees Th bu dd inte feeth o d tank fro a I'Om a containe a pneumati y b r c conveyor. Then s chargei 3 U0 d e e inte nexth grindeth th o t d an r vibration sieve froe feeth m d tank throug e screth h w feeder. The produced UCU is put into a product hopper whic s alsi h o a feeuse s a dd e hoppeth f o r hydration process. The powder sizes of the b03 is approximately 10Q,wm to 200/^m. On the other hand, e over-sizeth d powde s returnei e rfee th do t tand k from the recycle hopper by a pneumatic conveyor. (2) Hydration The feed UOs powder is charged into the hydrator froe feeth m d hopper throug a rotarh e y th valve s A . I)0s stirrei 3 d with watea constan t a r t feed rate, a changeUO e s hydrateth s UQ e producinte Th th o. t 3 hydratt U0 inte producpu th os i e t hopper whics i h also used as a feed hopper of the dehydration and reduction process e reactioTh . n equatios a s i n fo l lows: t a room temperature U0 nH20 L'03-nH

269 (3) Dehydration and Reduction

The feed U 0 3 • n H2 0 is charged into the fluidized-bed reductor from a feed hopper. The temperature in the reductor is maintained at approximately 550 °C with external electric hearters. The fiuidizing and reacting gas is cracked ammonia, a mixture of Ha and N 2. The product U02 is taken out of the reductor through a screw feeder and put into a product hopper whichis also used as a feed hopper of the hydrofiuo- rination process. This reductor, 2.2m in length with approximately 10c n mdiameteri s constructei , f o typd e 316 stainless steel. The off-gas from the reductor, Ha-HaO-Na gaseous mixture s burt i wit,ou n h excess air by the Ha gas burner. The equation of dehydration d reductioan s followsa s i n :

UÖ3-nH20(S) + H2 (g) :illl4 U0a (s) f (n-1) H20(g) n o i t a n i r o u l f o r d y H ) (4

The product U02 from the redactor is charged into the Muidized-bed hydrofluorinator which is operate t aboua d t 400°C with external electric hearters. The U 0 a is fluidized with H F diluted with t intpu a oe d producs Th an takei t « F ou . nU N2 t product hopper which is also used as a feed hopper of the F2 fluorination process. The reactor, 2.0m in length with approximately 8cm in diameter, is constructed of Monel alloy. The off-gas from the hydrofluorinator, HF-H2Q-N2 gaseous mixture s chillei , d to about 0°C to condence H F gas and is drawn through the alkali scrubber. The equation of hydro-f 1 uorin- s atiofollowsa s i n : : U02(S) 4- 4HF(g) ^^^ UF<(S 2H+ )2 Q(g)

(5) F2 Fluorination e producTh < froe F hydrofluorinatoU th tm s i r charged into the f l u i dized-bed fluorinator. Sintered A 1 2 0 3 which does not react with fluorine is charged inte d fluorinatomaterialth e be o th e th d an s ,a r charged U F < is mixed and diluted with this A1 a0a. y reactiob s s convertega i « e n U F UF wit o t hd fluorine s dilutega d wit 2 2 gas N F he equatio e Th .th f o n fluorination is as follows:

UF<(S) + F2(g) > UF6(g)

e chemicaTh ld CoFe an tra ar n whic22 i F pg M h packed is set subsequently to the fluorinator to remove a small amoun f radioactivo t e impuritieP s. F suc s a h and TRU. The fluorinator, 2.0m in length with approximately 8cm in diameter, is constructed of Nikle metal.

270 Afte rs coole i reaction d s trappean ga d F l d, o coltw in d trap n scries i e s firsTh .t cold6 tras i p cooled at about 0 °C and the second at about -30°C. f gaseof e s Th froe colth m d traps contai a smaln l quantity of UFs and F2. Those gases, therefore, are treated by the N a P chemical trap, the Al20j chemical trap, and the aikdli sciubber to remove UF6 gas, F2 a e smal residuan th i ld s gasan amountga ,l , respectively. The solid U F s trapped in the cold traps is warme o dliqueft wito t abouwdteC t ° ho 0 hp 8 u ytr B t cylinderint12 e pu liquefie th a s oi d G an F .! I d Some non-volatile fluorides remain in the fluidized-bed togethe d removean r3 wit20 1 d A h into h receiveas e th t regulaa r r intervals.

3 . THE RESULTS OF EXPERIMENTS

1 3. Improvin e chemicath g l reactivities ) (1 Hydrn io at It occured at times in the operation that the fine powder in the hydrator aggregates to grow up to the large particle s sama ss severaa e l centi meters in diameter o solvT e problem.th e e operatinth , g factors, water feed speed, the ratio of U03 to water and stirring speed, were changed, and the optimal condition were determined. The hydrator is operating smoothly unde e optimath r l conditio t presenta n . It was identified by means of x-ray diffraction that the UOs hydrate produced at room temperature had the crystal forms of U03-2H20 and/or UQ3-0.8H20. a 3 surfacwhic Q d U ha Whee h th en are f approximatelo a y 0. 5rrf/g was hydrated, the surface area of the UCU hydrate increase o approximatelt d y 3mVg greater than e U03th tha f .o t Figur 2 showe e increase th s th f o e surface area caused by the change from UOa to UCU hydrate. (2) Dehydration and Reduction The fl u i d i zed-bed reductor was operated smoothly for a long period. The production rate was approximatel 5kg-li/h. 2 e ycontinuou th n i r s operatione Th . reaction reactivit s almoswa yU 0/ t e perfecth d an t rati f produco o t 1J0 s abou2wa t 2.05 e hydrogeTh . n s utilizatioga s approximatela n w efficienclo s a s wa yy 45%, becaus e amounth e f hydrogeo t s suppliega n s wa d twice of the stoichiometry. The experimental conditions in the f l u i d i zed-bed reductor, the fluidizing properties and powder properties are summarized in Tabl. I e

271 bo X "6

2 3 ) r h ( E M I T N 0 I T A R D Y II

FIG.2 TIME DEPENDENC F SURFACO E E AREA

TABLE I . EXPERIMENTAL CONDITIONS

REACTION FLUIDIZÎNG INLES GA T IIOLD-UP POWDER PROPERTIES TIME IN \ TEMP VELOCITY VOLUME REACTOR SURFACE AVERAGE BULK AREA POWDER SIZE DENSITY -•V (t) (en/sec) (Vol/o) (hr) (rn'/g) (Aim) (g/cc) DEHYDRATION 2.90 200. 8 2.52 550 15.0 45 (H,) 4 & REDUCTION u03nH20 4.49 180.0 3.51 HYDROFLUOR1NATION 400 23.0 50 (HF) 3 U02 0.98 197.4 3.01 F FLÜOR1 NATION 420 10.0 30 (F ) 0.3 2 2 UF< i

e kinetiTh c curves measure e relatioth d n between the reduction time and the reduction ratios of the 3 hydrateU0 e th y differentiab d s an s UO l thermal analysis (DTA) are shown in Fig.3. As shown in Fig.3, the reaction ratios of the U03 hydrates are higher than that of the U03. (3) Hydrofluorination e fluidized-beTh d hydrofluorinato s beeha r n operated with no troubles. The production rate was approximately 2.3kg-U/hr in the continuous treatment. e hydrofluorinatTh n reactivitio s higa s appr a s h wa y- oximately 95%. The hydrofluorine gas utilization efficiency was approximately 90%. The experimental conditions are summarized in Table I .

272 100

70 CONDITION 60 TEMPERATURE 550t

INI ET H2 VOL 30% 50

40 • PRE-HYORATÎO 4m'0 ( ) V g / * POST-HYDRATIO Om3 ( N 1) /g 30 APOST-HYDRATIOV (3 5m'/ g)

10 20 30 40 50 60 70 REACTION riME(min)

FIG 3 TIKE DEPENDENCE OF REDUCTION RATIO 3Y MEANS 0^ D T A

e kinetiTh c curves measure e relatioth d n between the reaction tine and the reaction ratios by DTA are shown in Fig.4. The reaction of the UQ2 derived from 3 hydrate0 Ü e s th rapii sd completan d e compared with 2 prepare0 e U direc th e y th b d t f o reductio t ? th f o n f l u i d i zed-bed lids. 2 fluorinatioF ) (4 n The p roduciion rat e of the F 2 fluorination F U usin e th g 4 which was p roduced with over 90% of the hydrofluorination ra s tapproximatel1wa 0 . Ikg-u/hr2 y . e caseth In , the fluorine utilization efficiency was a s higa s a h pprox imate 1 y 99% at 420°C. However, the producti on rate of the F2 fluorination using the UF« whica w h s producei w d e hydrofth f to h - unde % 90 r luorinatiod n un rati s wa o er 80%. Tne experimental conditiona s e r sunimad . ze ri I e l b Ta n i The r eact ion rate and fluorination utilization efficiench t f o C a ° y 0 37 t 4 produceF U e y PNC-procesb d s are higher t han those of the UF4 produced by CTF-ÏÏ, at 420 °C. This facy ma t mainly be due to the smaller f quano y t ti UQ2 in the UF 4 produce y PN'C-procesb d s than that in U F 4 produced by CTF-n .

273 100

CONDITION 60 TEMPERATURE 400t % 30 L INLEVO F H T 50

• PRE-IIYDRATIO 4m'0 ) ( /Ng 40 * POST-!l') g : DRAV Um T IO3 ( N POST-HYDRATIOA ) g / 1 5m 3 ( N 30

10 20 30 4Û 50 60 70 REACTION T[WE (-'n)

FiG 4 TIME DEPENDENCE OF HYOROHUOR ! KA^I (LN P4TIO BY MEANS OF D T A

e kinetiTh c curves measuree r e th d at ion between e reactioth e reactionth timd an e n ratios e ar A T bD y shown in Fig. 5.

Main results of the above experiments, dehydr- atio d reductionan n , hydrofluorinatio 2 P fluori d an n - nation, are shown in Table U.

3.2 investigat e behaviorth eg n d an s ; n 1 1 a n i m i l e e th of the rad oactive impurities. A small amount of radioactive impurities containeU R n reprocesseT i d d an P . 3 F sucDG ds a h behave with uranium to the b F 4 form in spite o ( the changes of the uranium forms in dry process. Most of these impurities ctiange to these fluorides in the f l u i d i zed-bed fluorinator. Some of these fluorides havin e loweth g r boiling point temperature than that 2 fluorinatioF f o n temperature would behave . wit6 P ü h However n practicai , l operation e wholth , e quantity f theso e fluoride a fluidized-be n i s d fluorinator, such as R u F s, SbFs, N p F s and PuFs, does not always behave with UFG. It is inferred that these fluorides

274 100 -

ÜF, PRODUCED IN CTF-ÏÏ UF, PRODUCED BY PNC-PROCES:

e— •<

ce. CD CONDITION TEMPERATURE 400"C INLET Fj VOL 30%

10 20 30 40 50 60 70 REACTION TÏME(nin)

FIG,5 TIME DEPENDENCE OF F2FLUORINATION RATIO Bï MEANS OF D. T. A

TABLE. Ü RESULTS OF CONTINUOUS TREATMENT

S UTILIZATIOGA N "\ CHEMICAL REACTIVITY OPERATING RATE EFFICIENCY ^\ (96) _ (kg-u/hr) (%) DEHYDRATION 100 2.5 44 (Hj) & REDUCTION

HYDRGFLUORINATIDN 95 2.3 90 (HF)

2 FLUORINAT1DF N 100 2.1 99 (F2)

275 would rema in in a fluor nator as e a b soli o t de du reducte o thest d e lower va 1 ency states, The behav or of PuF6, as an example, s ifoia s l ows:

PuFs(g) UR«(S) PuF<(S) UF6(g) The fluorides gas released from the fluorinator is trappey b d the subsequent chemical trap packed with d CoFan 2 2 .F Mg n volatilno e Th e fluorides, suc, SrFas h z CeF<, AmF d CniFoan 3 , zed-bei almos d i u l f dt e remaith n i n fluorinator s removei d an ,d togethed rbe wite th h material, AlaOa a wast s t a regulaa ,e r intervals. e havW e investigated mainl e behaviorth y f o s Pu and Np in p2 fluorination. The results are as follows: (1) In spite of the /er y small quantity of Pu s sufficientli 3 inÜ0 UOon i u ,P y separated from UFs by the f l u id ized-bed fluorinator e chemicath d an l e amountrapu Th P f .o t containins undei 0 s dpm/g-U5 r UF n i g . e valuTh s lesi e s thae limiteth n d valuf o e e lowesth t detectio f alpho n a counting analysis. (2) The behavior of Np is somewhat uncertain, because of the large error of analysis. However, in spite of the large error, it would be certain that the removal ratio of e chemicath y b Np l a tralittl s i p e lovr ,e literature value tha. th n ] th ni e2 [ e U containin) TR (3 Becaus d an e amountP th . eF , f o s in UOs are very small quantities, these behavior e uncertainar s . These contentn i s e lesar 6 s U F than these e valuespecth n i -s ificatio e uraniuth f o n m enrichment pilot plant in PNC. The chemical data of repre- sentativ ee specificatiosamplth d an e e ar n shown in Table DI .

TABLE I ANALYSIS RESULTS AND SPECIFICATIONS OF F. P AND TRU

ITEM ANALYSIS RESULTS SPECIFICATIONS

< 105 Z9 r- 3 The total activty Nb 95 < lu'3 e maximuth f o m gamma Rul03 < 1Q-3 rays emitted from F. P Rul0 < 6 IO-2 95Zr, 3SNb, ""Ru, Csl37 < IO-3 '"Ru, 131d Csan , Cel44 < 1Ö'3 "4Ce is TOTAL < 0 OlSjuCi/g-U 05j"Ci/g-0 U Np < 150 e totaTh l activity Pu < 50 e maximuth f o m alpha TRÜ Am < 100 rays emitted from 0 5 < Cm s i m C d an , Am Np, ,Pu TOTA < 350dpm/g-ll j 1500dpm/g-U

276 4 . CONCLUSION The fundamental data and technical know-hows e beinar g obtaine e CTF-Ïth y Ïb d operatione th d an , initial object is being nearly reached. The facility would be scaled up in the near future, and after the operation of the expanded facility durin a certaig n period a e pla th f ,o n commercial conversion plant woul e startedb d .

REFERENCES

[1] NAKAI.T., SAITO.N., 1 SH 1 MOR !, T. , "Plutonium", Complete books of Inorganic Chemistry, XVÏÏ-2 (1967)195. . GolliherR . W ] , [2 etal, U.S.Patent , 3 615, , 267(1971) PANEL DISCUSSION

(Summarized by H. Page, U.K.)

In his opening statement to the Technical Committee Meeting on Advances in Uranium Refining and Conversion, Mr. J.L. Zhu, Director of the Division of Nuclear Fuel Cycle, remarked that the contents of the papers schedule r presentatiofo d n indicated that, ther d beeha e n considerabl ee nuclea innovatioth f o e fronr th d fue en t a nl n cycla n i e increasing numbe f Membeo r r States sinc previoue th e s conferenc Parin i e s in June 1979, despite the recent slowing down of nuclear power plant e conferencth f d becomo constructionha d t i en e e quitth ey B .eviden t l participantal o t s that this indeewa s e caseth d .

While it can be seen that the major uranium supply/processing countries continue to refine and modify their traditional conversion processe responsn i s o changet e marken i s t demand, feed producan d t specification, environmental constraints, regulatory practices etc., several other interesting developments have recently emerged.

Self sufficiency in fuel production is evidently one of the recent trende nucleath n i s r fuel cycle policn increasina f o y g numbe Membef o r r States. This accounte launchin th man e w laboratorr th ne y fo s f o g d an y pilot scale projects described at the conference and explains the extensive research and development effort aimed at production of uranium fuel cycle intermediates, such as fluorides, metal and oxides, suitable r processinfo o nucleat g r fue Membey b l r State r theifo s r home based nuclear reactors growina n I . g numbe f caseo r s Member State movine ar s g into nuclear fuel cycles which do not involve uranium enrichment.

An important issue, not only to the front end of the nuclear fuel e cyclwholea th s a o et cycl , t concernbu e e uraniu th e fat th d f so e an m plutonium recovered from the reprocessing of irradiated fuel. Member States without indigenous uranium and concerned about the economics of future" fuel cycles based on "natural" uranium are showing a great deal of interes recyclinn i t g reprocessed irradiated uranium (REPU) back inte th o fuel cycle either in routes involving enrichment or as mixed uranium/plutonium oxide fuels (MOX).

279 A further development with potentia r significanfo l t impacn o t traditional uranium conversion technology by the middle of the next decade is atomic vapour laser isotope separation (AVLIS) and existing converters accustome o providint d g uranium hexafluorid e feeo th t d s a e enrichment facilities will need to keep in close touch with the progress of this technology which requires a feed of uranium metal.

However, whil e technicath e l option r uraniufo s m refinind an g conversion differ from countr o countrt y y dependin n criterio g a sucs a h for example:

e type th f nucleao s) i r reactor systems involved; ii) the availability of natural uranium or other nuclear fuel resources

there remain, nevertheless, certain common features of the refining and conversion processes whic e relevan ar he industr th whola o t d s a yan e which merit inclusio e agendath n i f nfuturo s e meeting e Technicath f o s l Committee.

In addition to future developments in the principal conversion process technologies, examples of issues of common interest identified as appropriat r futurfo e e consideraton include:

1. Technology for refinery raffinate treatment and reagent recycle aime t procesa d s econom d reductioan y f environmentao n l impact;

2. The influence of factors such as equipment design and operating parameter n scale-uo s p from laborator o pilot y t plan o industriat l scale; e productionTh . 3 , characterisatio d qualitan n y contro f nucleao l r fuel precursors suc s uraniua h m dioxide n relatio i ,e subsequenth o t n t performance of the nuclear fuel;

. 4 Provisio f opportunito n r coordinatefo y d discussion, involving both regulator d operatoran smethodologe th n o e sauthorisatio th f o y d an n f control of radioactive and toxic effluent from uranium refining and conversion facilities;

280 5. The recycling of uranium recovered from the reprocessing of irradiated oxide fuels.

Alread numbea y f utilitieo r s have contracte e reprocessinth r fo d g of oxide fuel d significanan s t quantitie f recovereo s d uranium will become available for reuse at the beginning of the 1990's. Factors influencing decisions regarding recycling or storage of this material of interest to existing converters will include:

i. The evaluation of existing experience involving reprocessed uranium in:

a. conversion to uranium hexafluoride (UF ) o

b. enrichmen. UF f to 6

c. reconversion of UF, and fuel fabrication 6

d. irradiation of fuel containing recycled uranium

e predictioiiTh . e totath f lo n quantitie d specificatioan f o s r fo n the uranium which will arise from reprocessing operationse Th . levels of transuranic elements, fission products, 232U and its daughters will be important in this context.

e definitioiiiTh . f feed o producn an d t specificatio r eacfo n h stage in the sequence, reprocessed uranium to finished fuel.

iv. An assessment of the consequences of the foregoing on existing plants, processe d practicesan s .

n vevaluatioA . e comparativth f o n e cost f fueo s l prepared from natural and recycled uranium.

n summaryI , ther s generaewa l agreement amongs l participantal t s thae Technicath t l Committee Meetin d bee ha gvaluabla n e forur fo m cooperatio d informatioan n n exchange betwee Membee th n r e Stateth n o s rapidly changing technologies at the front end of the nuclear fuel cycle.

28! However contrasn i ,e historicath o t l situation wher majoe th e r uranium converters interfaced only with uranium suppliers and uranium enrichers the emergence of reprocessed uranium as a new factor in the fuel cycl d highlighteha e forua e neer th dmfo d involving reprocessor, convertor, enricher, fuel fabricato reactod an r r operato givo t r e effective consideration to the features of the closed fuel cycle which w havno e potentia r impactinfo l n eaco g h participant e existinTh . g Technical Committee on uranium refining and conversion was not an appropriate foru r thimfo s Divisioe purposth d f Nucleasa eo n r Fuel Cycle s invitewa o considet d r establishing suc foruma h .

The 1986 Technical Committee fleeting closed with a recommendation that the rapid rate of innovation in the fields of refining, conversion and nuclear fuel intermediate production technology would justife th y convening of the next meeting in 3 years time in the Autumn of 1989.

282 LIS PARTICIPANTF TO S

ALGERIA

Boualia, A. Commissaria x Energieau t s Nouvelles 2, Boulevard Frantz Fanon Alger, Algeria ARGENTINA Vercellone, Jos. eA Jose Rogue Fune<- 1593 Cerro de las Rosas Cordoba 5,009 argentin».

AUSTRALIA

Bull, P.S. Counsellor, Atomic Energy The Embassy in Austria Mattiellistrasse 2-4/III A-1G40 Vienna BELGIUM

De Regge, P.P.M.H. Studiecentrum voor Kernenergia Boeretang0 ,20 B-2400 Mol, Belgium

BRAZIL

Abrao. ,A ïnstit'it Pesquisae d o s Energéticas e Nucleares Caiza Postal 11.049-Pinheiros Cidade Universitaria 05508 - Sao Paulo - SP, Brazil

CANADA Ashbrook, A„W. Eldorado Resources Ltd. 400-255 Albert St. Ottawa, Ontario, KIP 6A9 Canada

Didyk, J.P. Atomic Energy Control Boarf o d Canada P.O. Box 1046 Ottawa K1P5S9 Canada THE PEOPLE'S REPUBLIC OF CHINA Chao, Lebao Burea Nucleaf o u r Fuels 2102-1x P.OBo . 0 Beijing People'e Th s Republi Chinf o c a Zhu, Chang-En Beijing Research Institute of Uraniu Processine mOr g P.O. Box 234 Beijing, e People'Th s Republi Chinf o c a

283 ARAB REPUBLIC OF EGYPT

El-Hazek, N.M.T. Nuclear Materials Corporation l MaadE i Kattamiya Road Maadi Post Office Box 530 Cairo, Arab Republic of Egypt

FRANCE

Rigo. ,L Compagnie Général Matières de e s Nucléaires (COGEMA) Service des Usines et de la Compta- bilit Matières de é e Bassd e 2, Rue Paul Dautier, B.P. No. 4 78141 Velizy Villacoublay, France

Faron. ,R COMURHEX Tour Manhattan CEDEX 21-92087 Pari a DéfensesL , France

FEDERAL REPUBLI GERMANF CO Y

Becker. ,B Reaktor-Brennelement Union GmbH Postfach 11 00 60 D-6450 Hanau 11 Germany, F.R. Leyser. ,E Deutsche Gesellschaft für Wiederaufarbeitung von Kernbrennstoffen mbH (DWK) Hamburger Allee4 D-3000 Hannovel r Germany, F.R. Wehner, E.L. NUKEM GmbH Rodenbacher Chaussee 6 0 Postfac8 O O 1 1 h D-6450 Hanau 11 Federal Republi Germanf o c y Sondermann, T. Reaktor Brennelement Union GmbH Postfach 110060 D-6450 Hana1 1 u Federal Republi Germanf o c y

INDIA

Kansal, V.R. BHABHA Atomic Research Center Trombay, Bombay - 40O085 India

ISLAMIC REPUBLI IRAF CO N

Gharib, A.G. Atomic Energy Organizatio f Irano n Fuel Department of AEOI P.O. Box 11365-8486 Teheran, Iran

284 Kamali, J. Atomic Energy Organization of Iran Esfahan Nuclear Technology Center P.O. Box 81465/1589 Esfahan, Iran P.O. Box 11365-8486 Teheran, Iran

IRAQ

Abdul Fatah, A.M.A. Departmen f Chemicato l Engineering College of Engineering Baghdad University Jadriya, Baghdad, Iraq

JAPAN

Mochiji, T. e ProcessinMillinOr d an g g Division Ningyo-Toge Works Power Reactor and Nuclear Fuel Development Corporation Kamisaibara-mura Tomata-gun Okayama-ken 708-06 Japan

KOREA. Republif o c

Chang, I.S. Korea Advanced Energy Research Institute P.O. Box, Daeduk-Danji Choong-Nam Republi Koref o c a 300-31

PAKISTAN

Shabbir. ,M Pakistan Atomic Energy Commission (PAEC) P.O. Box 1114 Islamabad, Pakistan

SOUTH AFRICA

Jackson, A.G.M. Atomic Energy Corporation of South Africa Limited P.O. Box 4587 Pretoria 0001 South Africa

Ponelis, A.A. Atomic Energy Corporation of South Africa Limited P.O. Box 4587 Pretoria OO01 South Africa

Roux, A.J.A. Atomic Energy Corporatiof o n South Africa Limited 458x Bo P.O7. Pretoria OO01 South Africa

285 SOUTH AFRICA (cont.)

Scholtz, T.E. Atomic Energy Corporation of South Africa, Limited 458x P.OBo .7 Pretoria 0001 South Africa

Tiltmann, D.E. Atomic Energy Corporation of South Africa Limited Private Bag X256 Pretoria 0001 South Africa

Venter, C.J.H. Atomic Energy Corporatiof o n South Africa Limited P.O. Box 4587 Pretoria 0001 South Africa

SWEDEN

Lindholm, H.I. Swedish Nuclear Fuel and Waste Management Co. Box 5864 S-102 48 Stockholm, Sweden

TURKEY

Ipekoglu, B. Cefcmece Nuclear Research and Training Center Havaalan1 . No x P.OiBo . Istambul, Turkey

U.K.

Bleasdale, P.A. Ministr Defencf o y e MR (NUC)1 Building 10 C35 Mearinge Th s Burghfield NR. Reading RG3 3RP, U.K.

Labaton, V.Y. British Nuclear Fuels pic Fleming House Risley, Warrington Cheshire, WA3 6AS, U.K.

Page, H. BNFL (British Nuclear Fuels pic) Springfields Works Salwick, Preston Lancashire PR4 OKJ, U.K.

Todd, R. Nuclear Installation Inspectorate Peter'. St s House Balliol Road Bootle, Merseyside, L20 3LZ U.K.

286 Webster . K . ,R Authority Fuel Processing Directorate (UKAEA) B 10, AERE HARWELL OXON 0X11 ORA, U.K.

YUGOSLAVIA Tolic, A. Institute for Nuclear Technology and Other Mineral Raw Materials Beograd, Yugoslavia

ORGANIZATION

INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA)

Ajuria . ,S Divisio Nucleaf o n r Fuel Cycle Rojas, J.L. Division of Nuclear Fuel Cycle Ugajin, M. (Scientific Secretary) Division of Nuclear Fuel Cycle

287