IAEA-TECDOC-693

Use of fast reactors for actiniae transmutation

Proceedings Specialistsa of Meeting held in Obninsk, Russian Federation, 22-24 September 1992

INTERNATIONAL ATOMIC ENERGY AGENCY Duz/—A\

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Printed by the IAEA in Austria March 1993 FOREWORD

n importanA t feature activitieth e Internationaf th o e f o s l Atomic Energy o assist Agenc s i ty Membe re safth Statee n i handlins d an g managemen f nucleao t r powe d spenan r t nuclear fued radioactivan l e wastes that arise from the use of nuclear energy.

e managemenTh y issuef ke radioactivo t n e i sth f o ee on wast s i e today's political and public discussions on nuclear energy, especially the long term disposa f higo l h level radioactive wastes. Rather than waiting for their radioactive s decayprincipalli t i , y possibl o reduct e e th e period of toxicity of the actinides and long lived fission products through transmutatio f thesno e isotope fission si n reactor acceleratorsr so .

The recycling of plutonium in liquid metal fast breeder reactors (LMFBRs) would allow 'burning' of the associated extremely long life transuranic waste, particularly actinides, thus reducing the required isolation timr higfo eh level waste from ten f thousando s f yearo s o t s hundreds of years for fission products only. This additional important mission for the LMFBR is gaining worldwide interest. In the past years, an increasing number of studies have been carried out on the advanced waste management strategy (i.e. actinide separation and elimination n i variou) n sa internationa t countriea d an s l level. Recognizing this, the International Working Group on Fast Reactors (IWGFR) decided in 1991 to include the topic of actinide transmutation in liquid metal fast breeder reactor s programmit n i s f specialisto e s meetingst I . was apparent that due to the expansion of supporting research programmes and the fragmented nature of information arising from these various activities it would be beneficial to co-ordinate this topic at an international level.

Support for this approach was provided by Member States and, against this background e IAEth ,A organize e Specialistth d f Fasso Meetinte Us n o g Breeder Reactor Actinidr fo s e Transmutatio Obninskn i n , Russian Federation, from 22 to 24 September 1992.

e SpecialistTh s Meeting made clea e increasinth r g interese th n i t issue concerning long lived radioactive nuclides and toxicity aspects in nuclear waste management and the desirability to work out a co-ordinated approac s solutionit o t h .

e potentiaTh f o fasl t reactor r reducinfo s e lonth gg term radiotoxicit f speno y t fuel justifie n activa s e programm f conceivino e d an g developing technologie r fo recoverins g transuranic r fo in—cors e transmutation. EDITORIAL NOTE

In preparing this material press,the Internationalthe for staffof Atomic Energy Agency have mounted and paginated the original manuscripts as submitted by the authors and given some attention to the presentation. The views expressed papers,the statementsin the general the made and style adoptedthe are responsibility namedthe of authors. necessarilyviewsnot The do reflect governmentsthosethe of of the Member States organizationsor under whose auspices manuscriptsthe 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

Summary of the Specialists Meeting ...... 7

BACKGROUND PAPER

Status of transmutation ...... 13 L. Koch

NATIONAL PROGRAMMES ON TRANSMUTATION OF MINOR ACTINIDES AND LONG LIVED FISSION PRODUCTS IN FBRs (Session 1)

United States national program on actiniae recycle (Summary) ...... 21 KhalilH. Long lived wastes transmutation studie Francn i s e ...... 2 2 . M. Salvatores, C. Prunier, P. Bergeonneau, A. Zaetta, H. Sztark, G. Vambenepe, J. Vergnes Minor actinides recycling in an EFR type fast neutron reactor ...... 24 H. Sztark, G. Vambenepe, J. Vergnes, A. Zaetta Partitionin transmutatiod gan n researc developmend han t program (OMEGA) in Japan ...... 0 3 . T. Mukaiyama Statu f minoo s r actinide transmutation stud t CRIEPya I ...... 7 3 . Sasahara,A. KurataM. Statu f woro s transmutation ko Switzerlann ni d (Summary) ...... 2 4 . G. Ledergerber, P. Wydler, R.W. Stratton Scientific research program on actinide transmutation by use of fast reactors ...... 43 L.A. Kochetkov, A.G. Zcykunov

PHYSICS ASPECTS OF TRANSMUTATION OF MINOR ACTINIDES AND LONG LIVED FISSION PRODUCTS IN FBRs (Session 2)

Reactor aspect f electronucleaso r transmutatio f actinideno heave th n yi s water high flux blankets. Analysis of problems ...... 49 E.V. Gai, A.V. Ignatyuk, N.S. Rabotnov, A.A. Seregin, Yu.N. Shubin Some physics aspects of minor actinide recycling in fast reactors ...... 55 U.K. Wehmann Rol f faseo t reactor reduction i s f lonno g lived waste ...... 8 5 . A.G. Tsikunov, V.S. Kagramanyan, L.A Kochetkov, V.I. Matveyev Actinide transmutatio advancee th n ni d liquid metal reactor (ALMR) ...... 1 6 . C.L. Cockey Minor actinide containing fuels for transmutation purposes ...... 67 Koch,L. Coquerelle,M. RichterK. Physics considerations in the design of liquid metal reactors for transuranium element consumption ...... 70 H. Khalil, R. Hill, E. Fujita, D. Wade ENGINEERING ASPECT TRANSMUTATIOF SO N USING FAST REACTORS (Session 3)

Radiowaste transmutation in nuclear reactors ...... 77 A.N. Shmelev, G.G. Kulikov, V.A. Apse, V.B. Glebov, D.F. Tsurikov, A.G. Morozov Minor actiniae transmutatio fission ni n reactor fued an sl cycle considerations ...... 6 8 . T. Mukaiyama, H. Yoshida, T. Ogawa Homogeneous recyclin f minogo rtypR actinideeEF fasn a t n i sreacto r ...... 4 9 . H.W. Wiese, KriegB. Characteristic transmutatioU TR f o s LMFBn a n ni R ...... 9 9 . T. Wakabayashi, M. Yamaoka Experimental investigations concernin problee gth f minomo r actinide transmutation ...... 9 10 . S.M. Bednyakov, V.A. Doulin, Matveenko,LP. A.M. Tsibulya, A.V. Zvonarev A concep f specializeo t d fast reacto minor fo r r actinide burning ...... 4 11 . V.I. Matveyev, A.P. Ivanov, EfimenkoEM. The proposed fuel actinide cyclth f eo e burning fast reactor DOVITA (Summary) ...... 8 11 . A.V. Bychkov, A.A. Mayorishin, O.V. Skiba Reduction of minor actinides in nuclear waste via multiple recycling in fast reactors ...... 119 A.F. Renard, Pilate, S. Fuente, La A. JourJ. Vambenepe, net,G. VergnesJ.

List of Participants ...... 125 SUMMAR SPECIALISTE TH F YO S MEETING

The Specialists Meeting on Use of Fast Breeder Reactors for Actiniae Transmutation was held in Obninsk, Russian Federation, from 22 to 24 September 1992.

The meetin s sponsore wa ge Internationa th y b d l Atomic Energy Agency on the recommendation of the International Working Group on Fast Reactors e Instituts anhosteth wa d y b d f Physico e d Powean s r Engineering (IPPEn )i Obninsk. The meeting was attended by 23 participants from eight countries (Belgium, France, Germany, Japan, Russia, Switzerland, United Kingdom and USA)o Internationatw , l organizations (IAE 2 observer1 d CEC an Ad )an s from Russia.

The meeting was opened by Mr. L.A. Kochetkov, Deputy Director of IPPE, who welcomed the participants on behalf of the Ministry of the Russian Federatio n Atomio n ce IPPE Energth .d Referrinan y e subjecth o t gt of this Specialists Meeting, he emphasized the importance of the advanced waste management strategy (i.e. actinide separation and elimination). Mr. V. Arkhipov of the IAEA presented a review of IAEA activities in the field of minor actinide transmutation. The meeting was chaired by Mr. L.A. Kochetkov. Mr. L. Koch presented a background paper on the status of transmutation, prepared at the request of the IAEA. The paper discusses the objectives and the proposed schemes for partitioning and transmutation of long living radiotoxic nuclides particularn I . e constraint,th e th r fo s efficienc T procesP& a sf o yguidin a fas D worr tR& g fo k reacto r fuel cycle are described. Session 1; National programmes on transmutation of minor actinides and long lived fission products in FBRs Chairmen: L.A. Kochetkov Wehman. ,U n

This Session gave an overview of the national programmes followed in France, the USA, Japan and Switzerland. The programmes reflect the large effort whic s beini h g undertaken worldwide wite goa o th reducht l e th e quantities of the toxic wastes and the long term risk of geological repositories. In the technical presentations, the potential of fast reactor f standaro s X fueMO dl desig f specia o s wel a ns a l l minor actinide burner cores was illustrated.

Sessio Physic; 2 n s aspect f transmutatioo s f minoo n r actinide d lonan s g lived fission products in FBRs Chairmen: L.A. Kochetkov, L. Koch

e paperTh s presente n thii d s Session cove a varietr f topico y s related to transmutation of actinides and fission products, ranging from heavy water blankets for a LANL type accelerator-driven reactor to the adaptation of fast reactors to transmute plutonium and radiotoxic waste from LWRs together with self-generated waste. Here Europeas th a e R EF n well as the GE designed ALMR are considered and the possibilities of the Russian BN-80 discussede 0ar .

The advantage of high thermal fluxes in fissioning Am-241 and Np-237 by their neutron capture products, Am-24 d Np-238an 2 s pointei , d out. These nuclides witd neutrood hd protoan n n numbers possess very high fission cross-sections. In order to reduce the competing decay reaction, high thermal neutron fluxes in the area of 10^-^ n/s cm^ are needed, which cause secondary problem e reactoth n i sr itself, e.g e buildu.th f o p tritium. One possibility to circumvent these problems is to use a stationary liquid blanket.

Session 3; Engineering aspects of transmutation using fast reactors Chairmen: L.A. Kotchetkov Sztar. ,H k

Topics covered in this Session relate to transmutation of minor actinides and fission products, mainly with reactors devoted to this purpose, with specific core concepts and adapted fuels and fuel cycles.

Theoretical studies, mainly on the Russian side, review the physical parameters which gover e burninth n g capabilitie e corth e f o whils e improvin e safetth g y parameters, with special e voidinconcerth r gfo n coefficient e nee o Th increast d. e numbeth e f freo r e neutrons available for transmutatio f minoo n r actinide d eventuallan s y fission products i s emphasized, leadin o verw t arrangementg ne y r corfo se concepts, fued an l also coolant. These new ideas are in a very preliminary stage, the feasibility of such systems has to be confirmed.

In the Japanese OMEGA programme, the part concerning fast reactors s alswa o developed, with specific studie f reactoro s s devote o burt d n actinides; M-ABR with harder neutron spectrum, P-ABR with higher neutron flux, SLLC (Super Long Life Core) with no need for refuelling. In this case also, only the neutronic feasibility has been assessed, further studies being necessar confiro t y m these concepts.

When comparin e "figure proposeth gth l f merito eal d f o solutions" , it appears again that a thercommoa lac f s o ki e n mode o normalizt l e th e hazard reduction, particularl r whafo yt concern e reprocessinth s g process e relateth d dan reprocessing losses. Thi apparens i s n thii t s Sessiont ,bu also when comparing this Session to previous ones, for example for the same EFR core model whic s beeha h n treatel threal n ei dsessions e potentiaTh . l of fast reactors for hazard reduction seems obvious to everybody, but the o exprest y wa s thi se same th potentia :t no transmute s i l d masr faspe st reactor unit, ratio between the number of fast reactor and the number of light water reactors, long term radiotoxicity, etc. Also, the need to recycle the fission products was discussed, some of the participants being in favou f controllero d deep disposal. The separation techniques for the different wastes, especially the separation efficiency and the related losses in the residual wastes, were also discussed, while this topi s outsidwa c e scop th ef thi o e s meeting (separation of rare earth and curium). Nevertheless, it appears that if further progress in hazard reduction has to be achieved, improvements in separation techniques mus e obtaineb t d sinc e amounth e f losseo t s during fuel operations (reprocessing, fabrication) becomes the major part of these hazards.

e experimentaTh l backgroun d futuran d e programm s explainewa e y b d Russian papers, as well for the nuclear data as for some fuel process. Experiments performe n Japai d nn Russii (FCA d aan ) (BFS) showed quite similar results, with some discrepancie n actuai s l calculations. Further experiments are planned in BFS involving some 50 to 100 kg of Np-237. Concerning fuel, experienc s beeha en accumulate n BOR6o d 0 with oxide fuel obtained by vibropacking, irradiated up to 7.7% burnup for uranium oxide and 4.5% for mixed oxide. This type of fuel would allow simpler fabricatio d reprocessinan n g technology d coul,an e demonstrateb d n BOR6o d 0 provided funding availablee sar .

8 In summing up this session, one can observe that advanced fast reactor concepts can improve significantly the transmutation capability of classical fast reactors, provided that similar improvements are obtained in reprocessing efficiency.

Conclusions

The IAEA Specialists Meeting made clear the increasing interest in the issue concerning long lived radioactive nuclide d toxicitan s y aspects in nuclear waste managemen desirabilite th d an t a co-ordinate woro t t y ou k d approac s solutionit o t h .

e potentiaTh f o fasl t reactor r reducinfo s e lonth gg term radiotoxicit f speno y t fuel justifie n activa s e programm r conceivinfo e g and developing technologie r recoverinfo s g transuranic r in-corfo s e transmutation.

The specialists agree that future progress in solving transmutation problems could be achieved by improvements in:

- Radiochemical partitioning and extraction of the actinides from the spent fuel (at least 98% for Np and Cm and 99.9% for Pu and Am isotopes);

- Technological research and development on the design, fabrication and irradiatio minoe th rf no actinide s (MAs) containing fuels;

- Nuclear constants measurement and evaluation (selective cross-sections, fission fragments yields, delayed neutron parameters) especially for MA burners;

- Demonstratio e feasibilitth e f safd o econominth an e f o y A burneM c r cores;

- Knowledg impace th maximuf f o to e m tolerable amoun f raro t e earthn i s americium containing fuels. BACKGROUND PAPER The toxicity potential of the spent fuel is known to diminish with time (Fig. 1). STATU TRANSMUTATIOF SO N mose Th t toxic nuchde daughtes 24e it 'A sar d man r nuchde 237Np (whic s alshi o independently formed during nuclear energy generation by other processes) If the L KOCH transuranium elements are partitioned from the spent fuel and transmuted in a fast Institute for Transuranium Elements, theif o reacto% r 5 initia 0 o rt l value the radiotoxie nth c hazar reduces di factoa y db r of 200 compared to the initial values m general [2] 238pu 1S the only nuelide where Commissio Europeae th f no n Communities, this factor is not achieved due to build-up during the transmutation processes. Joint Research Centre, Althoug e radiotoxicitth h f fissioo y n products decays rapidly durin e firsth g t 600 years the hazard posed by the remaining nuclides is considerable, because some Karlsruhe of these nuclides are readily soluble in water and could migrate back to the biosphere Abstract During the last 20 years the feasibility of partitioning and transmutation schemes have been studied resulte .Th s were presente t severada l topical meetings The objectives and the proposed schemes Tor partitioning and transmutation of long living radiotoxic over the last few years [3 -13] nuclides are discussed In particular the constraints for the efficiency of a P & T process guiding R & D worfasa r tk fo reacto r fuel cycl describee ear d 2 OBJECTIVES 1 INTRODUCTION n ordeI develoo t r p 'Partitionin Transmutationg& ' scheme r radiotoxifo s c nuclides one should find answers to the following questions in order to define clearly During nuclear energy generation highly radiotoxic by-products such as the worD & k R neede scope e th th f edo minor actinides are formed, which accumulate in spent fuels or in reprocessing - Which nuclides and to what extent should be transmuted7 waste streams on the ton scale (Table I) 7 Becaus presene th f priceeo w uraniur tlo sfo littlr mo o theren incentivee ear s - Which of the proposed P & T schemes is the most promising one to recycle plutonium in LWR Therefore several states (USA, Spam, Sweden, and procesT & ? P efficien sbe e w th Ho -n ca t others) have opted for direct disposal of the spent fuel, whereas some states still hang on with the reprocessing of spent fuels in order to recover plutonium (F, GB, J, CIS) 2.1. Nuclides to be transmuted Other states consider both options e g FRG Initially the recovered plutonium was Durin e meetinggth s mentioned above several candidate radionuchder fo s foresee fueo nt l fast reactors becaust ,bu e sam e ofth e reason (low uranium pricee )th transmutations were identified, but not all of the listed radionuclides (Table II) pose deploymen thif o t s reactor type wil delayee b l t leas da year anothe0 r 4 fo to st 0 r2 the same radiological hazard to the public when they are eventually released from a Hence - as an alternative - the recycling of Pu in LWR is considered, where the geological repository If one assumes that such a repository will remain intact for reprocessing technology is established instead of its use in fast reactors This, several thousand years (as it is the experience for human made structures on earth) 137 however, will further increase the formation of minor actimdes (Table !)[!]. then nuclides with short half-lives like l37Ce, 9°Sr, Cs, i2iSn, i26Sn would decay. From the remaining only those which form monovalent ions could be transported bac watee y biospherekb th o t r : 14C, 36C1, 99Tc, 1291, l35C 237Npd san . Since eth latter nuclide has three parent nuclides: 24lpu, 24iAm, 245Cm all actimdes have to TABLE 1 FORMATION OF CERTAIN LONG-LIVING be considere potentialls da y hazardous. Durin migratioe gth n throug geosphere hth e RADIOTOXIC NUCLIDES IN 235U AND MOX the nuchdes will be diluted with natural stable isotopes, which could reduce the FUELED LWR (AMOUNT PER GWta) specific activity to insignificant levels depending on the geological formation. To support this argument the annual average uptake of the natural element by the "ICRP-man" [14 ]compares i annuae th o dt l limit uptakf so eacr efo h radionuclide Initial Fuel 235U (W/o) 32 07 [15], From this follow tolerable th s e isotopic abundanc consideree eacr th efo f h o d 239+241p (W/o) 32 radionuclide uptakee th n si n natural element wha(Tablo T . t eextenu) t sucn ha u isotope dilution will occur during the release from the repository is presently being investigated [3]. This concerns all nuclides except 99Tc and the actimdes Therefore, Burn-up (GWth d/t) 33 33 as a first priority, the artificial elements should be transmuted: 99Tc, 237Np and its parent nuclides 24lp , 241 Am, 245Cm spent fuel 237Np(kg) 141 42 u 241 Am (kg) 22 229 2.2. Proposed partitioning and transmutation schemes 243Am(kg) 28 54 1 There are several nuclear processes under investigation which could transmute a radionuclide into a less toxic one (Fig 2) Photons in the range of 242Cm (kg) 02 20 lOMeV produce s a Bremsstrahlund g from accelerated electrons induce 244Cm(kg) 08 285 d e fissioemployetransmutatiob an n actimde) m nn) ca r , fo d (y an sy (b n 99Tc(kg) 260 260 transmutation purposes [16-18 e direcTh ] t spallatio radionuchdef no higy b s h CO energetic charged particles has been proposed f!9]as well as the use of an inactive TABL . CANDIDATEU E RADIONUCLIDE TRANSMUTATIOR SFO N

Yearly uptake Iso topic 3,9 a of Nuclide T'/uta) ALIfBq] abundance S 3 elements [%] Ug 3 t7 [g) I"! 1.1 E5 114C1 5.7 E3" 1.5 E6i IE-8' 3 ** 1.9 E31 36C11 i E5 0 1.3E43. ' 5E-71 i*5** «, 7.3E-2i 1 E2 3 3. 12911 ' E7 6 1. 7E-2I _ o 3 . Q OX 2 5' 3.6E-3' 135CS1i E4 1 5. 20E61 3E11 !*!. 79Se2 6.5 E42 9-02. 93Zr2 1.5E62 90Sr3 28.53 S, 12ISn3 503 126Sn3 10.53

137CS3 30.173 3 99Te4 21E5-I 237Np4 E6-1 2. 1

238PU4 87.74

239PU44 E4 4 2.

240pu44 E3 5 6.

anthropogenic factor 241PU4 14.44

242PU4 3.7 E5-* 24IAm4 4324 3 o E» 242Am4 1414 243Am4 7.3 E34 245Cm4 8.5 E34 246Cm4 4.7 E34 247Cm4 1.5 E74 248Cm4 3.4 E54 1 Nuclides forming monovalent ions with natural isotopic diluent. z- -, re n o. o 2 Nuclides formin monovaleno gn t ions. 3 Nuclides with half-lives below lOOa decaying into stable nuclides. 4 Nuclides and parents forming monovalent ions with no natural diluent.

spallation source (e.g. Pb-Bi) to produce intense neutron fluxes [20 - 24]. The latter 5' "n concepts are not very different from nuclear reactors - discussed below - except that ô 5° they can use subcritical arrangements with thermal [20, 22] or fast neutrons [21, X drive] 24 acceleratorsy nb . Include installatione th n di reprocessine sar g facilities 5! Ô1 which separate the remaining nuclides from the transmutation products (if needed). The radionuclides are introduced either continuously as aqueous slurries or TABL . MINOEUI R ACTINID E- CONTAININ G FUELS IRRADIATED IN FAST REACTORS •S 2. • & 3 "V PFR y l Am«(" , 8005)0,95 24lAm2O3 PFR

244Cm203 PFR

(24IAm6244CmRE7)203 PFR 5 m Np02 KNKH

(2-HAmQ5Uo5)Ot92 KNKu

(Npo 4$Uo 55)O0 20 PHENIX T) PHENIX crë' (Npo 2^'Amo 2Uo e)Oi 95

0 73Pu(U 0 25Npo 02)O8 )9 PHENIX (Uo 73PUQ 2sAmo 02)0] gg PHENIX

solutions [20] or as molten s^lts [24] to the reactor or are added to fuel assemblies consistin alloyf go s [21 burneA ] M simila e r th concep o t r t being studie JAERy db I [25]. Other concepts propose using existing LWR [26] and FR [2, 27, 28] or HWR [22] for transmutation. Not only Pu produced in the LWR fuel cycle but also the minor actinides could fuel fast reactors. Several year MA-containino sag g mixed oxide fuels were developed and tested [21 (Table IÏÏ). The MA partitioned from the s a n thai u t P fuel e spenamoun R e th eithe. b f TheLW o t n abouo t rtyca % 10 t Z II > homogeneously mixed to the FR MOX fuel or concentrated in a few assemblies of the reactor. Their concentration woulfuele th abouf ,e do wel b % t2 l belo limie wth t that would influence the Na void coefficient [29]. For higher MA concentrations the geometr core th ef yo needaltere e b o st d [30] partitionine .Th minof go r actinides o. e A 3 (and selected fission products) from the spent fuel is not yet established. R & D work improvo recovert s undem i A y e electro-refininwa y re th yb f Zr-basego d alloys [31] and the coextraction of Am, Cm is achieved by theTRUEX process [32]. Due to the similar chemical behaviour of lanthanides and Am, Cm especially in the latter cas producte eth s contai lanthanidemuce o th n to f ho directle b o st y use fuelss da . Studies are under way to develop new processes to separate the two groups of elements [31,33). ü 2. O 3 ? N Since this meeting is dedicated to the use of fast reactors for transmutation purpose sI wil l briefly compare this concept wit above hth e mentioned onese Th . 1 3" present strategie copo st e wit risine hth g anthropogenic waste stream l kindal f so s are to develop processes which

£* - generate less waste, - recycle (and reuse) waste, - incinerate waste in special installations, - condition waste for geological disposal [34,35]. o g. c 2. A %• Z * e «v e optionTh e listear s d accordin publio gt c acceptance givin e firse gth th t e • highest and the last the lowest ranking. If these criteria are applied to the proposed concepts of P & T one should add four more: nuclear material safeguards, transportatio nucleaf no r materials between different sites, technical feasibilitd yan Ol t leastno last , impactbu nuclean o t r energy generating costadvantage .Th e th f eo accelerator-driven reactor concepts as well as the MA burner is that they can be used In 2 CD in symbiosis wit nucleay han r fuel cycl transmutr eo remainine eth g nuclear waste Tl/ beina ( —2 = absorptioe —gth n cross sectio neutroe th n i n flu. x0) after a "stop of nuclear energy generation" (Ausstieg scenario) or destroy nuclear 00 weapons material quickly. The technical feasibility of such systems, however, is not If one assumes that the time of the minor-actinide-containing fuel in the core of yet develope staga o dt e comparabl existine th o et g nuclear fuel cycle whicy ,b e hon the nuclear reactor is limited by the mechanical stability of its cladding (which is alsachievn oca e sameth e goals (with lower efficienc respecn i y transmutatioo t t n presently abou yeart3 fasa n sti reactor thad )an t during this perio t leasda t half fo rates but certainly less cost). Moreover in the public opinion the recycle of waste is minoe th r actinide fissior so n products transmutedhave b o et , the crose nth s section preferre s incineratioit o dt specian ni l installations. Hence e presenth f i , t nuclear shoule db reactor copn sca e wit posee hth d transmutation problems, they woul preferree db o dt any other concept. o > 1.5b(5-10i5n/seccm2 fast), MA's together with Pu can be recycled in a self-generated mode in existing b(5-1014n/seccm8 1 o> 2 thermal). LWR and FR [2], The differences between the two reactor types are: The LWR in an equilibrium fuel cycle contains more than 10 times transplutonium nuclides The need to recycle the minor actinides due to the limitations of present compare fasa o dt reactor fue equilibriun i l 391- neutroe 6 m.Th [3 n flu presenf xo t claddings leads often to a wrong perception. If one expects a reduction of the LWR is too low to transmute 99Tc and eventually '291 efficiently. This is not the case radiotoxicity to 99.6% (which is equivalent to about 8 refueling cycles and which also wherR blankee F th e fon ei th r t regio nthermalisea d neutron flu betweef xo - 5 n0. correspond life-time th nucleae o st th f eo r power station) thathee easile se tnon n yca 15 2 in total only twice the mass of the radionuclides have to be processed. 1 • 10 n/sec cm can be produced [401. Inherent with the transmutation of MA is the Wit beinhN e masgth s whic transmutes hi reactoa n di refuelina f o r g cycle build-up of 238pu which further denatures Pu in the sense of nuclear material safeguards especiall 240pw witlo R cas un ys F i h contentf it eo . Howeve t woulri a e db length , assumint , g tha abous i t same th tTl/s ea 2 (transmutation half-life th f o e nuisance to the fabricator of mixed oxide fuels - not of metallic fuel for the IFR where nuclide), the e decreasnth e initiath f efunctioa o l numbee s masa th u f o sN nf o r already remote fuel fabrication is foreseen. The latter concept eliminates the need of cycles (n), can be described as: fuel transportation, whic fuel- eve d casn R hi an oxidn sf - e o wilmorLW R e F l eth N(nt N= ) 02-ni/T difficulte b e highe, th mainlo t r e neutroydu n radiatio 244Cf no f o fresmn i d an h feedbaco a-deca(n o t e n Nk= du 2 yO 1/ considered ) 242Cm, 252Cf in spent fuels. MA actinide a r fuefas fo e tl ar sreactors o thae s additiona, th t l energy with generation and breeding rate would compensate for the extra costs in fuel make-up N (t)mas e reworkee ,th b o st eacr dfo h cycle. [2]. Since the energy generating costs of a LWR are considerably lower than that of a £ N (t), the total mass to be recycled is only 2 times the initial mass N0 FR this argument is presently not valid. The development of a P & T scheme 1/2"£ + )1 ( « £N(tN = ) (according to the present R & D programmes) will take more than 15 years, until . 0 N 2 = then the uranium prices may well favour the FR over the LWR. R witF he on-sitTh e reprocessin a gviabl e seemb e o t sconcep r fo t Under such circumstances one can estimate the reduction factor of the transmutatio presentld nan onle yth y existing reactor thareducn ca t s wasteit e radiotoxicity presene .Th t reprocessing technolog subsequend yan t fuel make-us pha efficientl recyclingy yb shoult forgotten.I e b t dno , however, tha alternativn ta e fuel losses of minor actinides to the waste streams on the order of 0.3% per cycle. Hence cycle based on the thorium molten salt reactor [35] could prove to be even more accumulatee th d losses durin l reprocessingal g stages woul 0.6e db % which together advantageous productioA , sincM e eth n ther considerabls ei togetheyA loweM d ran wit remainine hth g radionuclid0.4e th %f o e th f eo afte% 1 cycle8 ro t p u s wilm su l with fission product transmutee b n ca s s wellda . Toda technicae yth l feasibility original radiotoxicity, i.eoverale th . l reductio initiae th f no l radiotoxicity wile b l (despite extensive efforts at ORNL in the past) is not yet proven. hundred times. Thi stils si significana l t reductio environmentan a f no l hazarf di compare thao dt othef to r noxes exhause likth en NO(i 2 t Xgase,SÛ automobilef so s and fossil fuel power stations) having a reduction by only a factor of 3. 2.3. Efficiency of P & T schemes If one has agreed on the kind of nuclides to be transmuted, the question arises: 3. OUTLOOK to what exten dschemT shoul& P e da reduc radiotoxicite th e a particula f yo r nuclide? Earlier answer thio st s question proposed "hazard reduction factorsf o " If nuclear energy generation is to be used by future generations then the several order magnitudf so e [41] which aime eliminatiot da nuclidef no s dowe th o nt limited suppl 235f yo U will rene interese w fasth e th t n breedeti r concept. Regardless "maximal permissible concentrations". Any realistic approach, however, has to of what future fast reactors will look like, they will hav inherene eth t potentiao t l consider losses during the P & T processes, which have still to be buried in a burn most radiotoxic waste. Therefor e developmenth e e minor-actinideth f o t - geological repository largw Ho e. these losses are, depende availablth n o s e containing fuel cycle for fast reactors is an essential task in keeping the nuclear technology, therefore one should retreat to the ALARA principle and reduce the energy option open. radiotoxicity of a certain nuclide as low as reasonably achievable. For the existing Apart from basic researc f morho e effective partitioning processe e basisth c fuel cycl ewhe- coule non d - extende attaiT & nP reductioa o dt n unde0 facto10 f o r physico-chemical propertie f MA-containino s g fuels studiee havb o d theiet dan r the following constraints [34]: irradiation behaviour tested. Since one expects a reasonable reduction of the radiotoxic nuclides during the For the IFR concept the electro-refining process has to be extended to extract life-tim determinn a reacto f ca o ee on re lowe eth r limicross-sectioa f o t e th r nfo l minoal r actinides froe spenmth t fuel wit a hsufficien t decontamination from neutron reaction causin e transmutationgth half-lifee Th . , Tl/2 thif o , s process si lanthanides and to isolate technetium which accumulates with other elements in the equatioe giveth y nb n cadmium cathode. Investigations in these directions are under way, where the developmen n appropriata f o t e fue s alsi l o included [31, 331. This fue s beini l g [17] KHLEBMIKOV, S.V., OSETROV, O.I., Proceedings of the Workshop on developed in a joint effort between the European Institute for Transuranium 'Nuclear Transmutation of Long-lived Nuclear Power Radiowastes', Institute Element CKEEPd san wild testee I an lb join a n di t irradiation experiment witA hCE Nuclear Power Egineering Obninsk and Institute of Theor. and Exp. Physics in PHENIX. Moscow, Obninsk, July 1991 oxide-fuelee th r Fo d fast reacto reprocessine rth g covering minor actinided san [18] TAKASHTTA . al.et , , OECD/NEH. , A Specialists' Meetin 'Acceleratorn go - t technetiubeeye nt developeno s m ha comparabl a o IFRde t th r . fo s ei stagt i s ea Based Transmutation', Würenlingen, March 1992 Nevertheless promising extraction schemes are under test which would not only [19] WENGER,H.U.,et.al.,ibid. extract these nuclide spenR stLW t fro fuealsoe bu mth l later coul e useb do t d [20] IRELAND,!.,ibid. reorocess the FR fuel [33, 42]. Minor-actinide-containing fuels have been tested in [21] TAKAHASHI , ibid,H. . PHENIe th X reactor (Tabl [43]) post-irradiatioe eHI .Th n examinations revealeo dn [22] KAZASrrSKY, V.D., ibid. unexpected behaviour under irradiation follow-uA , p irradiation experimens i t [23] TAKADA . al.et , , SpecialistsH. , ' Meetin 'Accelerator-Driven o g n Trans- presently under preparation which should lead to higher burn-ups. These fuels are mutation Technolog Radwastr yfo Othed ean r Applications', Stockholm 1991 developed jointly with the CEA, France and the European Institute for [24] KAWATA, T., et. al., Symposium on 'Separations Technology and Transuranium Elements and have been or will be irradiated in PHENIX. Transmutation Systems', Nat. Acad. of Sciences, Washington, January 1992 [25] MUKAIYAMA . al.et ,, Proceeding,T. 'Inte th . f sConfo Fasn o . t Reactord san Related Fuel Cycles "FR 91'", Kyoto, October 1991 [26] LEFEVRE , SymposiuJ. , 'Separationn mo s Technolog Transmutatiod yan n Systems', Nat. Acad. of Sciences, Washington, January 1992 [27] CHANG,Y.I.,ibid. REFERENCES [28] INOUE, T., et. al., "Characterization of Fuel Alloys with Minor Actinides", ANS Winter Meeting, San Francisco, November 1991 [29] BALZ, W., et. al., Proceedings of the 'Int. Fast Reactor Safety Meeting', ] KOCH[I , 'TransuraniuL. , m Elemen Symbiosis"R tF Fue- R l CyclLW ,n i e Snowbird, Utah, August 1990, Vol. 1 (1990) 153 Proceeding e '5tth hf o sInt . Conf Emeginn o . g Nuclear Energy Systems', [30] THOMPSON, M.L., Symposium on 'Separations Technology and Trans- ICENES, Karlsruhe, 3-6 June 1989 mutation Systems', Nat. Acad Sciencesf .o , Washington, January 1992 [2] KOCH, L.,'"Minor Actinide Transmutatio WastA n- e Management Option", [31] SAKATA, M., et. al., Proceedings of the 'Int. Conf. on Fast Reactors and Journa Less-Commoe th f lo n Metals, 122, (1986)371-382 Related Fuel Cycle 91"'R s"F , Kyoto, October 1991 [3] Proceedings of the '1st Technical Meeting on the Nuclear Transmutation of [32] HORWITZ, P., Symposium on 'Separations Technology and Transmutation Actinides', JRCIspra, 16-18 March 589 R 19777EU (1978, ) Systems', Nat. Acad Sciencesf .o , Washington, January 1992 (4) Proceeding '2ne th df s o Technica l Meetin Nucleae th n go r Transmutatiof no [33] CECILLE , JAERJ/OECL. , D Meetin 'Partitioninn go Transmutation'd gan , Actinides' IspraC ,JR , 21-24 Apri 692l R 19809EU ,(1980 ) Mito, Japan, November 1990 [5] OECD/NEA Expert Meeting on 'Options Making Extra Gains of Actinides', [34] KOCH , Proceedings^oL, , Workind 2n e th fg Group Meetin Targetn go d san Paris, April 1989 Fuels', CEC, JRC Karlsruhe, 23-24 June 1992 (to be published) [6] KOCH, L., WELLUM, R., (Ed.), Proceedings of the Workshop on 'Partitioning 1 [35] FURNKAWA , ProceedingK. , e IAEth f Ao s Advisory Group Meetinn go Transmutatiod an Minof no r Actinides ,Karlsruhe CECC JR , , October 1989, "Partitioning and Transmutation of Actinides and Selected Fission Products EUR 12849 (1990) from HLW', Vienna, October 1991 [7] JAERI/OECD Meeting on 'Partitioning and Transmutation', Mi to, Japan, [36] KOCH, L., et. al., Proceedings of the '4. Int. Transplutonium Element November 1990 Symposium', Baden-Baden (1975) 459-469 ] Proceeding[8 e Workshoth f o s n 'Nucleao p r Transmutatio f Long-liveo n d [37] OLIVA, G., Proceedings of the '2nd Technical Meeting on the Nuclear Trans- Nuclear Power Radiowastes', Institute Nuclear Power Egineering Obninsk mutation of Actinides', JRC Ispra, 21-24 April 1980, EUR 6929 (1980) 17 and Institute of Theor. and Exp. Physics Moscow, Obninsk, July 1991 [38] GUARDINI, S., SMITH, B.G.R., ibid., p. 43 [9] Proceedings of the IAEA Advisory Group Meeting on 'Partitioning and [39] SCHMIDT, E., ibid., p. 471 Transmutation of Actinides and Selected Fission Products from HLW', [40] SALVATORES, M., et. al., Proceedings of the IAEA Advisory Group Meeting Vienna, October 1991 on 'Partitionin d Transmutatioan g f Actinideo n d Selectean s d Fission [10] Proceeding Specialistse th f so ' Meeting on'Accelerator-Driven Transmutation Products from HLW', Vienna, October 1991 Technology for Radwaste and Other Applications', Stockholm, 1991 [41] CLAIBORNE, H.C., "Nuclear Induced Transmutation of High-Level [II] Symposium on 'Separations Technology and Transmutation Systems', Nat. Radioactive Waste", ORNL/TM-396 41972( ) Acad Sciencesf .o , Washington, January 1992 [42] MUSIKAS . al.et , , C. "Result, Prospectd san somr sfo e Extradante th n si [12] OECD/NEA Specialists' Meeting on 'Accelerator-Based Transmutation', Minor Actinide Partitioning", Proceeding Workshoe th f so 'Partitioninn po g Würenlingen, March 1992 Transmutatiod an Minof no r Actinides' ,Karlsruhe CECC JR , , October 1989 [13] Proceeding Workind 2n e th g f sGrouo p Meetin Targetn go Fuels'd san , CEC, [43] PRUNIER . al.et ,, 'TransmutatioC. , Minof no r Actinides: Behaviour under KarlsruheC JR , 23-24 Jun publishedee b 199o (t 2 ) Irradiatio f Americiuno Neptuniud man m Based Fuels", Proceedinge th f so [14] REFERENCE MAN, ICRP 23 (1975) 'Int. Conf. on Fast Reactors and Related Fuel Cycles "FR 91"', Kyoto, October [15] Strahlenschutzverordnung BundesgesetzblatD ,BR , (1977l t ) 2905 1991 [16] KÄSE, T., et. al., Specialists' Meeting on 'Accelerator-Driven Transmutation Technology for Radwaste and Other Applications', Stockholm, 1991 NATIONAL PROGRAMMEN SO TRANSMUTATION OF MINOR ACTINIDES AND LONG LIVED FISSION PRODUCT FBRN SI s (Sessio) 1 n

Chairmen

L.A. KOCHETKOV Russian Federation U. WEHMANN Germany UNITED STATES NATIONAL PROGRAN MO The processes being studied includ decladdme eth spene fueR th f l gLW o tpin s ACTINIDE RECYCLE by suitable mechanical and/or chemical means and the subsequent pyrochemical (Summary) decomposition of the spent fuel to provide (a) a transuranic product containing the transuranics plus a portion of the lanthanide fission products, and designed for introductio ALMe th o Rnt electrorefinme cyclth t ea g uranium-rica step ) (b , h H KHALIL component (composin spenR LW t fuee bule th glf th k masso ) suitabl storagr efo e Argonne National Laboratory, and potential future use as the source of LWR or ALMR fuel, and (c) waste Argonne, Illinois, streams that can be converted and packaged into forms acceptable for geologic United States of America disposal In the current, early phase of the actmide recycle program, three different pyrochemical separation concepts, referred to as the "salt transport," "magnesium The current U S National Energy Strategy includes four key goals for nuclear extraction," and "zinc-magnesium" processes, are being investigated using policy maintain safety desigd an n standards, reduce economic risk, reduce regulatory riskestablisd an , effectivn ha e high-level nuclear waste prograA m small-scale experiments to establish chemical feasibility and to identify the most attractive process for further development In the subsequent phase, the scale will potentially effective means of reducing the long term radiological toxicity of high-level wastes destined for geologic disposal (primarily spent LWR fuel) is to be increased to examine the feasibility of engmeenng aspects of the selected extract the transuranic irradiation products and to destroy them by transmutation process, such as materials compati- bility, equipment scaleup, and process in reactors or accelerator concepts Actmide recycle was considered previously in integration The program goal is to provide sufficient information to support a technical feasibility decision in 1995 Large-scale demonstration of the process the United State d rejectean s d becaus f proliferatioeo n concern littld ean s perceived benefit to the viability of light water reactor (LWR) deployment or to with LWR spent fuel is planned pending a favorable decision at that time disposal The current U S program, using new fuel processing and waste o addresT s fuel fabricatio m-reactod nan r performance issues, wor s alsi k o management technologie modulara d san , passively safe advanced liquid metal underway to demonstrate fuel fabrication by injection casting and to achieve reactor (ALMR) concept, offers the prospect of overcoming these concerns acceptable burnup level n si fue l containind g an mino ) rAm actmided an p s(N lanthanide fission product impurities Initial fuel casting studies have been The United States national progra Actmidn mo e RecyclS directeU s ei e th y db Department of Energy (DOE), who is responsible for development of energy concerned with verifying adequate contro f fabricateo l d fuel compositions Upon actmide Th D e recyclR& polic d yan e progra mcoordinates i d wit Integrae hth l establishment of adequate fuel casting parameters, fuel elements will be cast for Fast Reactor (IFR) metal fuel development prograALMe th d Rm an program l al , irradiation theid an ,r m-reactor performanc , fuel-cla g e e( d compatibility) wile b l actmide Th sponsore* e E recyclDO y edb approach comprises technologr yfo established by periodic examinations as a function of exposure the recovery and conversion of the transuranic content of LWR spent fuel to a helo T p asses abilite sth f currenyo t nuclear datcomputationad aan l procedureo st metallic fuel for pyrometallurgicay mb l subsequenmeans it d san t consumption i predict the relevant actmide transmutation rates, actmide samples will be irradiated an ALMR fuel cycle in EBR-II for the purpose of measuring their reaction rates and companng these measurements with calculational estimates Actinide Recycle from LWR Fuel Neutronics studie beine sar g performe evaluato dt performance eth safetd ean y Th ecloseR extensioIF e d th fuef no l cycle pyroprocess technologe th r fo y characteristics of core concepts in which the actmide management objective ranges extraction of transuranics from LWR spent fuel for consumption in ALMRs is being from efficient breedin transuranit ne go t c consumption Related system evaluations undertaken m the actmide recycle program at Argonne National Laboratory wile conducteb l f LWR/IFo d R synergistic fuel cyclee scenarioth f o d an s Pyrochemical processe beine sar g develope purpose th r fo f deconomicalleo y radiological hazard implication f theso s e scenarios orden i , o optimizt r e th e extracting the transuranic species from LWR spent oxide fuel and for management of transuranic mventones concentrating these transuranic metallia n si c form suitabl r introductioefo s na feed matenal into the IFR fuel cycle The goal of the process development is to recover at least 99 9% of the LWR discharged transuranics for use in ALMRs Advanced Liquid Metal Reactor (ALMR) Development reactoe Th r desig sodium-coolea s ni d pool type reactor Powee baseth n do r S Rosen and C E Weber "The ALMR as a Future Energy Source for the United States " Reactor, Innovative Small Module (PRISM) concept originated by General Electnc ro Proceedings of International Conference on Fast Reactors and Related Fuel Cycles The design is being performed by a GE led industry team The current reference Kyoto Japan (1991) Paper1 1 design is a 471 MWt modular reactor (nine modules constitute a 1440 MWe plant) 10 ALMR passive safety features assure accommodatio f anticipateno d transients LONG LIVED WASTES TRANSMUTATION ro without scram A passive reactor vessel auxiliary cooling system assures STUDIES IN FRANCE safety-grade decay heat remova reference Th l e core desig breedea s ni r core utilizing ternary metal fuel (U-Pu-Zr) Actmide recycl incorporates ei d inte oth M SALVATORES, C PRUNIER, reference core design by designing cores to study the possibilities of transmuting actmide botn si h breede burned an r r configuration scurrene whicth e hus t reactor P BERGEONNEAU, A ZAETTA desige NucleaTh n r Regulatory Commission (NRC )s i reviewine th g Commissariat à l'énergie atomique, pre-application Preliminary Safety Information Document (PSID) and preparing a Centre d'études nucléaires de Cadarache, Safety Evaluation Repor componentsty (SERKe ) , including electromagnetic pump Samt-Paul-lez-Durance stators and seismic isolators, are in the testing phase Future work includes commercial participation in building a prototype test module Safety tests in the H SZTARK prototype will be the bases for final plant design certification by the NRC Framatome, Lyon Overview of Technical Papers G VAMBENEPE, J VERGNES Electricité de France, Two technical papers fro Unitee mth d States wilpresentee b l t thida s meeting The first paper is "Physics Considerations in the Design of Liquid Metal Lyon Reactor Transuraniur sfo m Element Consumptio n" Thi sKhalilH pape y b ,s i r France R Hill, E Fujita, and D Wade (all ANL) and will be presented by H Khahl This paper addresses concepts which exploit key features of the metal fuel cycle and the hard spectrum characteristics of ALMRs Physics aspects of actmide Abstract transmutation are discussed for a range of transuranic management options In this paper Frence th , h activitie LMFBRr sfo fiele f th basi do n i , c data validation, reactor The second paper is "Actmide Transmutation in the Advanced Liquid Metal and fuel studies are reviewed Reactor (ALMR) y Christinb " e Cockey (GE) This paper addressee th s possibilities of actmide transmutation in the reference metal-fueled ALMR Actmide transmutation core designs are presented as well as lifetime actmide mass consumption values for different configurations 1 INTRODUCTION

Reactor incineratio f transactmideo n long-lived an s d fission product s considerei s a s a d possible complementary strategy to the deep storage of high-level radioactive wastes

The main objective is to envisage the reduction of the quantities of toxical wastes to be stored, the longterm safety (and then the risk reduction) of geological repositories being a separate issue

In France, CEA, FRAMATOME and EDF make studies on this topic The French utility EDF, since it is the main producer of HLW, is much concerned by the management of these wastes in this respect, EDF takes a financial part m the programme developped in CEA FRAMATOME, as reactor constructor and fuel manufacturer, participates also to this programme

Concerning transmutation potential in LWR and accelerators, studies are also performed but scop e describee th b wilt f f thieo o no lt s thiseminaou dm s si nott i s rea 2 INTEGRAL EXPERIMENT DATR SFO A VALIDATION SUPERFACT 1

In France, a large experimental data base is available to validate the basic nuclear data Both 19 PINS PHENIX CLUSTER irradiated fuel expenment d integraan s l expenments have been performed, whic n givca he pins location informations on MINAC and fission product data

2 1 IRRADIATIONS IN FAST NEUTRON SPECTRA

Irradiations have been performe PHENIn di f Xo

- special fuel pins with variable Pu vectors (TRAPU experiment), - separated isotopes samples (PROFI PROFId an L1 Lexpenments)2 , - special fuel pins with differen (SUPERFACm A d t an content p N f To s expenment), (Figure 1)

PROFIL-1e Th , PROFIL- TRAPd 2an U expenments have been analyze termn di JEF-f so 1 data [1] Indications have been obtained on integral capture and fission cross-section, on some (n,2n) reactions (in particular for Np-237), and on some branching ratios in the decay schemes

analysie SUPERFACe Th th f o s T expenmen stils i t l underway

2 2 LWR AND HCLWR NEUTRON SPECTRA

As these reactor types are out of the scope of the present meeting, we will not detail the corresponding work programme

Note only, thathesr fo t e reactor types, both irradiation expenments (pool reactor MELUSINE at GRENOBLE) and expenmental analysis (JEF-2 library) are performed

3 REACTOR STUDIES homogeneous heterogeneous actmide pin standarn dpi actmide pin Reactor studies are devoted to actmide burners (LWR, FBR) with homogeneous and heterogeneous recycling

Previous studies have shown the order of magnitude of the potential incineration in a large fast reactor Figure 1 Progressively destroyR FB actimdee e th sth , s initially loade wels d(a s thos a l e producey db operation)n ow s it afted an , r approximatel cycles7 o t e actmid 6 th ,y e amounts reacn a h equihbnum, with a reduction by a factor of 10 of the initial quantity of Np-237 - heterogeneous recycling, i e the Np-237 and the Am-241 are added in large quantities n pi 2 UO a o t ) % 0 2 d an % 5 (4 In the PHENIX reactor, a first demonstration of this potential has been made (the SUPERFACT expenment [2]) Two types of pins have been irradiated for 380 FPD, each type Present studie focusee ar s d being representativ recyclino tw e th gf eo mode s n homogeneouo ] 1 s recyclin n LMFBRi g s with different fuel typeg oxide ( d s an e nitnde), and of different sizes, and on heterogeneous recycling in special subassembhes (O - homogeneous recycling, i e Np 237 and Am-241 (2 %) mixed to the fuel of standard CO UO2/PuO2 pin , (e g in the periphery of a standard core, or in the decoupling regions of a modular ro core) where the actmides/fission products can be irradiated during all the life of the MINOR ACTIMDES RECYCLINN A N GI power plant Preliminary results indicate a reduction of the Np-237 quantity of at least EFR TYPE FAST NEUTRON REACTOR , whe% n0 2 irradiate e externath n i d l radial blanke a SUPER-PHENI f o t X type reactor , H SZTARK 2] on homogeneous recycling in LWR with different moderating ratios Heterogeneous Framatome, recycling is also investigated Lyon

A special attention is given to the reactivity coefficients, which can potentially be worsened G VAMBENEPE J ,VERGNE S by the addition of minor actimdes, and to the optimization of the spatial power shapes Electricit Francee éd , Lyon For long-lived fission products (such as Tc-99 and 1-129), for which the epithermal capture proces maie th ns si mechanis f transmutationmo explores i t i , feasibilite dth transmuto yt e them A ZAETTA at the periphery of a FBR core in special subassembhes containing a moderating material Commissaria l'énergià t e atomique, Centre d'études nucléaire Cadarachee d s , This technique was successfully proved in PHENIX for the production of Co-60 from Co-59 Samt-Paul-lez-Durance targets surrounded by a calcium hybnde moderator France Complementary studie performee sar morn do e advanced burner systems (accelerator, hybnde systems, etc ), as well as studies on fuel cycle consequences of the introduction of long-lived Abstract wastes burner reactors framewore Inth frence th f hko nuclear programme typR , reactorf baseeo PW n do s associated with reprocessing of spent fuel elements, the management of the nuclear wastes and particularly the Minor Actimdes has to be defined 4 FUEL STUDIES Amon possible gth e solution transmutatioe th s f thesno e long lived elements using Fast Neutron Reactors such SUPERFACe Th T experimen demonstrates ha t gooe dth d performanc (UNp)Ue th f eo 2 fuel as SPX and EFR is envisaged This paper aims at presenting the studies which are being performed on an EFR type and some problemproductioe th g e ( s n therma , Np)Ol Am conductivity , fue(U e l th r fo ) coref o , , oxid150We 0eM fuel with very cashige f minoth eho n i bur , r nup actimde s homogeneously distributed 2 insid e efficience fue Th e th Mino lth f o ry Actimdes transmutatio a functio e initias a nth f o ln contente th , Present studies are devoted to nitnde fuels (for the homogeneous recycling) and to different consequence e insertioth f o s f thesno e Actimde majoe th n r so core parameters, mainl safete yth y parametere ar s types of inert matrices for the heterogeneous recycling A matrix NpQ + MgO has been already presented Finally consequencee th , f thiso s Minor Actimdes recyclin lone th gn gtero m radiotoxicit givee yar n experienced successfull paste othet th bu ,n i yr matrices (ZrO2, A1 2Oinvestigatee 3ar ) d

Also, a joint experiment with ITU/Karlsruhe and CRIEPI/Japan, is planned for irradiation 1 INTRODUCTION of metallic fuel pins (actimdes and some rare earths in UPuZr matrix), in PHENIX framewore Inth frence th f kho nuclear programme typR , reactorf baseeo PW n do s associated Finally reflexioa , beins ni g made, abou future th t f SUPER-PHENIeo thin XI s contexd an t with reprocessin f spengo t fuel elements managemene th , nucleae th f o t r waste particularld san y as expressed very recently by French Officials, the SUPER-PHENIX plant could be used as a Minoe th r Actimdedefinee b o t ds sha waste incinerato than I r t respect, NERSA (owneplante th f )o r must propose some specific experiments this work is just at the early beginning Amon e possiblth g e solutions transmutatioe th , f theso n e long lived elements using Fast Neutron Reactors such as EFR is envisaged The present study aims at presenting the capability of a EFR type of core, 1500 MWe, oxide fuel with very high burn-up, to transmute Minor REFERENCES Actimdes homogeneously distributed insid e efficience fue Th e Minoth el th f o ry Actimdes transmutation as a function of the initial content, the consequences of the insertion of these [1] A D'ANGELO et al Actimde majoe th n rso core parameters, mainl safete yth y parameters presentee ar , d Alsoe th , , 105,24 g En l 4 Nuc(1990i Se l ) consequence f thiso s Minor Actimdes recyclin lone th gn gtero m radiotoxicit discusses yi d

[2] C PRUNIER et al 2 CONTEX STUDIEE TH F TO S "Transmutatio f minono r actimdes Behaviour underp effecN d f irradiatioo tan m A f no based fuels" In orde woro rt k with well established material quantities presene th , t studies wil centeree b l d "Int Con Fasn o f t Reactor Related san d Fuel Cycles', KYOTO t 28-NoOc , , 199v1 1 aroun e frencdth h nuclear programme Today e frencth , h nuclear electricity generation rat s somewhai e t higheh TW r tha0 30 n - total thermal power 3600 MWth, electrica expectes i year t i pe l , r increaso dt e slightl reaco t d ratha y an abouf e o TWh 0 45 tr epe - homogeneous core concept with two radial enrichments, year around 2010 This nuclear electricit s mainlyi y producetypR ePW reactorsy db , using - 388 fuel subassemblies, distributed respectively into 208 and 180 between inner and outer uranium oxide fuelled assemblies zones, - 1 meter fissile height, With such type of fuel, the heavy nuclei production rates at the reactor output are given here - maximum fuel burn up 20 at %, below, per each TWh electrical produced - fuel managemen batches5 n i t ,efp 0 witd32 ha cycl e lengttotaa d lh an fue l residence time of 1600 efpd (a) 2900 kg of " Reprocessed Uranium " containing 98 6 % U238 ,09% U235 ,05% U236, The fuel itself is made of mixed oxide UO2-PuO2, with the following compositions for the reference case (b) 35 kg of plutonium containing 2 9 % Pu238 , 54 3 % Pu239 , 22 9 % Pu240 , 14 3 % Pu241 ,57% Pu242, depleted uranium 0 25 % U235 , 99 75 % U238, - plutonium cominR t froPW g ou ma (c) 2 5 kg of Minor Actimdes with 2 8 % Pu238 , 54 4 % Pu239 , 22 8 % Pu240 ,118 Pu 241 , 7 % Pu 242, with a Np23Am24% % 4 8 0 6 72 , 1, Am24 Cm244% 6 3, , 1 24 m A f o % 2 1

(d) 135 kg of fission products, with in particular With this fuel composition, the fuel enrichments in plutonium oxide are respectively 16 8 % Tc9f o g 9k 3 3 - r botfo h % radia 9 1 l2 zoned an s 112f o 9 g k 8 0 - When adding Minor Actimde coree th ,n s i thes isotopew ene homogeneousle sar y distributed All these quantities result from fuel burn up calculations, which are then compared with all over the fuel, in replacement of heavy atoms of uranium and plutonium The fuel enrichment results coming out from the analysis of samples from some well known irradiated fuel in PuO2 is adjusted in such a way to get the same service as for the reference fuel subassemblies - same burn-up, which means same cycle length and overall fuel residence time, For an averaged electricity generation rate of 400 TWh per year, we will then have the - same reactivity at end of equilibrium cycle following heavy nuclei production every year, assuming that only uranium oxide is used as fuel The core calculations are performed in a 2 dimension model in RZ geometry, using the Reprocesseton" 0 f 16 so 1 - d Uranium" ERANOS code system and the CARNAVAL IV cross section data set, with the standard 25 14 tons of plutonium, energy groups Thi scompletes i dat t JEFe ase 1 th y db cros s section Minoe th r rsfo Actimdes 1 ton of Minor Actimdes batche5 e Th s fuel subassembly managemen t simulateno s i t d exphcitel calculatione th n yi s Presently PWR'e th ,par w onlf e recyclinfe o t s ar y a g plutonium onld an ,y partialle th n yi whole Th e cor burns ei t from f lifbeginnino e d (160en 0 o t f lif go p efpd) eu , thequilibnum core e fueX th lf MO assemblieo f cas5 e o r cor th e fo % onl s i e( r 0 reload ywa 3 s pe t i 1991n i ,) cycle being represented respectivel 0 efp96 d efp 0 d Wit64 an d t ya h respec fuee th l o itselft t , M0 poweW 90 rexpectes i plants t i d an ,d tha plant6 1 t s will recycle (partially) plutoniue th y mb one lifetime in the core will represent one fuel cycle end of the century

In these conditions foresees i t i , n tha amoune th t f unuseo t d plutoniu Minod man r Actimdes 2 MINO3 R ACTINIDES AUTO-RECYCLING will approach 300 tons before 2010, only for France, the Minor Actimdes themselves representing some tens of tons In a first step, the auto recycling of Minor Actimdes (MA) was studied in order to determin burnine eth gr corthi R capabilitFo seEF purposee th f yfollowino e th , g assumptions were made

3 STUDIES OF MINOR ACTINIDES TANSMUTATION IN AN EFR TYPE OF CORE e firsfoth r t fuel cycle a give , s insertei n A amoun dM insidf o te fresth e h fuel, homogeneously distributed all over the core 2 %, 5 % and 10 % first of Np237 and then 3 1 CALCULATION MODEL f Am24o 1 Indeed contributorspenR m , thes ma PW isotope to e e fueletw th th e ,n i ssar complete th e corcomplete s burnon ei r fo t e fuel cycle (1600 efpd), to purpose presene th th r reference f e takes Fo eo th t wa s corstudies nR a ] e[1 EF ,e witth , e hth the burnt fuel is supposed to be reprocessed, the fission products being extracted and in following characteristics replaced by a mixture of heavy atoms (U, Pu) without any new M A , in such a way to ro correce th t ge t reactivity same th ed heavan , y meta previoue th lr masfo s sa beginninf go 1000 en fuel cycle, f pilo t e isotopiou e th c- evolutio simulates ni takiny db yea g1 r fue fo r l fabricatio1 d nan yea r fuefo r l coolin reprocessingd gan , 800 - the same process is being repeated for each new fuel cycle

With these assumptions establisn ca e burnine on ,hth burnine gth ratd egan spee r eacdfo f ho isotopeo tw e th s individually totae th als d l r amountaknA an o,fo M g f o intt o account those y, rt whic e producehar fuee th l n di S

The results of these evolutions are given on Tables I and II respectively for Np237 and Am241, in terms of masses of the involved isotope, and in terms of the sum of the 3 isotopes leading to Np237 , Np 237 + Am241 + Cm245 On Fig 1 and 2 are plotted the cases of 2 % Np237and2 % of Am241

TABLE I EVOLUTION OF THE M A MASSES FOR THE CASE OF Np237 INSERTION AUTO-RECYCLING 9 10

Masseg k n si Number 2 % Np237 5 % Np237 10 % Np237 of cycles FIG 1 Evolution massesA case the Np237M % ofof2 the for insertion Np7 E (3 iso)' Np7 iso)3 E( " Np7 iso)3 E( " 0 902 950 2257 2306 4518 4569 2 253 408 610 766 1251 1409 4 93 251 184 338 358 508 1200 6 52 219 75 237 120 277 8 41 213 47 216 58 223 1000 M f Np23o m iso3 7Am24+ ( su )E 1= Cm241+ 5

TABLE I I E CASA MASSE TH F EVOLUTIOAm24M O E R E FO S1TH F O N INSERTION AUTO-RECYCLING

Masses in kg Number Am24% 2 1 5 % Am241 1Am240% 1 of cycles Ami E (3 iso)' Ami E (3 iso)' Ami E (3 iso)* 0 1027 1027 2336 2336 4625 4625 2 339 337 672 714 1328 1383 4 182 234 245 301 428 492 0 1 234567 6 155 211 164 222 210 273 Number of Cycles 8 155 211 154 211 163 222

1 E (3 iso) = sum of Np237 + Am241 + Cm245 FIG 2 Evolution of the M A masses for the case of 2 % Am241 insertion e thase t n whateveca initiae e th On s i rl actinidamouns it frese th d hn ean i t fuel e finath , l Reference core : composition of the fuel is the same after about 5 to 6 fuel cycles, this final composition Pu23% 7 8; 48.3. 7 Pu239 ; 36.3 % Pu 240 ; 5.5 % Pu241 ; 5.8 % Pu242 dependin corge th onle n characteristicsyo . Indeed, this final compositio s constanni t evef ni there are no M.A. inside the fuel at the beginning. This equilibrium state is obtained when the Higher enrichment core: productio f M.Ano . insid core eth equas ei theio t l r destruction. 3.9 % Pu238 ; 45.2 % Pu239 ; 37.6 % Pu240 ; 6.2 % Pu241 ; 7.1 % Pu242.

burnine Th g rat thif eo s establishetype b f coreo n eca abouo dt : t Wit highee hth r enrichment core plutoniue th , m qualit termn yi f equivalenso t plutoniu9 m23 is slightly lower, but in both cases the isotopic composition is stabilized after 5 to 6 fuel cycles. Np23e th r 7fo alone% 9 9 , o t % 5 9 - Also, for both cores, as the plutonium quality is decreasing when recycling it, the fuel - 85 % to 96 % for the Am 241 alone, enrichment has to be increased at each new fuel cycle : this increase in enrichment is of the + Cm245 1 24 Np23m .m A su e 7+ th r fo % 5 9 o t % 5 7 - relativ% 0 orde1 ef o rbetwee e firs nth where lase t th cyclon tfuee d eth l ean compositio s ni fuee th lstabilized s i enrichment o s d an , . Concerning the burning speed, whatever is the initial amount of M.A. inside the fuel, almost half of this amount is burnt after one single fuel cycle. For the sum of the three actinides leading This shows that whateve e typ th f core eo s i r , wit r withouo h t M.A. recycling e Fasth , t . Np237o % t 0 5 o ,t thiorde0 e 4 sth f ratif o ro s oi Neutron Reactor allows indefinite recycling of plutonium.

3.3. INFLUENCE OF THE MAIN CORE CHARACTERISTICS ON THE BURNING CAPABILITY 4. CONSEQUENCES OF M.A. RECYCLING ON THE CORE PERFORMANCES

In order to check the influence of the core characteristics on the burning capabilities and also 4.1. M.A. MASS LIMITATION equilibriue onth m compositio fuele th ,f no comparativ e calculations were performed wite hth referenc corR s defineEF ea d previously witd an cor,a h f sameo e total powe t wite bu rhth influenc e reference th insertioth e , r th 1 cor R f - Fo eo defines e EF 3 f differena no § n do t following modification: s amounts of M.A. on the main core performances is shown on Table III.

- core diameter decreased by about 20 %, correspondin o muct g h smaller fuel pellet diameter, wit same hth e core height,

- in correspondance, the fuel volume fraction is decreased by about 10 % relative, and the TABLE III. INFLUENCED? M.A. INSERTION ON THE MAIN CORE PERFORMANCES fuel enrichment increase abouy d b sam e th t e amount, M.A. content Ap (0)' Ap (BU)b Ap (rods)' Ap (Na)d Ap (Dop)' same th r e peafo - k burn-up fuee th ,l lifetime goes dow abouo nt efpd0 98 t . (pcm) (pcm) (%) (%) (%) Due to the increased fuel enrichment, the neutron spectrum is harder, but this has no Np232% 7 - 2300 + 1400 - 4 + 20 -20 influenc M.Ae th .n eo burnin gcaseo capabilitytw se correspondingth r Fo . insertioe th o t f no 5 % Np237 - 4800 + 3500 - 9 + 40 -40 2 % of Np237 and 2 % of Am241, the transmuted masses during the first fuel cycle in the core 10 % Np237 - 10000 + 6100 -17 + 50 -60 respectivelye ar same th er energfo , reactoe th fueye t th producea l rr outpufo d dan : t Am242% 1 - 2600 + 2300 -4 + 20 -20 5 % Am241 - 6200 + 5400 -10 + 40 -40 - for the reference core : 8.3 kg/TWh for Np237, 10 % Am241 - 11200 + 9300 -19 + 50 -70 9.0kg/TWh for Am241, highee th r r fo enrichmen - t kg/TW1 cor8. eNp237: r hfo . 9.1 kg/TWh for Am241. 2.5 % Np237 -5800 + 4300 - 10 + 40 -40 2.5 % Am241 transmutee th , So d masse M.Af so identicae .ar botn i l h core s; indeed core th e n i wit, e hth higher initial enrichment, the M.A. production coming from the plutonium isotopes is slightly higher, which compensate increasee sth d mass fissioned wit hardee hth r neutron spectrum. " Ap (0) : initial reactivity (fresh core), (BUp : reactivitA b ) y swing with bum-up, On the other hand, the fuel composition at equilibrium, after 5 to 6 fuel cycles, can be (rodsp : rodA c ) s efficiency, d ro slightly modified due to the different core characteristics, mainly because of the different fuel (Nap A : sodiu) m void reactivit t beginninya f lifego , enrichments equilibriue Th . m compositions afte fue6 r l cycle respectivele sar y: 1 Ap (Dop) : doppler effect at beginning of life. In this comparison fuee th ,l enrichmen same bees th ha tt ne coradjustege o t e y sucn di wa ha 5 CONSEQUENCES OF M A RECYCLING ON THE RADIOTOXICITY OF WASTES CD reactivit t beginninya f lifego f equilibriuo , instead en f do m cycle A management strateg produce) y baseAm a nuclea y n waste+ b do p r (N spar k seems Tabln o e e se III n effecte ,ca th e Asson relateinsertioe th o givea dt f no n amoun f Np23o t 7 importan o analyst t orden i e o compart r differene abilite th th e f o y t system o reduct s e th e or Am24I are very similar, and variations are almost linearly dependant on the imtal amount potential radiotoxicity sourc f wasteeo s The mam tendancies are as follows

- for a given plutonium enrichment, the initial core reactivity is significantly reduced, Np237 Am24d an 1 having large neutron captur presene e th rate s A st compariso mads ni t samea e 5 1 REFERENCE PARK initial reactivity, this would increase n leaa o dt dependin % 5 d1 enrichmeno t g % 5 y b t amountA M e onth , e referencTh e par always ki frence th s h one, consistin PWR'n go witd sfe h uranium oxide enrichment% 5 2 3 fue t generatind a l an , TWh0 g40 r yea epe r - the reactivity swing over the fuel lifetime is strongly decreased, due to the production of highly efficient isotopes, in terms of equivalent plutonium 239 In fact, if one compares f life o considee w fuee d , th en lt A r tha• t the different effects with the same reactivity at end of equilibrium cycle instead of beginnin f lifego , this would leadecreasa o dinitiat e th f elo core enrichment insteae th f do - the fuel is reprocessed, the separated plutonium being used once more in PWR or in FBR , increase quoted just before, only the Pu losses estimated to 0 3 % during reprocessing operations join the wastes, consideree ar ) Minoe Cm wastes th d a + l al r m Actmides A - f - p s(N the contro efficiencd ro l decreases i y abouy db 5 t to 15 %, due to the harder neutron spectrum,

- the sodium void reactivity for the fissile zone is strongly increased, by about 20 % to 5 2 MINOR ACTINIDES TRANSMUTATION 50 % depending on the initial M A amount, with correspondingly a reduction of the Doppler coefficient by about the same amount mixea w pare dno parTh s ki k composed wit proportioha f somno e standar (thR edPW sam e s previousla y described f somo d e an )whicamoun R EF hf o tmak transmutatioe eth n working essentialls Ii t y this later effect which limitâte amoune A tha sth tM coulf o t insertee db n di The total power (PW REFR+ ) settled same stayth et sa leve l generatio TWh0 40 r f epe no such an EFR core, homogeneously distributed over the whole core This limitation is of the fueR lEF amouncontaine yea% Th 5 r f Mino2 o t sa r Actinide Am)+ p s(N , homogeneously order of 2 % to 2 5 % of the total heavy metal mass of the fissile zone, which means about 0 7 distribute whole th n di e core. core correspondinR Th eEF e th n i A g increasM f o sodiun n eo to 1 mo t void reactivitf o s yi the order of 20 % relative, which would need anyway specific improvements on the core itself The self-recycling is simulated in the following manner • but such improvements seem possible. - core burnt in one fuel cycle of 1600 efpd, cooling time of 1 year, - reprocessing simulatio• nm * removing from the fuel all the fission products, 42 BURNING POTENTIACORR EF EN A F LO * adding a Np + Am mass to re-obtain the maximum allowed content, * completing with a U-Pu mixture to keep the total mass, and the correct reactivity level, (Np23A M mas e f o 7Am241 + abouf Th sn o to 1 t ) corresponds approximatel annuae th o yt l - a period of 1 year for fuel manufacturing, and so on till an equilibrium situation is production of the french PWR park (400 TWh annual), fed with uranium oxide For such a reached PWR park witd 5 planautn R an 1 ,h a f oEF o t recyclinw ne type a th e , f 2 gdescribe o - 3 § n do GWe every year woul necessare db y This fue,6 afteo lt cycles5 r , which corresponds roughly The needed number of EFR plants is determined so that the FBR park yearly consumption to the plant lifetime, would allow to reduce the total amount of M A (Np237 + Am241 + in Np + Am equals the PWR park yearly production Cm245) by about 80 % as compared to the production of the LWR's Anyway, this type of autorecyclm vert no y s gi realistic , sinc always i t ei same sth e fuel which feed reactore sth , with In this management strategy, wastes find their origin in onl ycomplemena f plutoniumo t . - those occurring when partitioning the Np and Am from the PWR and FBR fuels the A more realistic mod f recyclineo woulA maintai o t g M e db constanna t percentagA M f eo partitioning quality for Np and Am was taken equal to 90 %, in the fuel, by adding M A at each new fuel cycle to complete the amount to the level of 2 to Pu losses estimated to 0 3 % when reprocessing the PWR and FBR fuels, 2 5 % this is the case which is described in the next chapter - the total Cm produced by both PWR and FBR parks 5 3 RESULTS different hypothesis concerning the partitioning quality of M A (99 %) and by recycling Cm as Np and Am, we observe a reduction of the radiotoxicity on the entire time scale by a factor Wit previoue th h s hypothesis proportioe th , incineratoe th f no par R s foune b kwa FB ro t d of about 7 to 30, as shown in the following table and on Fig 3 about 20 % of the total settled power

comparisoe Th n betwee wastee nth s annua referenc e lincinerato e th mas th r d fo s ean r park 2 3 4 5 6 7 shows that Time (years) 10 10 10 10 10 10 Radiotoxicity reduction factor masp N reduces i se th , abouy - % db 0 8 t with a partitioning quality of 99 % - the Am mass is reduced by about 75 %, ) Am + EFp R(N 36 45 14 7 20 14 massm C e , th mainl - Cm244e yth increases i , factoa 4 y df b o r EFR (Np + Am -f Cm) 14 16 7 8 33 25 e consequenceTh termn i s f radiotoxicito s followine e giveth yar n i n g table, expressen i d term f radioioxicitso y reduction factor functioa s a s f timo n e Then the analysis of each isotope contribution shows that Pu represents 60 % to 80 % of the s Time (years) 102 103 104 10s 10" 107 whole radiotoxicity yeartilimprovo t aroun 10 d aftel o % san S r0 d5 e agai radiotoxicite nth y reduction factor necessars i t i , improvseparatioo ye t th morw u P e no f enA o thaM ne thath f to Radiotoxicity reduction factor 20 23 1 1 30 45 43

Note that a strategy of Np + Am incineration is able to reduce the wastes radiotoxicity as compare reference th o dt e strategy (fuel close whole th dn ecycleo tim facto5 a o y et )b scale2 r , excepted around 104 years where the gain is almost zero This is due to the Cm244 contnbution which represents about 60 % of the total radioxicity This higher contnbution is understandable large byth e mass generate transmutatiom A face y th dtb y thab t thitd a snan time elementars it , y radiotoxicity is highly superior by a factor of 100 as compared to the elementary radiotoxicity of the other Minor Actimdes

More generally analysie th , f eacso h isotope contnbutio whole th n eno radiotoxicity shows that

- at short times (102 to 103 years), the Am241 and the Cm244 represent 80 % of the total radiotoxicity, \\ years* Cm24e 10 t th , a highes - 4i r, (abou%) 0 6 t EFR __iv\.»_ EFR ) Am + p (N v V ) •»•m Cm (NA p+ - at 10s years, the Am combined with the Np amounts to 60 % of the entire radiotoxicity, but wit hnoticeabla e contributio Pu23e , th r abouf % n9o fo 5 2 t - for very long times (106 to 107 years), the Np predominance (about 50 %) combined with the Am241 (about 35 %) is settled up (note • the radiotoxicity contnbution is associated to the original isotope)

These majoe resultth e r sar contributor shoA w wholM e thae th th o tset toxicit wastesf yo , as compared to plutonium and also to fission products for long times tog (yeafs) ro To reduce this toxicity, the Np and Am partitioning quality has to be improved and the Cm

CD whose contributio s importanni 4 year10 t als t a t tshorbu s o a recyclee b to t tim s deha Witha FIG 3 Evolution radtotoxicit) ofthe functiona as timeof Gi 6 CONCLUSION PARTITIONING AND TRANSMUTATION RESEARCH O AND DEVELOPMENT PROGRAM (OMEGA JAPAN I ) N These preliminary studies hav eFasa showf o t Neutroe nus thae th tn ReactoR r sucEF s ha recyclo t Minoe th e r Actimde reprocessine s th comin n spenf R a o e t b LW t g ou f fue n go ca l intermediate solution to reduce the long term radiotoxicity of long lived wastes However, a T MUKAIYAMA significant reductio radiotoxicite th f no y needs also improvement reprocessine th n so g process, Japan Atomic Energy Research Institute, mainly concernin partitionine gth g qualit r botyfo h Minor Actimde plutoniud san m Tokai-mura, Naka-gun, Ibaraki-ken, Japan As regard coree lemsertiose th th , f Minono r Actimdes insid somfues e eth ha l e benefitn so core th e physics examplr fo , estrona g reductio reactivite th f no y swing alst bu o, drawbackn o s Abstract the main safety parameter sn increase a suc s a h d sodium void coefficieno thest e e Du t drawbacks, the Minor Actimdes amount that can be inserted has to be limited to about 2 to In the Japan Atomic Energy Industry Forum report on the waste 2 5 % of the total heavy metal mass, with as a consequence a limited transmutation capability management of long-lived nuclides the importance of research and the percentage of FBR in the nuclear park must be of the order of 20 %, in order to compensate developmen partitioninr fo t d transmutatioan g n (P-T f long-live)o d nuclides the LWR production in Minor Actimdes was pointed out as long term efforts in developing a complete system for radioactive waste management. Based on this, the interested Japanese In a further step, it is foreseen to study an heterogeneous mode of recycling the M A m a organizations propose o initiatt e d Th a majoeT D prograP- R& r n o m FBR, in specific target subassemblies located either in central core positions or in peripheral development of P-T technology was deemed to be quite an interesting subject for ongoing investigation froe perspectiveth m f potentiao s l utilizatiof o n positions Indeed, this would simplif manufactunne th y g proces separatiny b s e plutoniugth m resources and possible long term advances in radioactive waste management subassemblies from these high activity targetsburnine th t bu ,g potentia influence wels th a l s a l e of these specific subassemblies has to be established The proposed project was called "OMEGA". The Japanese Government proposed an international cooperation for information exchange to cover the areas of nuclear physics, reactor physics, advanced technologies and REFERENCE physico-chemistr n additionI y , advancemen f technologieo t s suc s lasea h r and accelerator technology, will provide spinoff r othefo s r fieldf o s science and technology [1] - SZTARK, H , BARNES, D , WEHMANN, U , Core optimization of the European Fast Reactor EFR, International Conference on Fast Reactors and Related Fuel Cycles, October 28 Novembe , 19911 r , Kyoto1 1- 3 , 1 Japa l nVo 1 INTRODUCTION

In 1973, the Japan Atomic Energy Industry Forum published the report on the waste management of long lived nuclldes after the two years assessment studies by the groups of scientists and engineers in Japan In that report entitled ' A closed system for radioactivity", the importance of research and development for partitioning and transmutaùon(P-T) of long-lived nuclides was pointed out as long term efforts m developing a complete system for radioactive waste management

Even after the pessimistic conclusions for P T of high level radioactive waste(HLW) drawn by Oak Ridge National Laboratory(1980), IAEA(1982), etc, small number of groups in Japan, as well as Europe, the United States etc continued their studies m P T as basic research

Based on their studies, the interested Japanese organizations proposed to initiate a major R&D program on P T The development of P T technology was deemed to be quite an interesting subject for ongoing investigation from the perspectives of potential utilization of resources and possible long term advances in radioactive waste management

In 1987, Japan's Atomic Energy Commission(AEC) concluded that the potential benefits from the use of some elements in fission products and from recycling minor actimdes for power generation could be achieved provided that a well planned, efficient and effective R&D program Chemica- l propertie behaviod san actinidf ro e specie aqueoun si organid san c solutions could be formulated. The AEC then submitted in October 1988 a report entitled "Long Term - Analytical techniques and methods Progra Researcr mfo Developmend han Nuclidn to e Partitionin Transmutatiod gan n Technology", - Physical and chemical properties of actinide compound which plots a course for technological development up to the year 2,000. Collectio- evaluatiod nan nucleaf no thermodynamid ran c data The progra mcalles i d "OMEGA" whic acronye th hs i m derived from Options Making Extra Gains from Actinides and fission products. PARTITIONIN2 2. G programD TheR& s were jointly stimulate collaborative th y db e effortJapae th f no s Atomic Energy Research Institute(JAERI Powed )an r Reacto Nuclead ran r Fuel Development Corporation Advanced separation technolog minof yo r actinide fissiod san n product developee b o t s i s d (PNC). In the public sector, the Central Research Institute of Electric Power Industry(CRJOEPI) basie on th currenf so t trendchemicae th n si l separation processes processey dr d . an Bote t sar hwe also has been carrying out R&D on this subject. to be developed. This would include application of new extracting solvents, laser induced separation.and sublimatio volatilizatiod nan n processes. Studie r utilizatiofo s f separateno d The Japanese government (represente Sciency db Technologd ean y Agency; STA) proposed nuclides will als includede ob followine .Th studiede gb itemo t e ;sar internationan a l cooperation for information exchang coveo et areae rth nucleaf so r physics, seactor physics, advanced process technologies and physico-chemical characterization relevant to P-T Partitionin- W HL r gfo technology under the framework of the OECD Nuclear Energy Agency4n January 1989. The first - Partitioning in the reprocessing process information exchange meeting on this subject was held in Japan in November 1990. Eleven OECD - Recovery of platinum group metals member countries and two international organizations, namely IAEA and Joint Research Center of - Cost estimatio procesa f no s CEC, participate meetinge th n di . 2.3 TRANSMUTATION The program is conceived as a research effort to pursue benefits for future generations through long term basic R&D, ana is not to seek a short term alternative for established or planned Recently new approaches such as minor actinide transmutation using actinide burner fuel cycle back-end policies.The program is expected to serve to revitalize nuclear R&D in general, reactors, metal fuel FBRs(LMR) with dry reprocessing, intense beam proton accelerator as and also to attract capable young researchers dedicated to bringing the nuclear option into the 21st neutron source have been studied. Optimization of minor actinide recycling into Pu-LWR and centur healtha n yi y state additionn .I , advancemen technologief to s suc lases hacceleratoa d ran r MOX-FBR is to be studied. Studies of transmutation of Sr-90, Cs-137 with an electron accelerator technology, as advocated in this program, will provide spinoffs for other fields of science and are also proposed.Field studiefollowingse e b th o st e dar ; technology. - Nuclear data and fuel property data of minor actinides Developmen- computef to r code . OMEG2 A PROGRA JAPAMN I N Syste- m design study - Reactor fuel and accelerator target development The program is to be proceeded in two steps: the phase-I and II. The phase-I covers a - Fabrication technology of the fuel and target materials period up to about 1996, and the phase-H covers a period from about 1997 to about 2000. In - Development of high power accelerator for transmutation general basie th , c studieconductee b testind o t phase-e san e th gar n d i evaluato t I e various Cos- t estimatio R&Df no , construction fued ,lan cycle concepts and to develop required technologies. In the phase-II, engineering tests of technologies The areas covered by the program and R&D activities are illustrated in Fig.l. demonstratior o conceptf no plannede sar . After 2000, pilot facilities wil buile b l demonstrato t e technologythT eP- . prograe Th m cover followine sth g fields. 3. RESEARCH AND DEVELOPMENT ACTIVITIES

2.1 PHYSICAL AND CHEMICAL PROPERTIES OF MINOR ACTINIDES Unde framewore rth OMEGe th f ko A program followine ,th activitieD gR& undee sy ar rwa AND FISSION PRODUCTS at JAERI, PNC and CRIEPI.

Reliable data base of minor actinides and long-lived fission products is indispensable to proceed the program. Underlying studies will improve understanding of the science and 3.1 PHYSICAL AND CHEMICAL PROPERTIES OF MINOR ACTINIDES technology for separation and recovery of these nuclides from HLW, for fabrication of actiniae AND FISSION PRODUCTS fuels for recycling to reactors or accelerator driven systems for transmutation, and for utilization of these nuclides. Nuclear datthermodynamid aan c dat thesf ao e nuclide measurede sar , compiled P-Tr fo followine ,th D basie R& th f gs o studieA s wer fiele e th f carrie do n i t dou and evaluate reactor dfo r physic materiald san s development. Followin studiede gb itemo t e ;sar chemistr nuclead yan r dat JAERIt aa . GO a) Destructive analysis ofPWR spent fltelfl]: ro spenR tPW fuels with burnup betwee34.d an 1 GWd/MT9 n6. U were analyze masy db s spectrometry, alpha-ray and gamma-ray spectrometry. The results were compared with the calculation using nuclear data library JENDL- bumuo tw d p2 an codes , ORIGEN- JAERTd 2an s SRAC-FPGS. The intercomparison of analysis of nuclides generated from minor actinide isotopes irradiated in a fast reactor is to be carried out under the cooperative program with ORNL.

b) Flow-coulometry ofUflpfu ions(2J: oxidation-reductioe Th ionu P acidi n si d nan c behaviop aqueouN , U f ro s solutions swa studie flow-coulometry b d y with multi-step column electrodes e rapiTh d. preparatiod nan determination of the ion of desired oxidation state were developed.

c) Photo-acoustic spectroscopy induced by Fourier transformed laser beam[3]: 31 Fourier transform laser-induced photo-acoustic spectroscopy (FT-LPAS s beeha ) n OQ develope speciatior dfo actinidf no e element aqueoun si s solutions ultravioletd an , , visibld ean near-infrared photo- acoustic spectroscopy (UV-VIS-NIR PAS) and Fourier transform infrared photo-acoustic spectroscopy (FT-IR PAS thosr )fo solin ei d phases. These methods were applied speciatioe th o t f U(VIno NaHCOj/NaC/Qn )i j solution/precipitate systems with satisfactory result These methods are also applicable to speciation of solid phases, especially of amorphous materials.

d) Electrolytic deposition ofactinides in molten salt: electrolytie Th c depositio separationU studTR r yfo n using chloride salt undes s i wit y hrwa emphasi reaction so n kinetic equilibriud san m measurements.

e) Nuclear data: latese Th t versio Japanesf no e Evaluated Nuclear Data Library 'JENDL-3'[4] containe sth nuclear dat actinidef ao Fm-255o t p su . Sinc accuraciee eth thesf so e nuclide dat uncertaine aar e th , revision of these data are planned under NEACRP/NEANDC International Evaluation Cooperation date f Np-23.Th ao Am-24d 7an beine 1ar g reevaluated usin available gth e datf ao integral experiments as the first step of the cooperation. For neutrons which are emitted in the charged particle induced reactions, evaluation of the nuclear data for neutron energy between 20MeV and SOMeV are under way. For neutron above lOOMeV, validity of various theoretical calculation codes are examined using the experimental data. For the future evaluation high energy nuclear data, the neutron emission data are being collected both for thin and thick target experiments. c > B) > r B* 2. a « 1 PARTITIONIN2 3. G iä PO a R 8 <—i R Wet processes of partitioning are developed at JAERI and PNC. Dry processes are 1 i 1H 1 develope CRIEPt da PNCd Ian . i/ A. Studie JAERt sa I

a) Three group partitioning costand estimation: From 197 19843o t partitionin,a g proces developes swa separatinr dfo g elementW HL n si into three groups; TRU, Sr-Cs and others. The process consists of three steps; the first is solvent extraction of U and Pu with tnbutylphosphate (TBP) and the second is solvent extraction of Am and Cm with diisodecylphosphonc acid (DIDPA), and the third is adsorptions of Sr and Cs with B Studies at PNC inorganic ion exchangers The process was demonstrated by using actual HLW solution More werm C e d extractean m A f do wit% h99 tha DIDPA[59 n9 preliminare Th ] y assessment study B l Wet Process indicated that total volume of solid material is reduced to less than half of that of vitnfied waste, a) Extraction tests based on the "TRUEX"flow sheet f whico % s fissiohi 12wt n products oxid thad cos e r constructioan th t fo , et fromHLW d nan Extraction tests have been conducted to recover minor actimdes from HLW, based on the operation of a partitioning plant is less than 5% of that of a Purex plant, provided that a TRUEX flow sheet Fundamental batchwise tests were earne determino t t dou distributioe eth n partitioning plant is operated in connection with the Purex plant[6] ratios of minor actimdes and fission products Based on the above results, counter-current flow sheet tests with small scale mixer-settlers were also conducted usin deriveactuae gW th lHL d from b) Development of four group partitioning fueR lFB reprocessing tests[13] Future studie plannee sar improvo dt procese eth r backsfo - After 1985 foua , r group partitioning proces bees sha n develope whicn di hstea r pfo extraction of Pu, to develop a separation process of minor actirudes/rare-earth elements, to separating Tc-platinum group was developed in addition to the three group separauoa An effective improve extraction efficiency of Np, to clarify the conditions leading to third phase formation and method for separating TRU, especially Np, and Tc, was developed[7] soon

c) Separation ofNp b) Fuel cycle including minor actimdes separation process Experiments on a counter-current continuous extraction using a mixer-settler were earned Fuel cycle including minor actimdes separation process is going to be designed as follows, out to determine the conditions for extracting Np(V) with DIDPA The addition of hydrogen spent fuels from both LWRs and LMFBRs will be reprocessed by the advanced PUREX process, peroxide accelerates the extraction of Np More than 99% of Np was extracted when hydrogen in which minor actimde separatee b n sca d from other element fuelX s MO includins g minor peroxide was feeded so as to compensate for its decomposition[8]. actmides will be loaded to LMFBR and minor actmides can be reduced in the core effece Th f radiolysio t f DIDPso extraction Ao Np(Vf no simulates )wa irradiatiny db g DIDPA with Co-60 y -ray Little change in extraction rates of Np was observed when the solvent c) Recovery of noble metals from insoluble residue was irradiated with the dose up to 1 MGy[9] Recovery tests have been conducte extractiob P y db n proces separato st e noble metals from actual insoluble residue produce dissolutioy db f spennconfirmes o wa t t I fueld that noble d) Separation of Am and Cm from rare earths Rare earths as well as minor actmides are extracted wtth DIDPA Recently it was found that metals, such as Ru, Pd and Rh, can be separated efficiently by this process Mutual separation of preferentialle ar m C Ad man y back-extracte diethylenetnaminepentaacetiy db c acicomplexina s da g noble metals will be conducted agen aqueoun i t s phase while rare earths remai DIDPe th n i A solven rarf o thad el tearthan al t s remaininu P d solvene an th p back-extractegsucn e i N U an tar s hda TR d wit nitriM h3 n ca acid dan Procesy Dr B2 s oxalic acid solution, respectively a) Separation byfluoride volatilization process Vapor pressure measurement fluondef so basid san c separation progresn i test e a sar s sa e) Separation ofTc feasibility study of fluonde volatilization process Two method bees ha s n develope separato dt , precipitatioeTc deratratmy d nb an W gHL adsorption with active carbo s recoveresimulatea nwa n i c MorW T s df a HL o e tha% n95 b) Ultra high temperature separation precipitate by derutrauon of HLW by adding formic acid to make the solution pH aoove 2 0 The separatee b n ca extremelt heatinr a y dS b W d gan HL s y C high temperatures( 1800t:-2500 active th f eo carboe us n column resulte quantitative th n di e adsorption frooc fT msimulatea W dHL by adjusting the solution to a 0 5M nitric acid and the elunon of Tc from the column was achieved t;) Basic tests have been conducted using a simulated HLW It was shown that Cs can be easily quantitatively by the use of an alkaline thiocyanate solution as eluant[10] separate residue th d stabls edi an compac d ean t f) Demonstration of four group partitioning The four group partitioning process is to be tested with actual HLW at NUCEF (Nuclear StudieC t CREEPsa ! Fuel Cycle Safety Engineering Research Faculty) whic undes hi r constructio colA n d operatiof no the facility is scheduled in 1993 a) Pyrometallurgical process[14} Separation of TRU from HLW by pyrometallurgical processing and transmutation in a g) Integration of partitioning process into Purex process[12] commercial metalli thin I s c scheme R fueconfinee cycl ar R produceFB l U R FB e ,TR n di LW n di To integrate these partitioning processes into Purex process, a reductive stripping with iso This process consist followine th f so g four steps Both denitratio chlonnaùod nan ne steppr e sar normal-butylaldehydd an successfulls ewa y applie selecnvr dfo frou eP e mseparatioth d an p N f no processin partitioninr gfo g codecontammation product streamfl 1] A study to develop a system for automatic measurement in (1) Derutranon of HLW by microwave heating This method is preferable for converting HLW to extractioU TR n proces s commenceswa 199n di 0 operativo undec e th r e program witA hAE oxide becaus smale th lf eo amoun f secondaro t y safete wasteth operation d yi san n Technical Harwell To support these works, a TRU valency monitor based on the photoacoustic feasibility of this method was examined expenmentaly Based on the tests, a design study was CO CO spectroscopy is under development earned out for the derutrauon step CO Chlonnatio) (2 oxidee th f chloridesno o st . Chlorin s togetheega r with carbo s user nwa dfo fue less i l s than mino5wte Th % r actinides transmutation cycl r amounrate pe abou s eTh i e % 11 t -P" chlonnatio oxidee th f no , where carbon serve reductans sa chlonnauor fo t ne b Chlorin n ca s ega of minor actinides transmute s aboudi timex si t f tha o sf mino o t r acumdes generatea n di recycled Nearly 100% chlonnaüo oxid demonstrateW s ewa HL f no d I.OOOMWe LWR Minor actmide loading of 5wt% reduces the burnup reactivity loss by about (3) Reductive extraction to separate TRU from molten chlorides using liquid Cd-Li TRU with 50% some amount of lantharades are separated from molten chlorides by reductive extraction using liquid Cd-L servei reducin e solvenL e ith th s ss recovera r a gtfo d agenC d reducef yo tan d metals b) Minor actinides material balance In order to evaluate the punty of TRU recovered in the reductive extraction step, it is necessary to To estimate the effect of transmutation in LMFBRs, the total material balance of minor obtain the distribution coefficients between salt and Cd phases, and related thermodynamic data, actinides in Japan was calculated The nuclear generating capacity was estimated based on the especially activity coefficients for TRU and lantharudes The distribution coefficients, defined as a Long-Term Progra Developmenr mfo Utilizatiod an t Nucleaf no r Energ Japan(1987n yi e th d )an mole fractio phasd C n ei n divide moly db e fractio saln i n t phase e beinar , g measurer dfo introduction of the LMFBR on a commercial scale was supposed to start from 2020 This study irradiatiny b , acnmdesd [15]Y Np an d ,d usingU an an d U lantharude,g N , Pr , Ce s, sucLa s ha implies that minor acumdes transmutatio LMFBRn i s suppresse quantitiee sth storef so d minor Kyoto University Research Reactor. acumdes after reprocessing (4) Electrorefining of recovered TRU The electrorefirung with liquid Cd anode from the preceding reductive extraction step, solid cathode and molten salt electrolyte (KCl-LiCl eutecac c) Irradiation test JOYOin developmentand of fuel salt bees )ha n applie anode Th d e dissolutio cathodd nan e deposition study using actmidd ean To evaluate neutron cross section data, small amounts of minor acumdes will be irradiated in lantharude mixture will start from the end of this year in the TRUMP-S Program[16] The the experimenta "JOYR lFB O ' Irradiation test fuef so l pins containin smalga l amoun Am-24f to 1 developmen procesf to s technology includin reductive gth e extractio electrorefirund nan g stepe sar wil conductee b l followinge Th diteme th r futur e sfo ar s e studies fuel fabrication technology, to be started m 1992 physical and mechanical properties of minor acumdes loaded fuel, and additional irradiation tests

b) Measurement of activity coefficient B Studie transmutation so metallin i t CRIEPa c R fueFB lI The activity coefficients and free energy for formation of chlorides in salt and activity coefficient metalf havo sd , C en Sm s i bee , Nd n , obtainemeasurementF Pr , EM Ce y d, b La r sfo Transmutatio recovereU TR f no pyrometallurgicay db l proces studies si e th n di Eu, Gd and Y and for U, Np, Pu and Am by the cooperation with Rockwell International Corp , followin witgR areahFB metallif so c fuel Universit Missouri-Columbiaf yo Kawasakd an E DO i HeavS ,U y Industry unde TRUMP-e rth S Program starte n 1989[16,17di e datafforn Th aca ]mak o t dmodeline th e g processer fo s a) Calculation of transmutation rate and design study of fuel element and core separating lantharudes and actirudes CITATION-TRU code was developed by modifying the bumup calculation code CITATION. When 5,1 r 15wio 0 minof %o r actmide mixee sar d fuelwitR h,FB transmutation A series of the experiments starting 1992 is planning to develop the process technology rate per year is 11 to 12wt% in oxide fuel, and 13 to 14wt% in metal fuel One unit of lOOOMWe FBR with metallic fuel which contains 5wt% of minor actinides can transmute minor acumdes through demtration step to electrorefimng step with the simulated waste This separation fro munit5 f lOOOMWso e LWR[19] technolog alst solvenr ybu o wilfo appliee W b l t onlt HL scrubbin r dno y fo g liquid wast Puren ei x Based on the chemical analysis of minor actinides oxide samples irradiated m process, undissolved residues and hulls after dissolution by mtnc acid KNK-2 fast reacto bence th r h mark calculation transmutatior sfo n rat beine ear g earnen i t dou cooperation wit CRIEPd han KfKU IIT , 3 3 TRANSMUTATION WITH FISSION REACTORS b) Characterization study ofUPuZr alloy with minor actmide and irradiation study Optimizatio minof no r actmide recycling into MOX-FB Rstudies StudieC i PN t df a s o In orde evaluato t r characteristice th e f metaso l fuel with minor actinidese th , minor acurude recycling into metallic fuel FBR are conducted at CRIEPI Actmide burner reactor following item beine sar g measure cooperatioe th n di n with ITU[20], design study and the integral experiments to improve cross section data of minor actinides are mutual irascibilit actinidesf yo , carried out at JAERI allowabl- e conten minof to r actmide lanthanided san fueln si , - other fuel properties 3 3 I MOX-FBR and metallic fuel FBR Fabrication and irradiation studies of metallic fuel with minor actmide are in preparation Studies have been proceeding in cooperation with ITU A Studies on transmutation m MOX LMFBRs at PNC 2 Actimd33 e Burner Reactor studie JAERt sa I Characteristicsa) of minor actinides transmutation! 18] Parameter survey calculations have been performe investigato dt e basic characteristicf so a)Actmtde Burner Reactor(ABR) design study[2132] minor actinides transmutation in a 1 ,OOOMWe LMFBR The homogeneous loading of minor ABR design study was earned out to obtain a technically feasible ABR model An ABR is a actinides has no serious penalties to the reactor core performance, provided that minor actinides in fast reactor designe minom bu ro d t actimdes(MA ) efficientl majos It y r componen f fue to minos i l r actirudes and neutron energy spectrum is very hard to fission MA nuchdes which have fission b) Experiments for spoliation m a heavy metal bulk target[27] thresholds at neutron energy of about 700keV Two types of ABR design were obtained. The integral experiment with a lead bulky target and 500 MeV protons from the booster Na cooled MA metallic fuel ABR(M-ABR) The advantages of metallic fuel are the hard facilit Institute th t ya Higf eo h Energy Physics.KEK startes ,wa 199n di evaluat0o t reliabilite eth y neutron spectrum and the compact fuel cycle facilities when pyrochemical reprocessing is used. of the code NMTC/JAERI. The neutron energy spectrum, spatial distribution and the yields of He cooled nitrideA M particle fuel ABR(P-ABR) particlA e fue l reacto s verrha y high spallation products were measure futuren I d experiment e th , s usin gtungstea depletea d nan d power density since heat removal in a core is very efficient because of a large heat transfer surface uranium target plannede sar . per volume of particle fuel. Coated fuel particles are directly cooled with helium A particle fuel consists of acnmde nitnde nucrospheres and thin TiN coating When fuel operating temperature is c) Design studies of minor acttnide transmutation system[28] lower thathirde n on theif so r melting points fuee th ,l would perfor so-callea s ma d "cold fuel" The conceptual design study for an accelerator driven transmutation system has been where the atomic migration is suppressed to result in a low swelling and a little fission-gas release performe conceptuae dTh l design stud systea f yo chloridmf o e molten salt cor bees eha n starteo dt transmute minor actirades and long-lived fission products b) Study on ABR fuel and fuel cycle [23] thermophysicae Th l dat metalaA basM insufficiens f i seo t Phase diagrad man d) Development of an intense proton accelerator[29J thermodynamic alloystudieA startee M sar f so cooperation di n with ORNL Fabncabihtd yan The construction of Engineering Test Accelerator (ETA) with a proton energy of 1 5 GeV performance particularly during transients of particle fuel are the areas of future study. Forced and a current of 10 mA is planned Various engineering tests will be performed using this cooling of fuel, even of metal ingots, is necessary because of large alpha decay heat. Applicability accelerator for an accelerator-driven transmutation system ETA is a significantly large system In of pyrochemical processe nitride th o st estudiee b particl o t t d eye Wit fues ha hsuitabla l e particular, an average proton current of 10 mA is 10 -50 times larger than that of existing chlorinating agent such as ZnC/ 2, actuude nitrides could be selectively incorporated into a fused accelerators used for nuclear physics experiments To build ETA, therefore, considerable efforts chloride, whil coatinN eTi g fragmen lefs i t t unreacted. and resource requiree sar d As a first step, Basic Technology Accelerator (BTA) with a proton energy of 10 MeV and a c) Integral experiments for minor actiniae cross section evaluation [2425] current of 10 mA is to be built to study a low energy portion of ETA A study of high energy The integral experiment was earned out in the fast critical facility (FCA) of JAERI to portio acceleratoe th f no r (hig structure? h/ alss i ) ot mode planneho a ls poweF testA dR a , r evaluate and modify minor actmide cross sections Fission rate ratios and small sample reactivity source will be built and the electric magnetic characteristics of ETA will be measured worths of separated isotope samples were measured in seven cores where neutron energy spectra Various computer design codes are to be checked for their reliability The conceptual design were systematically shifted. Usin measuree gth d data, JENDL-2 cross section data were adjusted. stud varioud yan s optimization (trade-off )performanc e studth r yfo e parameter beins i A g ET f so earned out.

3 4 TRANSMUTATION WITH ACCELERATORS B Studies of transmutation by use of accelerators at PNC

A transmutation system driven by a proton accelerator is studied and the development of an a) Assessment study of accelerator assisted system [30] intense proton accelerator is under way at JAERI At PNC, accelerator assisted or fusion.driven transmutatioe Th n rate energd san y balanc Cs-13f eo 7 transmutatio acceleratoy nb r assisted system Sr-9r Cs-13sfo d 0an 7 transmutatio evaluatee n ar intens n a d dan e electron acceleratos ri systems have been calculate usiny db g Monte Carlo codes t seemI . s difficul achievo t t e required under development transmutation rate energd san y balanc sucy eb h reactions propose s photonucleada r reaction, spallation reaction, neutron reaction by secondary neutrons in spallation, and (n,2n) reaction by A Studies of a transmutation system driven by an intense proton accelerator at JAERI neutron from muon catalyzed fusion. Further efforts will be made to select the most efficient transmutation method, includin methodw gne smovine sucth s ha g target mertiamethoe th d dlan a) Development of simulation codes spoliationa for process[26] fusion transmutation method. The simulation code NMTC of ORNL for high energy nuclear reaction processes in the energy range from 15 MeV to 3 GeV was modified as NMTC/JAERI to calculate high energy b) Development intenseofan electron linear accelerator fission reaction above ISMeV and spallauon reaction of minor actinides The simulation code A high current electron linear accelerator wit electron ha n energ f lOMeyo curren a d Van f to 'NUCLEUS' for spallauon reaction of one nucleus was developed to evaluate a computational 20mA is under development PNC selected an electron accelerator for development of an intense mode spallauor lfo n reactio analyzd nan date eth a measure thin di n foil experiments These codes accelerator because they are widely used in the industries and the basic technology that will be wil improvee b l includo dt e pre-equihbnum neutron emissio fragmentatiod nan n procesn si obtained during the development is applicable to any type of accelerators With this energy and spallauon reactio mtercompansoe Th n routinee th f nr higo sfo h energy fission calculatiof no current, the accelerator under development is fairy larger than existing electron linear accelerators NMTC/JAER HETC/KFd an I s Aplannei e codTh de SPCHAI s beeNha n developer fo d A basic design study of the accelerator has been completed and trial fabrication of main calculating generation and decay of spallauon products To analyze nuclear reactions in the entire components acceleratinn suca s ha g klystroa tubd progresn ei an s ni s Reflectin resulte gth s CO energy region lower than 3 GeV, two code systems were developed by combining NMTC/JAERI obtained durin developmene gth f theso t e components desige th , n wil modifiee b l e firsTh dt CJ1 wit neutroe hth n transport codes wite ,on h MORSE-D othee th d r Dwitan h TWOTRAN-II operatio schedules ni 199dm 5 CO REFERENCES (as of October, 1991) [22] TMukaiyama,e , Minotal r acumde transmutation using minor actmide burner reactors, en Pro Cont cIn Fasn fo t Reactor related san d Fuel Cycle1 (Kyoto9 R F s , 1991) [I] YNakahara, et al, Amount of Nuchdes Constituting PWR Spent Fuels , Radiochimica [23] T Ogawa,et al,' Fuel Elements and Fuel Cycle Concepts of Actinide Burner Reactors ', Acta.50, 141-149(1990) JAERI-M 89 123(1989) YoshidaZ ] , Floal [2 t we , coulometr Uraniumf yo , Neptuniu Plutoniud man m Ions with [24] T Mukaiyama,et al,'Acumde integral measurements m FCA assemblies", Proc InL Conf Multi-step Column Electrodes", Pro t AnaIn c l Sei'91 (Makuhan, 1991) publishee b o t , d Nucl Dat Basir aAppliefo d can i (SantdSe a Fe,1985),Gordo Breacd nan h Pubhshers,438 [3] TKimura, et al ."Speciaüon of Uranium in Aqueous Solutions and in Precipitates by (1985) Photoacousuc Spectroscopy '.Radiochimica Actapublishee b o t , d [25] S Okajima,et al, 'Evaluation and adjustment of actmide cross sections using integral data [4] K.Shibata,et al,' Japanese Evaluated Nuclear Data Library, Version 3 JENDL-3 - VAERI measured at FCA ,Proc Int Conf Nucl Data for Sei and Techn (Mito,1988) 1319(1990) [26] TNishida et al ."Calculation Code System for Fission and Spallauon Products ', Proc [5] M Kubota, I Yamaguchi, YMonta, K Nakâno, H Nakamura, Proc Mat-Res Soc Symp , ICANS-XI(Tukuba, 1990) , 551(198426 ) [27 TakadH ] ."Integral a t ae l Experimen Lean to d Bulk System Bombarded with High Energy [6] M Kubota and H Nakamura, JAERI-M 85-066(1985) Protons", Proc Specialist Meetin Acceleraton go r Driven Transmutation Technologr yfo [7] M Kubota, S Dojin, I Yamaguchi, YMonta, I Yamagishi, TKobayashi and S Tam, High Radwastes and Other Application (Stockholm, 1991), to be published Level Radioactive Waste and Spent Fuel Management", Vol 11, 537, The American Society [28] H Takada et al ,"A Concept of Actmide Transmutation with Intense Proton Accelerator", of Mechanical Engmeers(1989) Proc 6th Int Conf Emerging Nucl Systems (Monterey, 1991), to be published [8] YMonta KubotaM , TamS , , Proc RECOD , 348(1991 1 '91 ,W ) [29MizumotM ] ."Intensl a t oe e Proton Accelerator Development Program", Specialist Meeting [9] YMonta, M Kubota, Solv Extra Ion Exch , 8(2), 283(1990) on Accelerator-Driven Transmutation Technolog Radwastr yfo Othed ean r Applications [10] I Yamagishi, M Kubota , J Nucl Sei Tech , 27(8) 743(1990) (Stockholm, 1991) publishee b o t , d [II] G Uchiyama, S Fujine, S Hotoku, T Kihara and M Maeda,"Study for Reprocessing [30] T Käse et al ."Transmutation of Cs-137 by accelerator.iW Improving Separation Efficienc Np"f yo , Proc RECOD'91 , 723(19912 l \b , ) [12] TKihara FujineS , , TMatsui, TKitamura MaedaanM , Sakagarm,"ApplicatiodM f no Laser Induced Photoacousuc Spectroscopy Syste Determinatioe th mo t Transuramf no c Elements", Int. Solvent Extraction Conf 1990 (ISE) C91 [13] TKawata et al ."Preliminary Study on the Partitioning of Transuranium Elements in High Level Liquid Waste", First OECD/NEA Information Exchange Meeting on Separation and Transmutation of Acnrndes and Fission Products, (Mito,1990) [14] TInoue, M Sakata, H Miyashiro, TMatsumura, K Sasahara and N Yoshüa, "Developmen partitioninf to transmutatiod gan n technolog lonr yfo g lived nuchde, s Nucl Technol 93 ,206 (1991) Sakata[15M ] Kurata,M , HJHijikat InoueT d aan , Equihbnum distributio rarf no e earth elements between molte LiC- 1 l neutecnKC c liquisald tan d cadmium "J Nucl Mater 185, 56 (1991) [16] J J Roy, L FGrantham, L Macoy, C L Krueger, TStorvick, T Inoue, H Miyashiro an Takahashi,"StandardN d potentia lantharudf lo actmidd ean e tnchlonde molten si n eutecüc LiCl - KC1 electrolyte" Mater Sei Forum 73 547(1991) [17] M Sakata, H Miyashiro and T Inoue, ' Development of pyrometallurgical partitioning of transuranium elements pno transmutatioo rt metallin i c fuel FBR", Int. Conn o f Fast Reactors and Related Fuel Cycles(Kyoto, 1991) [18] M Yamaoka et al ,"Charactensùcs of TRU Transmutation in a LMFBR", First OECD/NEA Information Exchange Meeting on Separation and Transmutation of Actimdes and Fission Products (Mito,1990) [19] A Sasahara and TMatsumura,' Transmutation of transuranium nuclides with FBR", ProConL In cf Nuclear Data fo<- Scienc Technology(Mitod ean , 1988) (19885 ,95 ) [20] TInoue, M Kurata, L Koch, J C Spirlet, C T Walker, C San,"Charactenzation of alloy fuels with minor actimde presentee b Winte S o t , sAN rn di Meeting (1991) [21] TMukaiyama,et al ."Higher actmides transmutation using higher actirude burner reactor Pro t ConIn c f Physic f Reactorso s Operation.Desig Computatiod nan n (Marseille 1990), \61I97 STATU MINOF SO R ACTINIDE TRANSMUTATION 2 NEUTRONIC SAFETY CALCULATION STUDY AT CRIEPI 1 CALCULATIO 2 N GEOMETRY

A reactor type is the lOOOMVie metallic fuel FBR core designed in A SASAHARA, M KURATA CRIEPI*3 Figur show1 core e2 sth e geometry desig e cor Th s composeei n d Central Research Institute of Electric Power Industry, of two homogeneous regions Inner and outer core region Refuel interval E werR ed a fue i an homogeneousl yea1 s loade d year3 i lA M an rr sfo d y Tokyo, Japan loade n inne i dd oute an r r core regions Reference designed fues U-Puwa l - Zr ternary alloy which contains 10* of Zr We considered to exchange a part of U and Pu in the ternary alloy for MA and RE Loading ratio of MA Abstract and RE was 2% and 5% of heavy metal (excluding Zr) in loading weight The study of the partitioning and transmutation of minor-actinides (MA) arf greao e t valu n futuro e e nuclear fuel cycle havW ee been studying E 2 R COMPOSITIO 2 D AN A M F NO neutronic characteristics of MA transmutation in a metallic fuel FBR and material characteristics of MA contained fuel In the present study, we The composition of MA and RE is that in equilibrium recycle analyze e reactivitth ) 1 d y temperature coefficien e reactivitth d an t y n feei dE Compositio R strea d an s comin i mA M f gno from high level wastf o e feedback behavior unde anticipatee rth d transient without scram (ATWS) with LWR spent fuel We considered actmides as U-234 to Cm-246 and RE as Pr- simple asymptotic method measuree « propertiee d th ,an ) A containe2 dM f o s d 141 to Eu-155 (16 nuclides) which affect on reactivity of the core fuel suc s phasa h e stability unde e cooperatioth r f CRIEPo n d JRCan I - Karlsruhe 2 3 CALCULATION METHOD Nuclear library is ENDF/B-V for MA nuclear data and JENDL-2 for RE nuclear data Delayed neutron data and group parameters for fuel are given by referenc4 e 1 INTRODUCTION

e instituteIth n e studyinar y e dr w , a partitionina g y b A M f o g

processing method from high level waste (HLW) and a transmutation of MA by a metallic fuel FBR A metalli'1 s somcha e fue R advantageFB l ) 1 , as s Neutron energy spectrum is harder than that in a MOX fuel FBR Therefore, transmutation rates of MA are higher than that in a MOX fuel FBR 2) Studie t Argonna s e National Laborator y processindr e d otherth an y n go s method (1 e pyrometallurgical partitioning) of alloy fuel suggest the simplicit y processine compactnesdr th e d e methoth an y th s d i gf an do s suitable for a metallic fuel Metallic fues severaha l l advantage previousls a s y stated t therbu , e is a disadvantage that the dry processing has low separation efficiency between RARE-EARTH A M (RE Thusd an ), som e loadeear amounE R d f wito t A M h in a core simultaneously When 90% of RE are removed in the partitioning process s predictei t i , d thae sam A wil th tM thas e e a f b l amouno t E R f o t partitioned from HLW From the previous study for MA recycling in the metallic fuel FBR, MA transmutation rate was 14-15%/year^' even if RE are containe e metallith n i d c fuel For the fuel contained MA with RE, it is has to be analized the core safety parameters (reactivity temperature coefficient) and to be examined fuel characteristics

In the present study, we analyzed Dthe reactivity temperature coefficien e reactivitth d an t y feedback behavior unde e anticipateth r d CD Inner Core transient without scram (ATWS) events with smple asymptotic methode w d an , ) OuteC2 r Core Unit(cm) measured 2) the properties of fuel containing MA and RE such as phase RB) Radial blanket PL) Gas plenum stability unde e cooperatioth r f CRIEPno JRC-Karlsruhd an I e SH) SUS316 shield BH) B4C shield

Fig 2 1 lOOOMWe Metallic Fuel FBR Design W We calculate e reactivitth d y temperature coefficient taking account 03 e coroth fe structure material thermal expansio d nuclidean n s density change base perturbation do n calculation

2 4 ANALYSIS CASE In this study e analyzew , e followinth d g cases, 1} Reference case Loading fuel in metallic fuel FBR core is mainly composed of U and Pu However self generated MA and RE are also loaded In this case, loading ratios of MA and RE are 0 9% and I 35% respectively and RE separation efficiency in discharged fuel is Calculatio equilibro nt i assumed to be 67% recycle A wit M E E loadincasR R % h22 }d e an f heavo g A % M Eacrati2 y f s o hi o metal respectively u LoadinP d an g E fue R s compose i , l MA f o d partitione d froan mR dischargeLW df o froW HL m d fue f metallio l c fuel FBRdepleted an , U dfroenrichmene th m t plant

3-) 5% MA with RE case Similarly in above caser each of MA and RE Temperature loading ratio is 5% of heavy metal respectively (doppler T+500)

2 5 REACTIVITY TEMPERTURE COEFFICIENT

Numbe| r density inj if fus ion calculât on I equilibrium recycle J 1 it thermap ex l enhancing neutron leakage The absolut ee dopplevaluth f o er constane delayeth d an dt neutron fraction decrease as an increase of MA and RE loading ratio Decrease of i the doppler constante decreasth o e mainlt f U-23 ar o see du y8 loading

Dopple• r constant •Effective delayed neutron fraction TABLE RESULT TEMPERATURF SO E COEFFICIENT CALCULATION

Température coefficient

1 Temperature coefficient Fuel* -0 09(4 71) 1 ATMS event calculation -0 086 0 096(1- ) 6% 1 Sodium bond 0 030 0 032() 67% 7%6 (2 ) 8 03 0

2 Flo 2 w g DiagraFi f mo Calculatio n Cladding 0 044 0 047() 8% 6 0 055(25 0%)

Duct 0 016 0 016(0%) 0 019(18 8%)

Coolant 0 264 0 282(6 8%) 0 339(28 4%) The perturbation calculatio s executewa n d with CITBURN n I code CITBURN code base n CITATIOo d N code e burnupth , , inpu d outpuan t t module Radial expansion -0 5 0 - 52() 40% 0 565(13%- ) are developed in CRIEPI Fig 2 2 shows calculation flow Input data are Doppler constant -4 94E-3 -4 48E-3I-9 3%) -3 23E-3I-34 6%) e nuclidth e compositio t a equilibriun m recycl d averagan e e effective (TdK/dT) microscopic cross sections of each core region The average effective microscopic cross sections have 18 energy groups structure and are values Delayed neutron 3 35E-3 3 29E-3(-) 8% l 3 10E-3I-) 5% 7 o differentw n i t temperatur t ea norma conditions i e l On operatins g temperature condition and the other 13 at 500 degree higher than the normal operating temperatur r perturbatiofo e n calculation * This value does not contain doppler effect amount, because U-23 s largha 8 e resonance capture cross sections between lOeV and 3keV MA nuclides have smaller delayed neutron fraction values y. looo than tha U-23f o t 8 Therefore E loadinR ,d increasan gA M rati f o e o caused Sodium boiling temperatur« decreas f coreo e effective delayed neutron fraction 900

6 REACTIVIT 2 Y FEEDBACK UNDER ATWS We analyzed anticipated transient without scram (ATWS) events such as

unprotected los f floo s w (ULOF unprotected an ) d los f heao s t sink (ULOHS) 700 and unprotected transient over power (UTOP) by simple asymptotic method*5 600 Fig 2 3-2 5 shows the sumple estimated outlet temperature under ATWS event e figurTh s e shows e outle thae trenth th tf to d temperature rises a s the increase of loading amount of MA with RE 500 ReferenceMA2% MA51 RE5%

3 PROPERTIE R U-PU-ZFO S R ALLOY FUEL CONTAININE R D AN A M G 4 FiAsymptoti 2 g c Outlet Temperatury b e Quasi-Static Method under ULOHS event 1 ITEM 3 MEASUREMENF O S T

U-Pu-Z candidate r th base f o de eallo on fuel s r transmutationi y fo s , not only because the high efficient transmutation is attained in the P metallic fuel FBR, but also because the chemical and physical properties of U-Pu-Zr-MA-R differeno s Ee b coul t tdno from thos U-Pu-Zf eo r ternary 1000 alloy 900 The properties of U-Pu-Zr-MA-RE, such as phase stability, thermal conductivity, compatibility with cladding and so on, are measuring in the cooperation stud CRIEPf o y JRC-Karlsruhd an I thosd An f eU-Pu-Zo e e rar 800 also measured as the reference 700 Concernin stude phasr gth fo y e stability, specimen U-Pu-Zrf so , containing 2% of MA and 2% of RE and those containing 5% of MA and 5% ot , respectivelyRE , were anneale variout da s temperature r aboufo shour0 5 t s 600 quenched an rooo dt m temperature These samples were analyze y alphadb - autoradiography, metallography and electromicroprobe 500 Referencr MA2* MA5% RE2% RE5%

5 FiAsymptoti 2 g c Outlet Temperatur Quasi-Statiy eb c 1UUU Method under UTOP event Sodium boiling temperature 900

2 3ALPHA-AUTORADIOGRAP H 800 • Fig 3 1 and Fig 3 2 show the alpha-autoradiographs of 2%MA-2%RE alloy 700 - and 5%MA-5%RE alloy after the heat treatment In these samples, major alpha activity emitter is amencium, therefore, small white spots are correspond 600 to the positions of americium Americium is homogeneously dispersed in 2%MA-2%RE alloy In 5%MA-5%RE alloy, however, larger spots are observed. According to EPMA, these spots consist of 30%Am-10%Pu-60%RE (wt%). o 500 l 1 1 1 1 oO R<ïferenceMA2% MA5% 3 3 METALLOGRAPHS RE2* RE5% Fig 3 3 shows the microstructure of 2%MA-2%RE alloy annealed at about Fig 2 3 Asymptotic Outlet Temperature by Quasi-Static 500, 600, 700 and 800 C The specimens annealed at temperature about 500 CD Method under ULOF event C sho 0 anw 60 dthre e phases Accordin FPMAo t g , U-rica thes e ar eh phasea , 500 - 950-C 500 'G 950-C

600-C 600t slowly cooled from 700°C

slowly cooled from 700°C

700 - 700'G

850'C 850'C Fig.3. l ajpha - autoradiograph of U-Pu-Zr based specimens, containing 2% of MA 2 Figalph. 3 . autoradiograpa- f U-Pu-Zho r based specimens, A containinM f o % g5 and 296 of RE. which were quenched from various temperature (X20) . whicRE hf o wer % e5 quenche d an d from various temperature (X'20) Zr-rich phase and an amencium-rare earth phase. The U-rich and the Zr-rich phase correspond to zeta and delta phase in the U-Pu-Zr ternary system. Abov ee e Zr-ricU-ricTOth th d , OC an hh phas e unifiee phasar ed on an eo t d t i correspond o gammt s a e ternarphasth n i ey system.

3.4. OTHER PROPERTIES

Thermal conductivity, density, mechanical properties, solidus- liquidus temperature and another properties of U-Pu-Zr and U-Pu-Zr-MA-RE are measuring. However, no significant differences are observed, because e consistear e E inclusionth R s d a d an d thesA an sM e f o almos l al t inclusions do not seriously affect the various properties of the alloy. Therefore, U-Pu-Zr-MA-RE type alloy can be selected as the fuel for transmutation.

. A SUMMAR CONCLUSIOD AM Y N

We analyze e reactivitth ) a d y temperature coefficient with CITBORN e reactivitth codd an e y feedback behavior unde e anticipateth r d transient without scram (ATMS) events with simple asymptotic method d measurean , ) b d the properties of fuel containing MA and RE such as phase stability under the cooperation of CRIEPI and JRC-Karlsruhe. In the present study, we obtained following results, (1) The absolute value of fuel (without doppler constant), sodium, cladding, duct and coolant temperatu-re coefficient increase as an increase of MA and RE loadin g e neutroe effecth ratith f y o ob tn energy spectrum hardening. The absolute e temperaturvaluth f o e e coefficient r geometrifo s c thermal 500'G expansion increases by hardening of neutron energy spectrum enhancing neutron leakage. loadin % e 5 casth f o ge n (2, I simpl)rati RE f d Mo e Aan estimate d coolant outlet temperature increases near to the coolant boiling temperature during ATWS events. It should be noticed that present ATWS event analyses are in very simple asymptotic method and a metallic fuel properties such as fuel thermal conductivity are not clear now. Therefore, these analyses show only trend of change caused by MA and RE loading. However, these ATWS events analyses indicat e necessitth e f coro y e design considering ATWS detailed analysis. Fro l resultsmal e resulte simplth th , f o se asymptotic analysi n i ATWs S events show thae metallith t s safetha c R fueyFB l s e i loadedimportanar t E potentiaI R . d tan A thale M th evet f i n '^i-:;v-^w- ' r f- > reactivity feedbac f thermao k l expansion increases slightly witE hR Mfd tan loading t dopplebu , r constan d othean t r parameters decease witE hR Md Aan ^:-j|l|lsàp* loading. e characterizatioth y (3B ) ne fueth stud , f l o ycontaininRE d an A M g ^•O'-'^-^é&^r-K americium and RE inclusions were detected in the grainboundaries. However, . •"-^>^?3S, there are no significant difference between U-Pu-Zr-MA-RE and U-Pu-Zr alloy * * /T^,;LiiU*V*^ 1^> on thermal conductivity, density, mechanical properties, solidus-liquidus cw ' r œ *©^ temperatur othed an e r properties > /. *n'j li'.-jd '"'•^•-. '

REFERENCES

, . TINOUE . l a . t e , Developmena P f o t . 7,t. ^oning anri Transmutation 700 'G 850'C Te^hpology for T.ong-Iiypfi Nuclear Technology , Vol . 93 , 201(1991) Fig 3 3 micrograph of U-Pu-Zr based specimens containing 2% of MA 2. A.SASAHARA, T. MATSUMURA, An Assessment of TRU Recycling i'n Metal and 296 of RE which were quenched from various tcmperatuic EBE, International Conference on the Physics of Reactors, Marseille, (after e'.huig XlOOO) France, April 23-26, 1990 3 T YOKOO,et al , A Design Study on the Metallic Fueled FBR Cores- STATU WORF SO TRANSMUTATION KO N ro Evaluation of the Pen ornyince of Metallic: Fuel and Core. CRI EPI-Report T88043,198 Japanesen (i 9 ) IN SWITZERLAND (Summary)

Studies on the transmutation of long lived radionuclides are concentrated at the at the Paul Scherrer Institute (PSI) m the department "Nuclear Energy" Merging out of existing activities on Nuclear FueReactod an l r Physic generao tw s l subjects have been identifie r expenmentadfo l work . PREPARATION of TARGET MATERIALS for fast reactor application utilizing wet chem istry processes for the direct conversion of nitrate solutions into a fuel form (denses spheres) or in a intermediate product to be pelletized (porous spheres) preparo t s i m e ai uramum-plutomum-neptuniu e Th m nitrid uranium-plutoniud ean mm tnde microsphere comparo t d an s e spher hybrid c fuean epa l d sphere-pelley t dr wite th h route pellets in a fast flux reactor This cooperative work is being shared with the Départe ment d' Etudes des Combustibles at Cadarache of the French Commissariat a 1'energie atomique (CEA) and PS I In parallel evaluation a , f possiblno e mamx material fore th f oxidem o n i s r nitrideso s preparee t ot b chemistr we y db y processin startes goxide ha Th df magnesiu eo e th d man nitride of zirconium are in the center of interest Preliminary tests in the zirconium-cerium oxide and the zirconium uranium nitride systems have been initiated in collaboration with Japan Atomic Energy Research Institute (JAERI) . NUCLEAR DATA AND CALCULATIONAL METHODS PROBLEMS e relateth o t d accelerator based transmutatio f actimdeo n s High energy nucléon meson transport computer codes used in the design of accelerator based transmutation systems capable havb o et f correctleo y predicting, among other things, the yield and the mass distribution of spallation and fission products generated in the tar get Simple code comparisons for the irradiation of thin samples of actimdes have revealed considerable difference predictioe th n i s f thesno e quantitie resolvo T s discrepanciee eth s an confiro dt hige mth h potentia high-energe th f o l y fission reaction transmutatioe th r fo s n of actimdes expenmentan a , l programme, ATHENA (Actmide Transmutation using High ENergy Accelerators), has been initiated In a first phase of the programme, thin samples of uranium, neptuniu d amenciuman m encapsulate aluminiun di m wil e irradiateb l d with 590 MeV protons from the PSI nng accelerator using the PIREX irradiation facility The distribution of the reaction products will be measured using different methods such as the e totath ld reflectioan ICy S fluorescencPra M X n e techniqu compared an e d with results of model calculations I-or accelerator based system whicn i s h proton usee sar d directl transmuto t y e actimdes, integral Information on the adequacy of neutromc design methods is scarce In a second ATHFNe phasth f eo A programme therefors i t i , e propose studo dt y neutromc behaviouf o r such systems with the help of zero power experiments at a separate beam of the accelerator e physicTh s related studie framewore th carriee n sar i t collaboratioa f dou o k n wite th h SCIENTIFIC RESEARCH PROGRAM ON ACTINIDE Département d'Etudes des Réacteurs of the CEA at Cadarache TRANSMUTATION BY USE OF FAST REACTORS The activities make use of existing knowledge and capabilities and utilize an almost unique combinatio f facilitieo n s topicso availabltw e , t PSIacceleratoTh a e . r transmutatio d targean n t L.A. KOCHETKOV, A.G. ZCYKUNOV fabrication, are regarded by the PSI as long-term basic research and help to preserve and further Institut f Physiceo Powed san r Engineering, develop competenc aree f th advanceao n ei d fuelreactod an s r physics Obninsk, Russian Federation

Abstract

e recenFoth r te Ministr yearth n i f sAtomio y c Energ f Russiao y n Federatio e nationath n l W managemenprograRA n o s mbeeha t n under development. The program of actinides transmutation (as the most hazardous pars s treatef RAWcomponentwa o it t) s a d . Accordin o thit gs program various investigations within this field have been supported for the o lasyears tw e t prograTh . m includes five sections dealing with actinide transmutation with the use of fast reactors.

e probleTh . f 1 o mradioactiv e waste (RAW) managemens ha t recently becom f immenco e e importanc n nucleai e e th r l poweal f o r countries involve s it solution n i d . Evidentl e nearesth n i yt future it will be an issue of the main priority in the IAEA activity along wit a safequardth h NPP'd an a s safety e nucleaTh . r power progress could hardly be anticipated without its adequate solution.lt doesn't mean tha e authoritied expertth t an s s absolutely neglect the problem. The time has just come -when the attentio o thint s proble s i mconcidere d insufficient .whee th n problems of finding the sound solution of RAW management are more obvious. Nowadays we are quite aware of the international aspect of the problem and the scale of the financial support necessary. The nuclear power development as it is makes it necessary to work out: - cirtain concepts of RAW management ; -nationa internationad an l l standards, regulation rulesd an s : -research and development programs; -practical recommendations based on aciantific and research works e assimilateth n o datd aan d expertize. When working out the concept of RAW management we should proceed from the assumption of at least long-tarm preservation of the U Earth natural radiotoxicity level. If wa leave alone low-level d middle-levean l RAW d considean , r only high level wastes - conventional fast reactor core design meant for actinides produced mainly (98 nuclean i £) r reactor core e probleth s f o m transmutation ( Pu,Am,Np,Cm) ; their management could be reduced to at least two independent - desig specializea f no d fast (fast-thermal) reactor, meanr fo t aspect solutee b o st d quite differently: transmutatio actinidef no d probablsan mose yth t hazardous - iaolatio e maith nf no par f o fissiot n products froe th m fission fragments; environmen e perioth r f 300-50o dfo t 0 y-3ars; - associated problems, such as - elimination (conversion, spatial ejection a numbe f f o )o r f radio-chemical transuranians.actinidea, in particular the Pu-isotopes, and the - fuel fabrication technology so-called minor actinides-Np,Am,Cm whose radiological hazard is » transport conserve r milliofo d n years.The majorit f o investigatory s - biophysical conside compaca r t multibarrier isolatio f fissiono n fragments « feasibility storage under surveillance to be the most natural and In additio e e logisenvesagenth ar s tle r fo d efficien solvinf o y e firswa tg th t problem. - physical studies; e mosTh t reasonabl f o solvin y e seconwa th eg d probles i m - technological investigations; evidently the fission fragments utilization by means of fast - development of core designs. reactors. The advantage is testified by the following; Sectio e (190-1994nOn ) stipulate definitioe sth f requirementsno , - the accumulated positive experience of the design and operation elucidatio f restrictionsno , conceivin e principlesth g , concept of fast reactors (fast reactors technolog presena s i yy tda and engineering proposals. The determination of requirements to reality) ; the dept f mosho t hazardous nuclides extracts from spent fuel speciphic physical feature f fasso t reactors whicn reallca h y deems essential. utilize (burn or transmute) not only Pu-isotopea but all the Section Two (1990-1994) nuclides calculational and experimental minor actinides withou y an additionat l r coso energt y investigations aimed for the provision of adequate accuracy of consumption for their transmutation ; physical characteristics responsibl safetr efo d economy yan e Th . - their transmutatio fasn i n t reactor show bese sth t ecological fragments yield, delayed neutrons characteristics, nuclear effect. reaction constants at a microlevel, group constants are to be refined. Concurrently calculational codes are to be up-dated. For the recent years in the Ministry of Atomic Energy of Russion Section Three (1992-2000) stipulate e elaboratioth s f o n Federatio nationae nth l programanagemenW RA n mo bees ha tn under engineerin e gBN-80 th desig r 0 fo n (BN-600) core. with minor development e progra.Th actinidef mo s transmutatio e mosth ts (a n actinides dope conventionas dit inte th o e lamoun th fue n i lt that hazardous part of RAW) is treated as its component.According to doe noticeablt sno y alte e physicarth l characteristice th f o s this program varions investigations within this field have been BN-800 core developed nowadays. Traditionall n thii y s e casth e financially supported for thr last two years. The program includes the following five sections dealing with actinide detailed parametric investigation f o physicas l anot thermophysical characteristics should b.s performed,the optimum transmutatio f faso te nus reactor wite th h s solutions on the core composition arrangement and handling of - physical and technical grounds, ecological and safety fue d absorbinlan g material shoult i i e b foundir sd , that will problems; provide for the required burn-up, stability of power density - constants, calculational and theoretical justification; se If-régulât ion. Section Four (1992-2005) stipulate e developmenth sn a f o t - design of convencional fast reactor cores, meant for actinides absolutely new core with an increased content of actinides and burning (transmutation): free from U-238. - desig a specia f no l fast reactor wit a high h amounf o t These developments should be baaed on technological achievements actinides ; that would form the ground for the next fifth section. The need - associated problems (radio-chemical, fuel production, for Section Four may emerge in the case of more intensive transport). actinide burning occurrence. currene Th . 4 te prograth statu f o sm feature e followinth s g studies: e fiel th f constantso dn i - ; - conventional BN-800-type cores witn additioa h f o minon r actinides; - on irradiation of minor actinides oxiie samples in the BN-350 reactor. Preparation is being conducted for experimental fuel elements exposur e BOR-6th n i e0 reactor. Investigation e beinar s g helo t d support the concept of management of long-lived (most hazardous) components of nuclear wastes. Fast neutron reactor e considereb y able ma o sb t attai eo t d a n supplementary vital incentive for confirmation and development, for they wil e ablb lo solvt e e probleth e f managemenmo e th f o t most long-live d hazardouan d s radwaate f nucleao s e rth powe n i r most rational way. In this context the joint efforts of the IAEA member countries and primarily of those committed to fast breeder reactor developmen e solutioe th t actinideth aime r f fo o nd s transmutation problem appears desirable and beneficial.

Summary

The program on actinides transmutation with the use of fast reactors has been elaborated.lt covers the period of 1990-2005 and is financed e Ministrbth y f Atomio y c Energ f o yRuasio n Fédérâtion.The program comprise e followinth s g five R&D aspects: - physica engineerind an l g grounds, ecologica safetd an l y problems; Ul - constants, calculâtional codes; PHYSICS ASPECTS OF TRANSMUTATION OF MINOR ACTINIDES AND LONG LIVED FISSION PRODUCT FBRN SI s (Sessio) n2

Chairmen

L.A. KOCHETKOV Russian Federation L. KOCH Commission of the European Communities REACTOR ASPECT ELECTRONUCLEAF SO R lines - different schemes are to be compared between themselves TRANSMUTATION OF ACTINIDES IN THE and to alternative reactor schemes. But it's too wide a task and we shal e limil th d considenarro o an tt t wi r onle yb whay ma t HEAVY WATER HIGH FLUX BLANKETS. called polar opposit o t efas t breeder transmutatio - heavn y ANALYSIS OF PROBLEMS water super high flux electronuclear blanket burners. Other reports to present meeting will, no doubt, provide ample opportunity for comparison. E.V. GAI, A.V. IGNATYUK, N.S. RABOTNOV, Although accelerator driven neutron multiplying transmutation A.A. SEREGIN, Yu.N. SHUBIN devices are formally called targets or blankets due to their subcriticality, in the case of large power producing industrial Institute of Physics and Power Engineering, systems they woul e b dreactor l righal s t wite hth somf o e Obninsk, Russian Federation problems characteristic for standard reactors and some new ones not less d reaseriousan l . First proposal n o neutros n Abstract transmutation were base n reactoo d r options (see [1]). Minor actinides (MA) are in fact threshold fission elements. At neutron energies capturV abovMe 1 e e cross section e onlsar a y e followinTh g problem reacton i s d neutroran n physic f higso h few per cent of fission cross sections, so fast breeder reactors flux heavy water accelerator driven blankets are considered and (FBR) were the first to be considered. But it was immediately showvere b yo n t e seriou 0 1 fluxee th n/c th f t o sa m: s understood that averaging over real fast reactor neutron spectrum results in fission to capture ratios 10-20 times larger difficulty of attaining needed neutron flux; high radioactivity spectruR PW thar t stilnfo m bu l considerably less than unity, d powean r releas primare th n i ey circuit; tritium accumulation; e relativelar m closeA d 0.2o yrt an p .N poo r r fuefasfo lt breeder y considerablan d san e hardenin e spectruth f o s gi m heavy water radiolysis and hidrogen release; positive welcome specializeo .S d fast breeders-actinide burners (FBAB) flux-reactivity feed-back; poisoning. At more moderate fluxes were natural option to propose but resulting improvement was not qualitative, about threefold. e Hence th interesth e o t t fission rates of the minor actinides become unsatisfactory which transmutation possibilities of electronuclear facilities. e accumulatioth lead o t s f o heavien r hazardous long-lived Four main types of actinide waste may be singled out and ordered by the complexity of the isotope composition: isotopes e desirabilit.Th f o yinvestigatio f statio n c liquid a. Military waste left after effective plutonium and uranium blankets which may solve some of those problems is pointed out. extraction froburw lo mn fuels with Np-23 a mai s 7a n component by a large margin. b. "Minor actinidea" (MA) waste left after radlochemical reprocessin f o commerciag P spentNP l , well burned fuel; Np-237 comprises abou halta f thio f s waste e resth ,t being 1.INTRODUCTION mainly Am and Cm isotopes. c. Commercial spent fuel with only fission products and most of Existing transmutation project splie b y t sma into following uranium remove y reprocessingb d plutoniums i t i ;f o A som ,M % e90 main categories: the rest, high flux reactors with therma r slightlo l y more hard d. Spent fuel proper as a product of open nuclear fuel cycle. (epithermal, resonance) spectra; Each of the groups a-c is a potential raw material for fast reactors (power plants or specialized actinide actinide transmutation and needs, in principal, individual burners); approach y isotop.Ke e Np-23 moss i 7 t "stubborne th allf "o d an , - subcritical assemblies and blankets both thermal and fast overall effectiveness of actinide transmutation schemes must be drive higy nb h current accelerator neutron sources. measure success it y burninn b di s g neptunium. e InteresTh transmutatioo t n potentia f o electronucleal r plant s growinswa e glasyearsw th fas fe tn i t. Durin e lasgth t fifteen months at least three international scientific meetings 2. ATW PROJECTS on transmutation were held: in Obninsk (Russia), July 1991; in Vienna (IAEA,Austria), October, 1991 ; in Villigen,Switzerland, LANL scientists have proposed what is probably the most March, 1992. The last meeting was devoted completely to radical concept of accelerator driven transmutation [21. Its electronuclear transmutation. The papers presented to the basic ides veri a y appealing. Minor actinides nuclee ar i meetings are the main part of the information base for present effective absorbers of thermal neutrons with low fission report e completTh . e comparative analyse f o reactos d an r probability vern i yt higBu . h fluxe f thermao s l neutrons planned CO electronuclear options should principlen i , done b , e alono tw g in LANL scheme the fission of the isotopes formed after the cn capture of neutrons by long—lived MA radionuclides playe th s Considerably less attention was paid so far to very serious O main role (see Tabl. TheiI) e r significant equilibrium problem e analogth s f whicso h plagued exampler fo , e molteth , n salt reactors and prevented their practical use in spite of concentrations may be reached only if the flux is close to 10 considerable long time efforts of ORNL where two experimental neutrons/cm s (relatively long-lived Am-242m is an exception). reactors wer w eyear fe buil d operatesa an t eachr e fo Th d. Thus the absorbers turn into superfuel. problems are: extremely high radioactivit decad an y ye th hea n i t primary circuit; complex, ever varying composition of the circulating liquid; possible sedimentatio f o radioactivn d an e TABL COMPOSITIO. I E D MAJOAN N R NUCLEAR PARAMETER F PVRSO s fissile materials in various parts of the circuit; coolant's MA DISCHARGE (THE CROSS SECTION VALUES TAKEN FROM [3]) radiolysis, especially dangerou case profusf th eo n i s e hydrogen production. These difficulties may be qualitatively aggravated in supei—high flux transmutation plants. Some new ones arise Minor actinides Capture products too. We shall consider them one by one with the emphasis on b , , o-o- , b , Nuclid- o e Nuclide P.,/P P 0^.P/ Np ° f Tl/2 superflux case needed to burn Np-237. 237 Np 1 . 0.573 176 0.02 ^2NP 203 2027 2,12 d DIFFICULTE 3TH . F GETTINO Y G HIGH EFFECTIVE FLUX 0.46 0.264 587 3 .2 „f,t Am( 533b ) ? 2100 16,1 h „..Am(54bl 13446950 152 a l verFirsal yf o thig h e schemflu th vitas i xr efo l whics i h 243Am 0.224 0.128 75,1 0.2 24? Am(3,8b) ? 2300 10,1 h illustrate Fig.. 2 Fig.y b dd l an ltake n fro] show[6 m s crucial „ , (-Ami 71 , 3b ) ? 26 min neutronics paramete e numbeth f neutrono - rr s requirer pe d 244 1600 Cm 0.049 0.028 15,2 19 ,0436 ?46Cm 2145 8500 a fission as a function of thermal neutron flux in Np-237 Cm 9.4-3 5,4-3 369 2145 ;™Cm 1,20,124 4700 a transmuting system. When it crosses a horizontal line indicating 24,3 Cm 8.1-4 4.6-4 130 617 2?3Cm 15.1 .024 18.1 a the numbe f neutronro s release fission i d n event absorber turns 3.1-4 1 .8-4 16 <5 ;"cm 130 617 28.5 a into fuel. Fig.2 demonstrates thae ratf th to fissio e n drops 246Cm 2.4-4 1 .4-4 1 .22 0.14 Cm 57 81 1 .6+7a

4.0 -, The scheme is oriented on the use of liquid fuel-coolant circulating between blanket and heat exchanger, part of it going tradiochemicaa o l bypas r continuoufo s s reprocessinge Th . scheme allow o avoi st o majo tw d r difficulties: remote controlled OT productio f fueno l pin f o highls y radioactive materiald an s £1 additional high level waste from construction materials of the fuel assemblies. Prospects of the method are estimated optimistically and the research activit considerables i y n lateI . r versions [4-7e ]th evolving concepfacn i ts spliwa t t roughly accordin selecteo gt d 3.0 - ABSORBER target waste group. The common feature of all schemes besides a 1-1.6 a linaGev,<^heava m s 0 i cy10 > water blanket containing _0o) solutions or slurries of actinides and/or technecium compounds. E High fluxe crosd san s sections allo principln e i wver us yo t e U low concentration e nuclldeth e b transmute f o o st s d still getting good burning rates. Heavy water fuel-moderatoi—coolant ( FMC ) circulates in the primary circuit of most probably FUEL three-circuit unit. This general concept was accepted as most promising one by the majority of electronuclear experts in our 2.0 ——1— 111)——I—I I I I | Mill1 ] —i i 11 iiii|—i i i mii|—i country too, the preferred versions and estimates being close to American ones e author[8-101th l Al s. poin t ou generat l 10 ,"3 1Q10M " 10 '5 10 " 10 " advantages of liquid fuel schemes: the absence of nonuniform Neutron flux, n/cm2 s power density and of radiation endurance problems characteristic for fuel pin systems and the possibility to adjust the Numbe. 1 g f neutrono rFi s require o fissiot d n Np-23 a therma n i 7 l compositio core th ef no continuousl y throug ha chemical bypass spectrum as a function of flux level [7]. The horizontal line without frequent reloadin f fasgo t burning solid fuel. indicate e neutronth s s release n fissioni d . only by refusing to produce electricity and using much colder water in external circuits. But even in this case r values much higher than 0.3 are hardly realistic. Impossibility to produce V typpoweAT en i r scheme woult onlno dy complicatd an - e considerably - the economic side of the problem but also undermin principae th e l advantag f subcriticaeo l a safet f i y standard critical NPP supplying the llnac with electricity become n indispensablsa e schemee W projecparth AT f to o .s S i t rightly planned as power self-sufficient and even electricity producer. e spatiaTh l nonuniformit e flu equivalens th i xf yo o t timt e nonuniformity. Wit centraa h l axial neutron sourc e volumth e e with maximum flus i limitex d spacan d e averaging reduces resulting effective value. The burning half-live e cruciaar s l parametery an f o s transmutation plant. They e depenloath d n o de planfacto th f tro which enters denominators. In electronuclear transmutation the Neutron flux, n/cm2s total load factor is a product of partial ones - for the llnac, target complex, blanket pland chemicaan t l by-passy ma t I . share Th Np-23f Fie. o g2 7 turned into Pu-238 Instea fissiof o d n reduc valus It e e considerabl prolond yan g transmutation cycles as a function of neutron flux. acting just like decreasing of the effective flux.

E DIFFICULTIE4TH . F HANDLINSO G HIGH EFFECTIVE FLUX fast with decreasing flux and its place is taken by production of Pu-238 which is hardly less of a radiational hazard than 4.1. Tritium production initial Np-237. Accumulatio tritiuf no m woul vere db y serious proble hign i m h One serious difficulty in getting high enough flux is this: flux heavy water blankets. It accumulates due to neutron FMC spends e onlfulth ylf o parcirculatio , tT ne th perio n i d absorptio deuteriumn i n . Thermal cross sectioprocese th f s no si pumpes i blanket t i e timd th e res? throug eT .Th f o t e pipeshth , Just 0.53 mb but in the flux of 10 neutrons/cm s it results in heat exchanger d radisan o chemicae ar l2 T units d an botf I .t hT tritium production rate about 1000 Ci per ton of heavy water a day and equilibrium concentration to be reached in some twenty much less than relevant decay half-time then effective neutron years is about 10 million Ci/t. Hundreds and even dozens of flux $eft enters a simple basic equation determining the Curies of chemically unseparable tritium in cooling water are equilibrium concentrationy ke e th f o s short lived well considered as significant problem in PWRs, so it's clear that fissioning nuclides tritium quantities 5-6 orders higher would create grave difficulties indeed if the fluxes like 10 cm s in heavy water ( 1 ) become realit. y whers nuclide'i eN s neutron number - density ,p X , decay 4.2. Heavy water radiolysis and hydrogen release constant differ„ . .$ s from physical neutron flux$ In the solid fuel reactors first protective barriers -fuel slug material and fuel pin shell - prevent not only accumulation V' (2 ) of radioactive fission fragments in the coolant but radiolitic destructio molecules it f no fissioy sb n fragment s sa well e Th . Operating NPPs have r values varying from approximately 1/20 radiolysi f wateso r lead productioo t s hydrogef no d oxygennan . (PWR) to 1/300 (BN-800 type FBR). For specially designed The latter reacts with the water forming peroxide. Hydrogen is blankets [7] T ~ 0.2 and the effective flux is reduced to almost mainly released because recombination processe e relativelar s y 2x10 instea e necessitf physicao dTh . 0 o kee1 yt l r pvalu e weak in pure water. Approximately one million of H2 molecules is as close to unity as possible conflicts seriously with t powecreateGW t A r f fissio y o . evere wateb ] d t th 1 n ac y[1 ri n thermophysical requirements C e heatinliqui.FM th n i dg fuel level it means the release of a few cubic meters of hydrogen blanket vers i s y effective volume process without separating (npt r seconpe ) heavn i d y water blankets 1105. This hydroges i n Oi walls. Coolin s muci g h less effective, slower, especialln i y very radioactiv o tritiut e du me componen d meanan t s high multi-circuit systems e b improve . T valuy ma ed substantially explosion danger. The radiolysis may be inhibited in normal en operating condition y recombinatiob s n catalyzers like coppet bu r TABL AVERAG. J--ENERGIED EII AN E MINO- Eß TH F RO S ACTINIDES to in case of an accident leading to intense boiling recombination F DECAO T PEYAC R goes down while radiolysis intensit n soms timei yte e s higher for the vapor than for the water [11 ]. Addition of the copper Nuclide keV keV inhibitors (the quantitie o t mol/1s p u neede) e absorar d b neutrons damaging neutron balance already vulnerable at high 238 190 647 fluxes. Peroxid strons i e g oxidizin corrosiod gan n agent. These problems speak for themselves. They are not directly connected 159 18 with the flux itself and are determined just by total power of Am 109 806 fragments released inte waterth o . The precursors of the delayed neutrons also circulate in the primary circuit (see section 4.4.3 below) so there will be not 4.4. Complications in blanket kinetics negligible releas f o neutrone s producing l fissional n i s Kinetic problem y see sma f mino mo r importanc r subcriticafo e l sections of the circuit with resulting production of hydrogen. facilities but it's only true for the dangers associated with It would be by a few orders of magnitude slower than in the the instabilitie steadf so y critical condition which cause most blanke ty becom itselma t seriouea bu f s complicating facton i r concern in traditional NPPs. Attaining and sustaining stable heat exchangers and pumps calling for special measures to deal neutron multiplication and flux on planned level would be with and seriously affecting safety. difficult task with heavy water e blanketexampleth s a ss considered below Illustrate. 4.3. Radioactivity and heat release In the primary circuit 4.4.1. Strong positive flux-reactivity feed-ôack If a radioactive nuclide is produced during the irradiation of a liquid circulating fuel in the core the share of its nuclei Equilibrium concentrations of the well fissioning nuclides stationara e cor ith t n ea y irradiation e regimexpresseb y ema d grow with the flux and so does reactivity. This positive in term decaf so ye timeconstanth sd an spen tX t insidd an e feed-back is a potential source of instabilities calling for outsid core th e] e 2 T.an[1 „ T d investigation. Her e shalw e l give e onlexamplon y e estimating and comparing reactivity time derivatives after a stepwise flux increascoreso tw r :fo euraniu e mth f U-23enriche o n i % 3 5 o t d ( l - e XT1 )( l - e XT2 ) d neptuniuan flu 0 1 x m with .equilibrium concentrationf o s - l = , , /N N (3 ) . e 0 flu1 Np-23th xd r an valuefo 8 0 1 s c tot The single most sensitive factor of the four determining k (which we shall designate simply k) is 77 and for present limited If all the exponential arguments are much less than unity this purpose we can approximately take k * 17. Then ratio turns into Nc/Ntot= t = Tj/ (Tj+ T? ). That's the case for dp 1 ak f dn dr, j ar 1 most of the fission and capture products so only a part of total ) 4 ( radiatio d decanan y heat e equa releases th i core d lt th ean n i d at at T7 an, <ït res othen i t r e partprimarth f so y circuit. Withou n o lint e chemical reprocessing this hearadiatiod tan n release grows with densite th wher s i f fissioninyo en g nuclide. e th e b Len t the time durin campaige gth n reaching typically cenr aboupe t t8 of total power. On line cleaning is rather slow and reduces only density of the absorbing nuclide. Then slightl short-livee th y d radioactivity which make a slion' s release e problesharth th f eo o es m survives. In the heavy water blanket burners there is a considerable (5 ) additio thio nt s vero effect ye higdu t h radioactivite th f o y short-lived well burning actinides Np-238, Am-24 d Am-244an 2 . Their periods were indicated in Table I , average beta- and n will start changing linearly with time after small "flux

gamma-energies are given in Table II. f jump" A$: n,= n.+ßt. ß=H o c f A$ for neptunium and u=-N,o'. ,A$ for The periods of all the nuclides in Table II are more than 10 t 1 It 1 a.j' a hours 1 hour,5 r mossfo t important Np-23 o s they 8affecma y t uranium. Taking that into account we get from (4-5) considerably the cooling after shut down. In general the 2 n extremely high levels of the radioactivity of the primary op "a 1 «Va AA _„ op a 1 "Va circuit liquid fuel-coolant deserve special detailed discussion. and (6 ) Gamma-heating of the uncooled shielding may become a problem. dt 'tf 'ft at for neptuniu d uraniuman m correspondingly. Np-23 d U-238an e 5ar TABL HALF-LIVE. EIV THERMAD SAN L CROSS SECTION U ISOTOPEE F SO S fissioning nuclide , Np-23 sd an U-23e 7ar absorbers 8 . Multiplying and dividing right hand sides of the expressions (6) by $ we can present both derivatives in the form rA$/$ and Nuclide Eu-151 Eu-152 Eu-153 Eu-154 Eu-155 substituting the values of cross sections and densities to calculate crucial coefficien determinintr reactioe o gth tw f no Half-life stable 13 a stable 8.5 a 4.96 a systems on the same relative "jump" of the flux. The results are er , barns 9200 12800 390 1340 3950 -3.8xlO~9s~1,3.5xlO~6s~1and 6.7xlO~ r uraniufo 6 m at$=1014, neptunium at $=10 and $=10 correspondingly. So the neptunium sensa systen i e s i mapproximatel y 1000 times more dynamic than uranium one in the considered conditions. And with uranium the practically coincides with that of Np-238 - about 50 hours - so feed-back is negative (stabilizing) while it's positive the rate of reactivity dropping after a shut down increases and (destabilizing neptuniumr )fo . after a few days a restart may become impossible without complete cleanin samariumf o t gou . u E isotopIf e yields become significan r heaviefo t A M r 4.4.2. Poisoning europium poisoning may be important as Table IV shows. Eu-151 Very high flux complicatespoisoning problems, sometimes poisoning will increase almost five-fold, Eu-152 poisoning - qualitatively. First, some "weak" poisons turn "strong". Second, almost four-fold etc. equilibrium concentrations are reached faster. Third, intense neutron capture transforms some decay chains strengthening minor 4.4.3. Circulating precursor delayee th f so d neutrons branches. Fourth, increased poisoning after shut down grows with The delayed neutrons plao n speciay l roln subcriticai e l the flux. Xenon, main poiso solin i n d fuel, gets eliminated from accelerator driven systems. But there still remains the problem the liquid very fast, in a minute or so [13], but even that is of accidental criticality no matter how small its probability probabl t fasyno t enough because xenon burning half-tim onls i e y case th f suc eo n i h d emergencian s e delayenumbew th lo y f o rd 26 seconds in the flux 10 . Prometium-samarium chain becomes neutrons aggravates the dangers. There is little direct most important and will be considered now. Heavier isotopes of experimental informatio delayee th n no d neutron parameterf o s samarium must be taken into account. The parameters of samarium short-livethy eke d actinides burnin n higi g h fluxes, total poisonin hign i g h fluxe e givesar Tabl n i n e III. evaluatee yieldb y sma d from systematics (14) which gives v,=0.01 . ThosAm e0.00d valuean r p fo 5 N e mucfosar r h 23lowe8 r than 242 TABLE III. SAMARIUM POISONING IN HIGH FLUXES for uranium. Beside a sconsiderabl e portio f o delayen d neutrons (up to about one half) are emitted outside the core *'-2 -1 Tr , hourr. s ^Sm decreasing the effective average further. Eq.(3) must be applied cm s 1/2 in full for o calculatmt e in-coreth e shar f eo delaye d neutron m S m S m S m P 149 Nd 149„ 149 150„ 151„ precursors because most of them are very short-lived. Whether the delayed neutrons induce fissio d multiplan n y significantly outside the blanket depends on the geometry of the circuit but 1.73 51 48 2.2a 128 1.9 casy ian ne neutron background woul vere b d y strong aroune th d d 80 12.8 104. 89 1 8 3 1.73 10 16 1.73 11 0.48 8d 1.28 190 circuit tubing because 10 -10 of all the neutrons are produced there.

e yieldTh specifif so c fission t productknowno r fo e n ar s LOWE. 5 R FLUXE R PLUTONIUSFO M DOMINATED MIXTURES short-lived MA so it's impossible to normalize their poisoning data properly yielde th heaviesf t so , bu t fragments grow with Plutonium isotopes are the main part of the spent fuel the mass number of the fissioning nuclei so the Pm-Sm chain actinide composition, uranium excluded, so the policy toward yielreasonabla s abouf i o d Z 1 t e assumption. plutonium is decisive in designing transmutation schemes. It may The periods in the Table III are to be compared with chemical be considered either as valuable major fuel element or Just as by-pass period, it's planned value being aboue tth montha t A . another waste componen disposo t . Intermediateof e concepte sar possible and the latest ATW version (7) seems to be one of them. flul thre al xm S isotopeeleve 0 1 l s reach equilibrium It uses blanket configuration similar to CANDU core, physical concentration so the poisoning is tripled. "Samarium well" Ul deepens manifol e lasth t s a dcolum f Tablo I nshow II d e an s d wastan flue 0 mixtur1 x r cenpe et whic9 8 plutonius i h r fo m CO unlike iodine well it does not disappear. Pm-149 half-life fuel (see Table 5). This version is already rather close to en TABLE V. INITIAL AND EQUILIBRIUM COMPOSITIONS AND NUCLEAR DATA The first major difficulty eliminate n statii d c blankes i t FUEE TH L F MIXTUR 0 E MODERAT•TH ^ F EO E FLUW VERSIOAT X ] 17 N the distinction between physica d effectivan l e fluxese Th . 2xl0 14 difference may be decisive. The second is not less Important: tf f the primary circuit becomes practicall radioactivt yno e th d ean numbe circuitf ro reducee b eved y an channen a o firsma d t o tw m l Isotope Rel. concentration [7] type single circuits with drum-separators are not unthinkable becaus tubee eth s made wallb ey smucma h thicker than n fuepi l T Initial Equilibrium Vb o-b , { J/2 shells and practically hundred per cent reliable. 237 Small actinide inventory of the blankets with high thermal N 0.0449 0.0317 176 0.02 2.14xl06a neutron flux is considered as one of their main advantages, Np 0. 0.0003 203 2027 2.117d increasing safet d enablinyan g continuous chemical reprocessing. Totap lN 0.0449 0.0321 In earlier versions inventory gain was estimated as "«Pu 0.0140 0.0230 540 17.9 87.71a approximately hundred-fold, in later versions it's almost Pu 0.5148 0.0610 269 740 24 lOOa vanished. Fuel circulation increase e inventorsth r times1/ y , 240 0.2372 0.1003 289 0.056 6569a i.e. manifold. It' e smoron e argumen n favoi tf o statir c 0.0467 blankets. 242Pu 0.0785 358 1011 14.355a _ 243Pu 0.0481 0.2630 18.5 <0.2 3.763x10 a 244Pu 0. 0.0002 87 196 4.956h Pu 0. 0.0026 1 .7 - 8.1x1a 0 CONCLUSION. 7 S Totau lP 0.8925 0.4968 reactoe Th 1. r physic d safetsan y problem f heavo s y wateA M r 241. 0.0513 0.0048 587 3.2 432. 6a 242 0. 0.0001 2100 16. Ih blanket burners wit e neutrohth ns flum looc x k0 clos1 o et 242m™ very serious indeed e mosth ,t Importan e followingar t e th : „43 Am 0.0001 0.0001 1344 6950 152a Am 0.0092 0.0874 75.1 0.2 7730 a difficulty of providing needed effective flux; tritium Total Am 0.0607 0.0924 accumulatio e flu th reacheds i xf i n ; high radioactivite th f o y 242- primary circuit liquid; water radiolys is by fission fragments . Lm 0. 0.0213 16 <5 162. 8d o243- 0 with profuse hydrogen release; samarium poisoning. CŒ 0. 0.0006 130 617 28. 5a 2. Tritium accumulatio d poisoninnan g schemee relath n i xs with (-244 * *. i~i n . c^-ffl 0.0018 0.2794 15.2 18 18. la 245« 0.0001 0.0051 369 2145 8500a relatively moderate flux of a few units by 10 but radiolysis 246Cm 0. 0.0004 1 .22 0.14 4700a and radioactivit e primarth n i yy circui te totath depen n lo d 7 power only and stay high. The fission rate of MA goes down to a 24flCm 0. 0 .0094 57 81 .9 1 .ôxio'a Cm 0. 0.0625 2.63 0.37 3.5x10 a hardly acceptable level with simultaneous e increasth f o e Totam lC 0.0019 0 . 3787 accumulation of heavier, long—lived and poorly fissioning nuclides like Pu-238, Pu-242 and Cm-248. 3. Static liquid fuel blankets deserve detailed consideration though the y hel yma o solv pt e onl ye problemsomth f eo s listed normal thermal reactor as far as thé neutron balance and thé above. burning rates of various isotopes are concerned, its effective facn 4 I .n accellerato a t subcritican i r l system plays rolf o e 14 n inertialesa s emergency shu d te fro an poindow th md f o nro t flu onls i x y 2xl0 e wel.Th l fissionine th - whic e A ar M gh vie actinidf wo e burning eliminate e probleth s f o delayem d heart of initial ATW scheme - play only secondary, almost neutron deficit only. Very high costs r caldetailefo l d negligible roleconsequenca s A . e very hazardous isotopes like discussiosucn i rolha e efficience th us e f s no because it f yo , Pu-238, Pu-242, Cm-244, Cm-24 accumulatinge 8ar . follows it s a s from above consideration, other reactor problems mort ge e complicated e accelerato.Th r problems and, especially, the difficultie f continuouo s s radiochemical processina f o g POSSIBILITA . 6 O EXPLOREYT : STATIC LIQUID BLANKETS liquid fuel mixture with short cooling times mus e considereb t d separately. The facilities with circulating liquid fuel may be called "inside-out" blankets because usual situatio vics i n e versa a- coolant circulates throug statiha c fuel. That's quite possible for the liquid fuel too and even looks preferable from many important points of view. The possibility is mentioned for heavy water electronuclear actinlde burners in [ ) so it's not quite clea y shoulwh r d fuel circulat t alla e. REFERENCES SOME PHYSICS ASPECTS OF MINOR ACTINIDE fl ] INTERNATIONAL ATOMIC ENERGY AGENCY, Evaluation of Actiniae RECYCLIN FASGN I T REACTORS Partitionin d Transmutationgan , Technical Report Serie o N 124s . IAEA, Vienna, 1982. UK WEHMAN N [2JBOWMAN ,, NucleaC.Dal .t e r Energy generatio d Wastan n e Transmutation Usin n a Accelerator-driveg n Intense Thermal Siemen, sAG Neutron Source, Rep.LA-UR-91-2601 , Los Alamos National Laborato- Bergisch Gladbach, ry (1991 !. Germany t 3 ] BELANOVA, T. S. , IGNATYUK, A. V. , PASHCHENKO.A.B., PLYASKIN, V. I. , Neutron Radiation Capture Handbook, Energoatomizdat, Moscow (1986). Abstract [4] ARTHUR, E.D. A New Concept for Accelerator Driven Transmuta- tio Nucleaf no r Waste, Fusion Technolog 0 (199y2 1 (,641-647. papee Inth r some result f studieso coren so s wit potentiae hth reducea r fo l E dSV (51 BOWMAN, C.D. Fission Energy withou tLong-Tera m High Level are described and the influence of MA recycling on these cores will be discussed Waste Stream, Contributio o ICENEnt 1 9 ConferenceS , Monterey, June 16-21,1991 . [6j IRELAND, J., CAPPIELLO, M., RIDER, W. Target-Blanket Concep- tual Design for Los Alamos Accelerator Transmutation of Nuclear Waste Concept (ATW), Ibid. [7] IRELAND, J., Overview of Los Alamos Concept for Accelerator Transmutatio f o nNuclea r Waste (ATW), Contributioo t n 1. INTRODUCTION Specialists' Meeting on Accelerator-Based Transmutation, Paul Scherer Institute, Villigen, Switzerland, March 24-26,1992. previoua n I shows studywa t ni / tha introductioe 1 / tth f Minono r Actmides (MA) CAPPIELLO , Target/Blankeal IRELAND, t e ,M. . J , t Conceptual Design for the Los Alamos ATW Concept, ibid. int largoa e sodium cooled fast reacto maie r th corns consequenceeha o st 18] KAZARITZKY, V.D., The ITEP Concept of the Accelerator Trans- mutatio f Radioactivno e Waste, Contributio o Specialistst n ' - increase the positive sodium void effect (SVE) with a simultaneous reduc- Meetin n Acceleratoi—Baseo g d Transmutation, Paul Scherer tion of the Doppler constant Institute, Villigen, Switzerland, March 24-26 ,1992. (91 KAZARITZKY, V.D., BLAGOVOLIN P.P , GALANIN A.D., A Theoreti- cal Study of Heavy Water-Minor Actlnide Blankets for Proton - to reduce the burnup reactivity loss Accelerator Transmutation, ibid. 110] TOCHENY ,, Desig L.Val t e n. Scheme Blanketf so r Actinidsfo e Since an increased SVE is an important drawback for the core behaviour in acciden- Transmutation, ibid. publie th l conditionr cta fo acceptancside d eon an n so fasf eo t reactor othee th n rso til] PIKAEV A.K., KABAKCHI S.A., MAKAROV E.I., High Temperature Radiolysis of Water and Water Solutions, Energoatomizdat, side, there is a strong incentive for SVE reductions especially for MA burning cores. Moscow, 198 Russian)n 8(I . 112] KEEPIN, C.R. Physic f Nucleaso r Kinetics, Addison-Wesley, Most of this work is part of a paper to be presented at the Tokyo Conference 1992121. Reading-Palo Alto-London (1965). 113] GANGRSKY, Y.P., DALHSUREN, B. , MARKOV, B.N., Nuclear Fission Fragments, Energoatomizdat, Moscow, 198 Russian)n 6(I . 2. STUDIES DEVOTED TO SODIUM VOID EFFECT REDUCTION

In this study the core size reduction was applied as instrument to reduce the sodium void effect Startin Europeae g th poin s twa n Fast Reacto witR thermaha rEF l powef ro cross-sectioa 360, 0MW whicf no 1 show hs i g Fi n i

Core size reduction was realized by, firstly, reducing the core height from 1 m to witm h05 constand an m t 07 core diamete andm 4 ,f secondlyro reduciny ,b core gth e diamete Thiachievem s 8 s1 wa ro t froreducin y db m m 4 numbee gth fuef ro l S/Ad san 01 en show2 g cross-sectionumbesFi a e th A S/ f pin r o r smallese spe th f no t core 01 Tab 1 Design and performance parameters of the core variants with different core size (EOEC)

core core core Explanations parameter unit core 1 2 3 4

thermal power MW 3600 2600 2000 450 inner core S/As core height m 10 07 05 07 outer core S/As core diameter m 40 40 40 1 7 radial breeder S/As fuef o l o S/An s - 375 375 375 84 control and shutdown rods no of control rods - 24 24 24 6 diverse shutdown rods no of pins per S/A - 331 331 331 271 dummy S/A fissile SVE $ 59 44 30 25 E SV total $ 55 36 22 19

fissile Doppler constant $ -1 7 -1 6 -1 4 -1 1

reactivity loss per S -89 -104 -128 -123 Fig. 1 EFR core cross-section full power year worth of control rods $ 25 23 19 31

desige Th n fou e datth rf adiffereno t core gatheree sar Tabldn i togethee1 r wite hth Explanations mai ncase performanc th value o r e fo tw thagive e e E sar t on nonlSV e yeth datr Fo a the fissile zones are voided and another one (total SVE) which also includes the void uppeeffece th f o rt axia l shore blanketh td fissioan t plenus nga m above inner core S/As Tabl showe1 strone s th whe $ g5 reduction2 goino t fissil e fro$ E th 9 gm f ne5 o SV outer core S/As from cor coro t e 1 ewhic 4 evehs ) i $ n9 mor1 o t e $ pronounce5 5 ( totaE e th SV r l dfo radial breeder S/As sinc enhancemene eth axiae th f l o tleakag casen i voidinf eo strongegs i flattee th n i r cores control and shutdown rods alternative shutdown This strong SVE reduction should offer the potential to design minor actinide con- assembly taining cores withou tlarg e levereachinE th f eo SV l coree gth s

3. STUDIES ON CORES WITH MINOR ACTINIDES

For thi sstudye parth f o t, whic bees hha n performed withi contracna t wit Euroe hth - Fig 2 Cross-section of the small core with 450 thermal MW pean Community, the cores no 1 2 and 4 have been chosen as potential MA-burners Tab 2 Performance parameters of cores increasE mucs i SV tha e A e hM etse th smalle witn % ca h 15 e r thaOn n withouA tM with different MA contents (EOEC) Thi causes i fac fissioe d th t an y thadb A n tM product s hav similaea r influence th n eo neutron spectrum so that the effect of the fission product built-up during burnup is weakenepresencA e M th o e t th e f edo du parameter MA- core core core content 1 2 4 Generally, one can conclude that the increase of the SVE at EOEC due to fissilE eSV 59 44 0% 25 MA introduction is increasing with core size reduction it changes from -t-1 6 $ in core 1 [$] 5% 65 51 3s 1i ) $ 8 4 ( valu e A th cor r t M e fo cor en witBu % i 4 5 e$ h4 1 3 2 + d corn i an $ e 2 9 1 to+ 15% 75 63 48 still smaller than the value for core 1 without MA (5 9 $) 0% -1 7 -1 6 - 1 1 Doppler constant 5% -1 3 -1 2 -08 othee Th r parameters mentione Tabl dn alsi e ar o e 2 strongl y influenceA M e th y db ($] 15% -07 -06 -05 introduction AQ per 0% -89 -104 -123 - The Doppler constant is strongly reduced full power year 5% -46 -67 -89 [$] 15% + 1 6 -1 5 -49 - The reactivity loss per full power year is strongly improved 0% 25 23 31 contro wortd ro l h 5% 24 21 30 - The control rod worth is slightly reduced but this is of no relevance for the fulfilment J 1* shutdowe oth f n requirement reducee th o t e d sburnudu p reactivity loss 15% 20 17 27

. 4 CONCLUSION In these cores 5 % and 15 % of the heavy metal content were replaced by MA with a compositiof no The study has shown that core size reduction is a very effective way to reduce the fissile th f eo zoneE SV t requiressI , however reductio,a totae th f ln o powe desige th f ri n Np237 Am241 Am243 Cm244 = 49 37 11 3, limits and the basic design data, like fuel pin diameter, are to be kept constant

whic typicahs i Ligha r lfo t Water Reactor UO2-fuel wit averagn ha e discharge burnuf po Due to their reduced SVE the smaller cores have the potential for an introduction of core e dons th abour wa se t fo i MWd/kwithoucor 0 e burns s t4 th A e, wa tM tstarMA g H - Mino% 5 1 ru Actinidepo t s without exceedin referenc e leveE th f SV lo e gth e e corTh e1 ting from the BOL state to beginning and end of equilibrium cycle with a total fuel resi- acceptability of the reduced Doppler constant of the MA containing cores has, however, dence time of 6 years The main core characteristics calculated for the end of equilibrium demonstratee stilb o lt d cycl collectee ear Tabldn i e2

interestinConcernins i t i E differenSV e poingth o e t gt th tou t burnup dependencen i REFERENCES cores with and without MA The BOL and EOEC SVE values in the three considered co- followine th re) e $ s ar n g(i /1 / W Balz et al, "Core design and safety aspects of large LMFBRs with minor actimde recycling" core 1 core 2 core 4 Int Fast Reactor Safety Meeting, Snowbird, USA (1990)

BOL 36 22 07 121 UK Wehmann et al "Design studies on LMFBR cores with reduced sodium void A M % 0 EOEC 59 44 25 effec minod an t r actimde burning" Int Conf on Design and Safety of Advanced Nuclear Power Plants Tokyo, 1992 BOL 74 58 39 ui A M % 15 EOEC 75 63 48 en ROLE OF FAST REACTORS IN REDUCTION activation product e arisinar se mosTh gt important ones from oo ecological point of view are the long-lived actinides (Pu, Am, OF LONG LIVED WASTE Cm), which determine radiotoxicit f o higy h level activity waste after abou 0 year50 t f coolino s g (storage e leveth d l )an A G TSIKUNOV, V S KAGRAMANYAN, f theio r danger doe t diminisno s h durin a milliog f yearo n s L A KOCHETKOV, V I MATVEYEV e mosth tf o dangerou e On s (from ecological poin f viewo t ) Institute of Physics and Power Engineering, elements among these actinides in the spent fuel is plutonium Obninsk, Russian Federation Its potential deleterious effect on environment about 10 times higher tha l nothe al tha f ro t actinide n totai s) lCm (Np , Am , Closed fuel cycle formatiot no e f i o repeaten e us d Abstract completely burned up and build up fission materials, means actually beginnin f wasto g e minimization procesf othei , ror s Role of fast reactors in the problem of lowering of minor actinides (HA) quantity in the fuel cycle of nuclear power is things equal, reduction of uranium demand Really all the considere e repor th e requiremen Th n i td A quantitH o t t n i y countries with developed nuclear power follow this approach in closed fuel cycl s forme i ee resultTh d f analysio s f BN-80o s 0 the fuel cycle organization Nevertheless, up to now under reactor possibilitie A quantitM r fo s y reductio O HAYAP t a Kn consideration were only fuel cycles with extraction of U and r furthefo u P r recycl n reactori e n thiI s s case becausf o e are discussed in the report The recommendations are given on 241 z37 choice of actinide extraction coefficient values during the long-lived =<. -emmiters Am(Ti/2=433 yr ) and Np(Ti/2=2 14 chemical reprocessing of spent fuel The coefficients are the 108 yr ) build up one should put forth the task of out of 8 9 0 - m C d an p 0 9998U - r ,fo m A d an followin u P r fo g reactor fuel cycle time reduction It is reasonable to put a question what type of reactor e mosf th io secologicall t e suitablus r fo ye dangerous plutonium' The question is particularly sharp for plutonium Developmen d constructioan t P witNP hf o nfas t reactord an s with high content of isotopes,having mass number M>239 their further progress are tightly connected with the closed fuel cycle creatio s necessari e concept I th t nonlr no yfo t y Conducted evaluations show e thaspecifith t c (per f nucleao r fuel breeding realizatio r solutiofo t bu f morno n e GW(e)yr)radiological e long-livehazarth f o d d actinidee th n i s urgent task for the present - environmental problems spent fuer fasfo l t reactor s lesi s s than thar thermafo t l reactors Th,e conclusio d uraniu an s i trur nbotX fo eHO m h Considerabl n eoperatioi quantit P NP f no y to-day, using fuel once-through fuel cycle produce radioactive waste, disposal of e mosth t f o importan e whicon s i ht task f globao s l character Calculations a showe s a ds i possible thaus t i to t e Fast power reactors can play an important role in solution of nuclear fuel in fast reactor not only plutonium but neptunium, the problem americium, curium In the paper the above-mentioned role of fast reactor in When considering this problem one should have in mind that the nuclear power system is being considered The analysis of the benefi f o transmutatiot s compare(a n o onct d e through the question shoul e concep e baseth b d n o dt tha n loni t g term cycle) is exhibited after relatively long tine interval perspective, under condition f nucleao s r power system normal Proceeding from this fact suc a requiremenh t determinine th g organization any increase of the environmental radioactivity efficiency of minor actinides transmutation can be formulated level mus e excludeb t d e totath l- conten f actinideo t l stageal f t closeo a ss d fuel cycle should amount 0 1-1 0 % as related to their quantity o realizT e conceptth e t leasta , o problemtw , s shoule b d produce n onci d e through cycle solved organizatio a reliabl f o n e buria f o fissiol n products The chosen time interval may correspond, for instance, to under surveillance ( for hundreds of years), the reactor life-time, or life-time of the nuclear power - organizatio f produceo n d actinides burning system This conditio s beeha n n formulated e basmainlth e n o y of requirements to the depth of actinides extraction from a matte s f A o factr , radiotoxit f o fissioy n products spent fuel in reprocessing (actinides content must be reduced remains higher than that of natural uranium for about the by 102-103 times) first 500 years Radiotoxicit l minoal f ro y actinide n thii s s fuel cycle Besides e procesth n f i nuclea,o s r reactors operatiot no n after 500 years of storage would not be higher than only fission e samproductth e t e a formedar tim st ebu , radiotoxicity of initial natural uranium Conducted evaluations showed e cas,f mino o th tha en ri t e maximuTh m life-tim P witNP hf o BN-80e 0 reactor assumed actinides recycle their content in the fuel cycle arrives at . Storagyr 0 6 e e tb otim f faso e t reactors spent fuel before the steady state level. Minimum of minor actinides will be in reprocessing assumed to be 3 yr. Efficiency of actinides (Pu, the system consisting of fast reactors only. In this case ) extractioCm e reprocessin, th NpAm ,n i n f o speng t fuel total quantit f minoo y r actinide n uranium-plutoniui s m fuel assume e 0.9b o 8t d (conservative approach). entering reactor e leveth f o 0.7%t lsa wile .b l Minor actinides quantit e totath n li y closed fue le bascyclth e n o e Three possible situation O "MAYAKP e sit f th e o e t ar "a s of fast reactor will not be higher than 1% of minor actinides, being considered. which coul e produceb e dnuclea th n i dr power system during 1. t BN-80commissionedno s i 0 . Transuranium elements from 150-200 years of thermal reactors operation having the same VVER-440, BN-35 d BN-60an 0 0 reactor e storear s d durin 0 yea6 g r powe d oncan re through fuel cycle. after year 2000 and then directed to the long-time burial. . BN-802 0 reacto s commissionei r yeae th r e n i th d200 n i 0 In reality fast reactors will represent only certain part traditional concep f o uranium-plutoniut m fuel cyclee Th . e nucleaoth f r power system. Under these conditions minor breedin go calleS rati . s 1 i equado o t "dirtyl " plutonium actinides content in the fast reactors fuel cycle will surpass produced by thermal reactors VVER-440, as well as extracted the 0.7% value. Obviously, such a reactor should be adapted to froe spenth m e tBN-80 th fue f 0o l reactor cor s assumei e o t d f o actinide e thus e s bearing fuel. Inclusio f o minon r be used as its fuel. In the case of "dirty" plutonium actinide e traditionath n i s l fuel t somi give eo t sspecifi c deficiency it will be covered by "pure" plutonium produced in features in subassemblies management and in the core fast uranium reactors and in the radial blanket of BN-800 neutronics. reactor. Minor actinide n thii s s t extracteversiono e ar t n bu d buried together with fission products. 3. The BN-800 reactor to be commissioned in the year 2000 e traditionath If l BN-800 fast reacto s r i userfo d using uranium-plutonium fuel mixed with minor actinides. actinides incineration, their quantite fueth l n i wily e b l Maximum content of minor actinides amounts to 4% of heavy restricted by the condition of initial core characteristics atoms in the very beginning, when there is a lot of minor retention, and first of all those accounting for reactor actinide e storagth n f i speno se t fuel. Then, when minor safety e minoTh . r actinides contene fueth l n e i shoultb t no d actinides in the thermal reactors storage will be exhausted, higher than 3-4% of total quantity of heavy atoms. Already their content in the fresh fuel of BN-800 reactor is this quantity appreciably affect e reactoth s r physical diminished. characteristics, especially sodium void reactivity, and requires partial refurbishin f o fueg l handling system, All data concerning estimates of different nuclides designee "standardth r fo dX fuel MO " . quantities on the site PO "MAYAK" by the year 2060, for three considered situations, are presented in the table i. o increasT e efficiencth e f actinideo y s burning woule b d reasonabl rejecto e t a fertilzae as u e material producing high activity actinide coursthe of irradiationin es . Further TABL . QUANTIT1 E F O RADIONUCLIDEY O P MAYA R THRET FO A KS E developmen f thio t s idea result n concepi s a cor f o et witn a h DIFFERENT OPTION F FUEO S L CYCLEG ,K inertial diluent such as, for instance, zirconium, or zirconium based material. Nuclide Nuclide quantit f 206o 0s a yyea r quantity 2000y______Optiof ao s n 1____Optio2 n Options e presenth y n traditionada Ca t l fast reactors solve th e practical task of bringing down minor actinides quantity in 237Hp 550 720(720) 785(785) 60(45) the fuel cycle? 2«iAm 420 2450(2450) 2040(2040) 370(170) 2«3Am ISO 150(150) 630(630) 195(45) To answer this and other questions, related to substantiatio a necessit f o n r expediency(o y o transmut)t e th e 24-iCm 40 5(5) 25(25) 100(10)

minor actinides e authorth , s attempte e baso obtait th d e n o n 238pu 165 110(110) 40(20) 275(75) of their analysis of possible use for these purposes BN-800 O "MAYAK"reactoP e sit th f o e n o r . 23Upu 17600 17600(17600) 17230(1750) 17470(1730)

210pu 4370 4400(4400) 3810(840) 3770(830) To the date of BN-800 reactor commissioning appreciable 2«1PU 2350 165(165) 420(90) 410(90) quantit f o actinidey s i suppose se b accumulate o t da s a d Ol resul f o spent t fuel reprocessing from thermal uranium Comment Nuclide quantities which go for final disposal are CD reactors VVER-440 and fast reactors BN-350 and BN-600. indicated in brackets O1 Analysi f theso s e figures make t possibli s o concludt e e TABLE II POTENTIAL HAZARD INDEXES OF ACTINIDtS IN WASTES AFTER o 10=>YEARS, LITERS OF WATER 1 In spite of small difference in the total quantity of nuclides in versions 1 and 2 the latter gives an essential Grouf o p Option of fuel cycle reductio f o nucliden s which need superlong-time burial actinides 1 2 3 Quantity of plutonium to be buried in the version 2 is an orde f magnitudo r e lower tha ne versio th tha n i t1 n Np 0 03*lQi° 0 04*101° 0 002*101° 2 More radica s solutioi l f ecologicao n l probleme th n i s e thase tn herca t onle quantit no e th versioy On f o 3 yn Pu 100*10io 14*10i° 14*10io plutoniu e n burieb ordea o f s magnitudt o i mrd e less than i n the versiot quantitiebu , 1 n f neptuniuo s d americiuan m e b o t m Am 170*1010 140*101° 12*10io n burieordea e f magnitudo rar d e less tha n version2 i n d an 1 s s wela l Recycl A resultH f o en reductioi s f theio n r quantity Cm 0 07*101° 0 3*101° 0 14*101° in all the system (storage+BN-800 ) Exclusion is ?**Cm Total 270*101° 154*10io 26*10io Wether such a minimization of actinides quantity is sufficient or not Answer to this question can be obtained by the way of comparison of the waste to be buried radiological all minor actinides increase e scalth sf o lowerine th f o g toxity with that of natural uranium Here the quantity of VVER-440 disposed wastes PHI up to 10 times, but for the natural uraniu s meani m t whic s user energwa hfo d y production variant the value of the actinides PHI in the wastes after by VVER-440 reactors and by BN-800 reactor during 60 of its 103years , wilhoweverbe l , approximatel o ordertw yf o s operation magnitude higher then ideal objective - 0 4*10 liters of water 10 According to estimates the quantities of fission products and artificial actinides accumulated due to VVER-400 reactors e analysiTh s shows e thaideath t l objective coule b d operation tile yeath lr 200d BN-80an 0 0 reactor durin0 6 g achieved only unde e conditiou th P r d an f lowerino nm A f o g years amount to 100t and 30 t respectively This process of (the main investors into PHI) discards value during energy production will result in disappearance in the nature reprocessing almos o ordertw t f magnitudeo s o e t froi ,% 2 m of t uraniuabou 0 13 t m Radiological y nuclidhazaran f s o di e r t extractio fo n do s A e w % n0m 02 .C coefficient d an p N f o s accepted to evaluate by the potential hazard index (PHI) It need special tightening n i compariso p u s n e witth h is determine a quantit s a d f wateo y r (air), whic s needei h o t d coefficients adopter calculationsou n i d , namely8 9 0 , dilute the specific content of a radionuclide to its tolerable consentration level Summing up aforesaid, one can affirm that in near future the fast reactors could solv e probleth e f limitatioo m o t p (u n The potential hazard index of 130t uranium is evaluated as demand of the Earth s radiation status quo conservation) of 0 4 1010 liters of water (taking into account daughter long-lived and especially dangerous part of the nuclear power product f uraniuo s m isotopes f actinideo I PH ) s burier fo d wastes like actinides different scenarios of fuel cycle development are expedient to compare with each other approximatel 3 year10 y s later, whet i n will be difficult to rely upon artificial barriers and when relative contributio f actinideo n n potentiai s l hazard index will reach maximum In table 2 are listed the calculated PHI of actinides in wastes after 103years for three above-mentioned options

It is clear from table 2 that putting into operation the BN-800 reacto e above-mentioneth f ro inty an o d option f fueo s l cycle organization leads to lowering of level of wastes PHI in comparison wite cas f th hfinao e l disposa f VVER-44o l 0 reactor wastes Under recycle BN-80th n i e0 reactor plutoniums i onl t i y possibl o timeo reductw t e I s PH e Involvement inte recyclth o e ACTINIDE TRANSMUTATIOE TH N I Administration Building NI Security Center —, ADVANCED LIQUID METAL REACTOR (ALMR) Switchyard Fuel Service Facility C.L. COCKEY Optional Fuel GE Nuclear Energy, Cycle Facility^ San Jose, California, United State f Americso a

Reactor Abstract Maintenance Facility The Advanced Liquid Metal Reactor (ALMR) is a US Department of Energy (DOE) Remote Shutdown Turbine Facility & Radioactive Generator sponsored fast reactor design based on the Power Reactor, Innovative Small Waste Facility Facility Module (PRISM) concept originate Generay db l Electric currene Th . t reference Control Building Steam Generator design is a 471 MWt modular reactor loaded with ternary metal fuel. This paper Assembly Facility discusses actinide transmutation core designs that fit the design envelope of the RVACS ALM d utilizRan e spenfue R s startua l LW t p materia makeupd an l . Actinide Stack transmutation may be accomplished in the ALMR by using either a breeding or High Security Boundary Reactor Facility burning configuration. Lifetime actinide mass consumption is calculated as well as (Below Grade) changes in consumption behavior throughout the lifetime of the reactor. Impacts on FIG. 1. ALMR Power Plant (3 Power Blocks) -1440 MWe system operational and safety performance are evaluated in a preliminary fashion.

CONTAINMEKr DOME . INTRODUCTIO1 N

REACTOR VESSEL AUXILIARY COOLING objective Th Advancee th f eo d Liquid Metal Reactor (ALMR) Program, under SYSTEM INLED TAN sponsorshie th e U.Sth f .po Departmen f Energyo t develoo t s i , competitivpa e EXHAUST STACKS reactor system aime t improvina d g safety, enhancing plant licensabilityd an , simplifying plant operations nationa.a s Thii , s lGE activity progray b d le , m involving wide participation by U.S. industries, the U.S. national laboratories, universities, and international contributors. One of the goals of the program is to develop a standardized desig licensee nb than certifiedd ca dt an curren e Th . t reference design (see Figures 1 & 2) is a 471 MWt modular reactor, with nine modules constitutin gplan144a e 0tMW (Ref. [1])reference Th . e core desig breedea s ni r

core utilizing ternary metal fuel under developmen Argonny b t e National Laboratory GRADE (Ref. [2]). This paper discusses actinide recycle (or burning) core designs that fit the LEVEL design envelop ALMe th utiliz d f eRo an e irradiate fue R startus a l dLW p materiad an l for makeup.

Actinide recycle (frequently referred to as "actinide burning" or "transuranic burning") has long been an intriguing concept associated with closing the nuclear SEISMIC ISOLATOR fuel cycl improvind ean g waste management concepe Th . t involves transmutation BEARING (31) or fissioning of the higher actinides (transuranic [TRU] isotopes) to shorter lived fission products. The primary incentives for transmutation of these TRU isotopes is o eliminatt s manea f theo y possibls ma e fro ultimate mth e waste streaa mvi processing spent fuel and to recycle the TRU as an ALMR fuel source. LWR irradiated fuel contains up to about 16% minor actinides (MA-neptunium, americium REACTOR curiumd an ) relativ totao et l TRU, dependin burnun go shows pa e Tabln i Th . eI SILO current U.S. LWR spent fuel inventory averages about 7.7 w/o MA. FIG. 2. ALMR Reactor Facility O3 TABLE I. COMPARISON OF ISOTOPIC COMPOSITIONS IN SPENT FUEL PO turbine generator A sodium-filled intermediate heat transport system (IHTS) provides normal heat remova e reactoth r fo rl throug n intermediata h e heat Réf. Equil Existing Current Future exchanger (IHX) and transports the energy to the steam generator The reactor R LW R LW ALM R LW S U facility is seismically isolated to provide high margins in seismic capability Inventory 33 GWd/t 50 GWd/t Plutonium The reactor, show pool-typa n Figuri n s i , e2 e containin coree o gth tw , Isotope) s(% intermediate heat exchangers, four primary electromagnetic (EM) pumps, and Pu-238 0.9 1.2 1.5 3.4 mtenm spent fuel storage Containmen leakageprovides w i t lo a y db , pressure- Pu-239 75.4 60.1 58.2 553 retaining boundary, which completely enclose e reactoth s r coolant boundary Pu-240 21.3 233 26.6 20.1 Containment consist lowea f so r containment vessel surroundin reactoe gth r vessel Pu-241 1.5 10.4 85 117 uppen a d r doman e ove reactoe rth r closure Pu-242 09 4.5 52 94 The current reference reactor fuel is metallic uranium-plutonium-zirconium Minor Actmides (%) alloy ferntid an ,claddin e uses ci th allor 9 assembld fo ygHT an y duct minimizo st e Np-237 265 47.8 42.0 428 Am-241 603 39.6 48.3 408 swelling associated with high burnups Metal fuel provides competitive fuel costs Am-242M 33 01 01 01 and excellent negative reactivity feedback during transient overpower and loss of Am-243 77 93 80 12.7 cooling event alternativn A s e oxide cor undes ei r development with international Cm 22 26 1.6 3.6 partner reference Th s e breeder corheterogeneoua s ei s arrangemen f blankeo t t drived an r fuel, wit controx hsi l rods Refueling occurs afte month4 2 r f operationso , Total TRU (%) with one-third of the driver fuel replaced each cycle The fuel is not shuffled, but Plutonium 974 923 83.2 834 blanket assemblie shufflee sar d each outag burnue Th e p reactivity swin smals gi l Minor Actmides 26 77 108 166 dunng the cycle for the equilibrium core, making axial swelling of fresh metal fuel TR UHeavn i y assemblies the largest contributor to the cycle reactivity swing Meta) (% l 259 088 102 176 Reactivity contro r normafo l l operation f startupo s , load followingd an , shutdown is accomplished by bank movement of the six control rods Each control rod absorber assembly has sufficient worth such that any one of the six can shut down the reactor to cold subcntical conditions Positive reactivity addition by The purpose of the work described in this paper was to focus on 'safe inadvertent withdrawal of the control rods is limited by electronically positioned actmide burners' The ALMR is a reactor system that uses passive safety features, mechanical rod stops and the actmide burning cores should not have a significant impact on these features The work was done in the context of this system and an evaluation was For safety margi evene th f losn o i t f primarso y coolant flow without scram, madimpacte corese th f th e o n so ' safety features (Ref [3]) three gas expansion modules (GEMs) have been included at the penphery of the GEMe cor holloTh e es ar w assembly ducts, whic ope e bottoe h ar flo th o nt t wma There are two basic processing concepts, aqueous and pyroprocess, to like the dnver fuel but are closed to flow at the top The GEMs are filled with vessel achieve actmide recovery from fuespen r R recycl ALMe fo l LW th t o eRt Aqueous cover gas before insertion into the core, and this gas is compressed as the GEMs processe pyroprocessed san usee b recover r dy fo sma actmidef yo s from botR hLW fill with sodium The GEMs are designed such that the sodium level inside the and ALMR spent fuel and recycle to the ALMR The pyroprocesses for processing GEMs at normal operating pressures is above the top of the active core When LW ALMd Ran R spen reference tth fuee ar l e processe ALMe th n Ri s program pumpin primare th loss g i n i t y pressuree systeth d man insids dropga e ee th th , Unlike the aqueous process, the pyroprocess does not separate Pu from other TRU GEMs expand d expelan s s sodiue mos resultanth e f o tTh m t void nea core rth e A specific Pu enrichment from LWR spent fuel therefore carries a specific MA penphery increases neutron leakage (provides negative feedback reduced an ) e sth enrichment In this scenario, to continuously consume MA for a reactor's lifetime, it net reactivit core th ef yo necessars i continuouslo yt y consumu eP An ultimate shutdown system is included as a means of bringing the reactor to cold subcntical conditions in the event of a complete failure of the normal scram . PLAN2 T DESCRIPTION system This device is manually actuated after the core has been brought to a safe stabl ecore statth ey ethermab l feedbacks The ALMR plant utilizes nine reactor modules arranged in three identical 480 MWe power blocks Each power block consists of three identical reactor modules, eac stean h ow wit ms hit generator , that jointly supply stea singla mo t e 3. CORE DESCRIPTION ALMR REFERENCE Two major types of ALMR cores were studied; radial heterogeneous breeder homogeneoud san s burners reference (seth r e FigurFo e . breedere3) , LWR spent fuel fissile materia selectes reference wa l th s da e startue p th fue r fo l ALMR and will be recycled in the ALMR until the TRU content approaches the equilibriu ALMRn valuU a r mTR efo . This material contain pluu sP s associated minor actinides from the LWRs. Only startup fissile material is required for the reference breeder ALMR. This typ f coreo e wilfissile b l e self-sufficien wild lan t consume LWR minor actinides loaded into the startup core plus two reloads.

If the ALMR is used to reduce the total amount of LWR actinides in waste destined for disposal, fissile material and other TRU isotopes can be transmuted or burned in ALMRs. Various amounts of TRU can be burned depending on the SHIELD DRIVER FUEL 66

desigALMe th f nRo core orden I . consumo t r largese eth t amount witU hTR f so REFLECTOR INTERNAL BLANKET 30 metal fuel, a burner core was designed. The approach used in this case to modify an ALMR from breedea changburneo a t o s rt core wa re th e fro mheterogeneoua s EXPANSIOS GA N MODULE CONTROL 6 arrangement with blanket assemblies to a homogeneous arrangement with no Q RADIAL BLANKET ULTIMATE SHUTDOWN blankets blankee Th . t assemblies were replaced with fuel assembliee th d san radial blankets were replaced with an extra row of shield assemblies in order to compensate for the effect of the deleted radial blanket assemblies on shielding. The core height was decreased to provide decreasing conversion ratios and increasing TRU consumption rates. Impacts on system operational and safety performance were evaluated in a preliminary fashion. The power level remained at 471 MWt, and the designs were evaluated according to current ALMR design and safety parameters.

The driver fuel assemblies are divided into two enrichment zones, lower enrichment in the core center and higher enrichment in the outer core rings. This was done to reduce the radial power peaking. Core heights ranging from 135 cm (53 in - reference) to 89 cm (35 in) were studied for actinide burning benefits and safety characteristics. Cycle lengths were adjuste accoundo t pear fo t k fast fluence limits on the fuel. The reference core has a 24 month cycle, while the 135 cm burner core has an 18 month cycle and the 89 cm core has a 13 month cycle. 2 10 SHIELD 0 DRIVER FUEL (LOW ENRICH) 30 2 4 REFLECTOR 4) CONTROL 6

corm c e9 A8 heigh t burner design doe t meeno s t ALMR design/safety EXPANSIOS GA N MODUL6 E 66 ULTIMATE SHUTDOWN 1 criteria due to high burnup reactivity swing ($12) and peak linear power constraints defines ,a metar dfo l fue ANLy b l . This cor considerees i beyone b do t d DRIVER FUEL (HIGH ENRICH6 6 ) TOTAL- 253 the envelope of ALMR burner core design. A 102 cm (40-in) core is marginal in performance due to its burnup reactivity swing of 11$, although this value is FIG.3. Core Layout Comparison probably within the bounds of control design. An added benefit of reducing core heighreductioe th s i t f sodiuno m void worth t thiminoa bu ,s i r safety benefit which show Tabln i e II). Tabl I showeI resulte sth f equilibriuso m diffusio burnud nan p offses i othey b t r more serious safety concerns. Greater core heights give larger calculations with no fuel recycle. This produces "snapshot" core values calculated safety margin larged san r actinide storage masses. approximatel lifetimA M d e an reacto U t Cyclya TR r . consumptioe3 n totals should be base n coro d e lifetime evaluations rather than "snapshot" intervalss a , MASU TR S. 4 CALCULATIONS consumption changes with changing composition core th es heighA . reduces i t d and the uranium-238 content is reduced, the TRU consumption rate is increased The TRU and minor actinide consumption rates were calculated for specific whil totae eth l core heavy metal inventor decreaseds yi consumptioU TR . s ni o> cores. As discussed above, the reference ALMR has a net annual TRU gain (but variable with the time the fuel spends in the reactor, but tends to approach an CO MA loss) wherea burnee sth r core consumptioU sTR t shone wa n (refe valueo t r s equilibrium valuf reactoo d e en near lifee rth . TABLE II. ALMR CORES EVALUATED FOR ACTINIDE RECYCLE

Reference Actini de Burning Cores Startup ALMR Core Activ e Core Height (cm) 135cm J35_ JSL. _8i Breeding Ratio 1.09 077 9 06 0.65 Cycle Length (months) 24 18 15 13 Burnup Reactivity Swing ($} -23 -91 -115* -119' Peak Linear Power (kW/m) 312 282 354 39.4 Sodium Void Wort) ($ h 585 463 263 1.51 TRU Enrichment FIG. 4. Minor Actiniae Mass (wt% in U-TRU-Zr) 30 15/19 17/21 20/24 TRU Inventory Every isotope exhibits a unique characteristic behavior in the actmide (kg/core 471 MWt) 2196 1867 1550 1424 burner designs Table III shows total loaded and consumption values for both the TRU Consumption Rate corem c s5 fro13 me th ORIGEN d an 10m 2c 2 calculation burnp N t steadils sou y kg/yr -69 387 511 569 throughout the plant lifetime A total of 87% of all Np loaded into the 102 cm core 0 4 3 3 1 2 3 -0 % inventory/yr ove actmidyear0 m c 5 r consumeds i 5 s e13 burnee th r ,fo r wit% valuha 74 f eo three isotopem Th eA s exhibit very different behavior Am-24 consumes 1i botn di h 'Doe meett sno ALMR design criteria cores (79% in the 102 cm core) Overall, the total amount of Am-242M is slightly increased Am-24 consumeds 3i , with values ranging from 16-23 varioue Th % s isotopes of Cm have differing behavior, but in general during the early cycles Cm is ORIGEN2 calculations were completed for two cores in order to show the produce onls i t i y compositio e d lateth dan n i r n lifetime (afte years0 4 r m C ) thay an t effect f continuallso y loading LWR-generate ALMinte U oth dTR R actmide burning isotop consumes i e amoune producem Th dC f o t smalds i termn i l f masso d san cores The transuranics were processed and reloaded into the core for 42 cycles large in terms of decay heat Pu is consumed throughout the reactor lifetime (64- The calculations included refueling and makeup with LWR TRU, as well as k-mfinity 73%) masd an s normalization One-group ALMR cross-sections weree overlaith o t n do standard ORIGEN2 fast reactor cross-section mont8 1 n hA cycles , three-batch corm c e2 load10 e largesa Th r total amoun large s f actmideit o t o t r e sdu homogeneous core with a 135 cm core height (conversion ratio 077) was required makeup, and also consumes a larger mass of actmides over the same compared to a 15 month cycle homogeneous core with three batches and a core amount of time 66 and 62 percent of the total TRU introduced to the ALMR is height of 102 cm (conversion ratio 0 69) Both changing TRU consumption rates consumed over 50 years in the 102cm and 135cm high cores, respectively differine th d an g behavio f eaco r h actimde isotope were evaluated durin corga e (3 3 and 3 1 MT/reactor) 68 and 63 percent of the MA introduced are consumed in lifetim cycle2 4 f eo s the two cores, respectively (385 and 350 kg/reactor) These values show pure consumption without the additional benefit of storing TRU or MA in the ALMR A Figur showe4 changine sth gcor m concentratioc e2 10 e th n f i eac no A hM e ar U loadee TR ar f f thesA o do M T inte f eM o o on cores 5 g k d 0 an ,tota 56 f o l t core-averageno e Thesar e batc d on e an r h dvaluefo resulte e ar Th s s loaded into these reactors concentratio f eaco n h minor actmide appear approaco t s equilibnun ha m value Neptunium (Np) is converted at a steady rate The rate of production of curium Figure 5 gives curves that demonstrate the storage and consumption (Cm )steada levelo t f y of svalue thed an n, gradually begin decreaso st e capabilit cor m f thesc yo e 2 doee10 corese botTh sh loaconsumd dan e more Amencium (Am slowls )i y consume batce Th dh naturgrape th f eho clearly shows corm c e 5 ovecrossovee 13 year0 tha Th U 5 re nth TR s r point where eth consumes i tham A t d dunng each cycle t alsbu ,o shows that loades mori m eA t d a additional makeup required by the 102 cm core is greater than the greater TRU the beginning of each cycle than is transmuted The percentage of MA in total TRU loaded into the 135 cm core is at 22 years The lower conversion ratio of the 102 stays approximately even at 10 7%-10 9%, due to the heavy TRU makeup required corm c e lead greateo st r consumptio corm c ne2 amount 10 lifetime e th r th f sfo eo for this core The reference heterogeneous ALMR, not requiring LWR TRU makeup, would have a substantially higher rate of change in the MA concentration Figur showe6 consumptioU sTR n rate resultm c s 5 fro coreso 13 m tw e th , (from 10 7% to 2 6%) actmid actmidm ec burnin2 e10 e burningth cord coree gan ecorTh s esho w TABLE HI. TOTAL TRANSURANICS CONSUMED IN 50 YEARS

* ' -S l-;v:^,ifiîi£:cm'<^vL- - »?y."ï ; 135cm * *Total'°' Total Total Total Loaded Consumed Percent Loaded Consumed Percent Isotope (Kg) (Kg) Consumed (Kg) (Ko) Consumed

Np-237 193 168 86.7 168 141 83.9 U-238 9376 4997 53.3 11569 5372 46.4 Pu-238 114 55 48.6 93 32 34.4 Pu-239 2471 1802 73 2211 1342 60.7 Pu-240 1293 741 57.3 1103 507 45.9 Pu-241 250 177 71 216 145 66.9 Pu-242 262 132 50.3 218 96 43.9 Am-241 319 254 79.4 271 201 74.1 Am-242M 4 -1 .- 4 -2 -- Am-243 66 10 16.3 53 3 4.8 StorageTRU 5. FIG. Cm-242 0 -28 108.2 0 -19 - Cm-243 0 0 - 0 0 - Cm-244 5 -15 - 5 -12 - Cm-245 3 -2 - 2 -2 -- Cm-246 0 -1 - 0 0 -

TotaA M l 566 385 68.0 501 308 61.4 TotaU lTR 4956 3293 66.4 4342 2429 55.9

TOTAM LH 14332 8290 57.8 15912 7801 49.0

parallel curves, but different consumption rates. The sharp changes in the first ten ConsumptionMA and TRU Rates6. FIG. burnue th o t f Np-237p percentago e e th yeardu s A e . sar f Np-23eo 7 (and other actinides) in the core is high relative to U-238, the burnout of Np-237 will have a large r rate effec f changth e o n o t f consumptioeo earle th yn i cycles . 5. IMPACT F ACTINIDSO E BURNINE ALMTH RN GO SYSTE M Figure 6 shows an interesting view of actinide consumption. The consumption rat f minoeo r actinides changes throughou core th te lifetime firse tTh . LWR actinide recycl currene th n ei t ALMR reference configuration will havea two cycles use LWR TRU for reload fuel. The following cycles use processed small impact on reactor operation. The greatest impact would be the additional ALMR fuel and LWR TRU makeup. The marked change in the graphs is the point shielding requirements necessary for the greater amount of MA. Actinide recycle recycle begins. The rate of actinide consumption is changing due to the changing usin ghomogeneoua s actinide burner core does have system impacts safetw Ne y. isotopic composition of the core. Also, the majority of the actinides loaded into the designs would be needed to accommodate the much greater reactivity swings (e.g., core are from the initial core. As the actinides are processed and core) reloadedm c 9 .8 e th ,r therfo $ e 12 i corsm d c ean 5 13 e th r fo $ 9 a higher probability of burnup due to the greater amount of core exposure. TRU OJ consumption is approximately level, with some small changes due to changing The greatest e reactoimpacth n o tr relate o shieldint s f neao g r core O1 composition. components. difference th o t e corn ei Thi du es s i assembl y typereference Th . e Œ> heterogeneous core has a row of radial blanket assemblies between the fuel and ACKNOWLEDGEMENTS en the reflector assemblies, while the homogeneous cores use only fuel assemblies Neutron producee sar outee th n dri rin f fuego l assemblie highea t sa r rate than i Work performed unde contracE DO r o DE-ACO3-89SF1744N t 5 the radial blankets in the reference heterogeneous core It is necessary to increase shielding for these homogeneous cores by converting a possible fuel assembly row into shielding Near core shielding would increaseneey b o d- t facto5 a 1 y df b o r orden i 5 counteo t r2 neutroe th r n damage incurred withou f shiel o extre th tw daro REFERENCES assemblies KwantW 1PatelR M , A Hunsbed,Magee M S ALM P U d " ,R an t Sodium Actmide recycle impacts fuel handling equipment design because of the Cooled Reactor Design and Performance," ANP-92 International Conference, nee handlo dt e fuel assemblies with higher decay hearadioactivitd an t y This si Tokyo, Japan, October 25-29, 1992 higheo t e rdu concentration f minoso r actmide r referencsfo e ALMR startud pan burner ALMR reload fuel than for ALMR reference equilibrium fuel It is anticipated 2 C E Till and Y I Chang, "Progress and Status of the Integral Fast Reactor that fuel handling equipment including transport cask accommodatn sca e these (IFR) Fuel Cycle Development", Proc International Conferenc Fasn eo t Reactors higher decay hearadioactivitd an t y levels without major design modifications Total and Related , Kyoto 6 Fue 1 l p ,Cycles Japan , I l Vo , October, 1991 core decay heat is less than a startup core, so there are no impact on RVACS 3 C L Cockey and M L Thompson, "ALMR Potential for Actmide Consumption," The impacts of actmide recycle on fuel processing and fabrication are ANP-92 International Conference, Tokyo, Japan, October 25-29, 1992 expecte relativele b o dt y inconsequential because these function designee sar o dt be operated and maintained remotely Remote/automated operation and maintenance of fuel processing and fabrication are required due to high radiation levels of plutonium (Pu238 and Pu240) as well as significant levels of fission products (pnmanly rare earth isotopes) increased levels of minor actmides should not require additional shielding, with the possible exception of transition areas, such as loading and unloading casks and fuel assembly transfers It will be necessary to evaluate process impacts, such as the effect on melting point temperatures, process heat loads, volatility, alloying othed an , r attributes which could affec processee th t s and/or the product These impacts are not presently known but, based on the charactenstics of the minor actmides, are expected to be minimal and controllable

For the homogeneous burner cores, there will be no blanket assembly fabrication, as all assemblies are fuel assemblies Even this change has no major impact on the fuel cycle facilities because for the pyroprocess, blanket assemblies were assume fabricatee b o d(plutoniumt e th n di ) fuel a cycl n i et facilitno d yan separate facility Two fuel ennchments are required in the core which will result m a minor complicatio n fuei n l processin d fabricatioan g n compare o onle dt on y ennchment

6. CONCLUSIONS

Overall, actmide recycle (consumption) appears promisin a wast s a ge management approach as well as providing fuel for ALMR startup and deployment for power production Processing spent fuel provides opportunitn a r removayfo f o l actmidee th selected an s d fission products fro waste mth e stream wels a ss a l potentially improved waste forms Actmide burning cores designe r safetdfo s ya well as consumption are certainly feasible within the ALMR design envelope The ALM verRs i y flexibl weld ean l positione providdo t e actmide recycl botn ei h modes, breede burned ran r Recycl fuef naturaa eo s i l l characteristi ALMe th f cRo MINOR ACTINIDE CONTAINING FUELS o desigT a nr fuespecifie fo l d irradiation conditions certain physicad an l FOR TRANSMUTATION PURPOSES physico chemical data have tobe known [10] They include datn ao effective neutron cross sections to estimate the burn up evolution in the fuel with time, L KOCH COQUERELLEM , K RICHTE, R mechanical properties such as Young's module, thermal creep plastic Institut r Transuraniuefo m Elements, deformation, Commission of the European Communities, chemical physical data suc s phasa h e diagrams, melting point, oxygen Joint Research Centre, potential, thermal expansion heat capacity vapour pressure From such data possible fuel compositions can be derived which need further Karlsruhe testmgof their technical feasibility

Abstract In the last 20 vcars Ihc Institute for I ransuranium LlcmciUs is developing MA containing fuels for 3 FUEL TESTING fas eaclori t s Ther severae ear l incentive eeyeli o st e these material fasn si l reactors increased energy ineration prolonged bur reactivitp nu t leasy no lasedueet i tbu d radioloxicit dischargef yo d waste f he basic knowledge to design fuels on an oxide and alloy b isis has been obtained from experimental I he scope of information required to license an irradiation experiment varies studie d simulatiosan n experiments ruel pin f differenso compositionA M t s have bee d wile nan b l wit chosee hth n reactolicenss it d rean authorit dependt I y fuen so l mastestee b o st d irradiated in Pill ''«MX First results of the post in adialion examinations are given and the computer model used to simulate the irradiation behaviour Certain features of a fuel under irradiation can be simulated in a laboratory fuel restructurin thermaa n gi l gradien electricay b t l heating, fuel cladd interactions compatibilit fuef yo l with coolant 1 INTRODUCTION Fuel swelling, radiation induced cree s welp a othes a l r radiation dependent Durin e lasth gt twenty year e Europeath s n Institut ransuraniu1 r fo e m fuel parameter determinee b n sca d irradiation onla n yi n experiment Thereforf i e Elements studie possibilitiee dth recyclo st e minor actimde fissioo st n reactor] s[1 their influence to the irradiation behaviour can not be sufficiently predicted a pre- Ther economin a s ei c incentiv recoveo t e r these element energr sfo y generation ni test with a smaller quantity (maybe in a different reactor) will be asked by safety the driver fuel of a fast reactor I he gam in extra electricity compensates the authorities Suc n exampla h e beloth s wi e mentioned irradiation experimenn i t additional cost reprocessinf so havy he fued e1 gan beel ) mak1 n1 p assesseeu e b o dt KNK II where a pre irradiation was made in FR 2 4 8% [2] to 5 2% [3] of the electricity generating cObt in case of the less favourable During post irradiation examination the irradiation behaviour is revealed PWR The gam in extra energy by MA fission will amount to about 4% of the energy Routinely the following tests are foreseen in our institute generatio consequentld an ] 1 n[ y increas e breedineth g gai s welna s prolona l e gth visual inspectio outef n o inned ran r claddin examinatioy ra X g n burn up reactivity of the fuel [4, 5] Moreover the recycling of MA would save 15% of profilometry, Eddy current, gamma and neutron scanning, uraniu n argumena m t less relevan o FBR't t s becaus f moro e e efficienf o e us t optica scannind an l g electron microscopy EMPA, uranium in this reactor type than in LWR Recently the aspect of the long term fuel density, porosity metallography, ceramography, oxygen potential, reduction of radiotoxicity as an ecological advantage of MA recycling attracts more elemental and isotopic assay by IDA MS laser ablation ICP MS attention than the economic benefit since the electricity generating cost of a FBR compett ye t no e n witca R h thosLW a f eo 4 EXAMPLE FUELA M F SSO 2 FUEI DESIGN Fable I summarises the fuel types that have been tested under irradiation, of which the post irradiation examinations of some of them have not yet been The renewed interest in partitioning and transmutation brought concepts concluded The l consisyal mixef o t d oxide e basiTh s c data neede r fuedfo l design other than the fast reactor into discussion, which are summarised in [6] The fuel have been presented earlier [10] First result pose th tf so irradiatio n examinations types proposed enclose aqueous solutions slurries molten salt alloys, nitrides and, were published [11] From these and recent results it can be concluded that the MA last but not least oxides Principally our institute is, open to engage in any fuel containing fuel of the homogeneous type doesn't behave significantly different from development provide appropriate dth e fundin gives gi n Presentl continue yw e with e standarth d fuel under irradiation Henc ea lon g term irradiation experimenf o t r studieou mixen so r basedZ oxided d an wild alloy] an sl[7 1 loos[8 k int desige oth n mixed oxide in a cooperation with CEA, France is foreseen in PHENIX reactor targetc ofT minod san r actimde inern si t matrix] ][9 In the test irradiation of an (Uos Amo 5X^2 m the thermal flux of FR 2 it researce Th mixef ho d oxide fuel limitee s b wil application t a no lo R d t FB a n ni became apparent thae stronth t g resonance shieldin ge sufficientl b coul t no d y Sincnucleae th s ei presentlR r reactoLW e choicf yth ro e transmutation studies with accurately predicted During irradiation only 15% of the expected linear power this reactor type offer the opportunity to study the impact of P & T on an existing increase was observed due to the strong radial giadient of Am transmutation in the fuel cycle and at the same time keep the relevant technology alive for a possible pellets ) (Bi[121 g ] Consequentl e stronth y g temperature gradient leado t s application of P & T in a more effective FR economy circumferential crack formation what made the outer region of the pellet to decay CD TABL I E MACONIAIMNG FUI-1S PIS Hn UNDLR l ABI L II RATIO Ot CAICUIAIIU IO LXPLRIMFNI AI ISOIOI'L CO IRRADIATION CONCENTRATIONS F OR (Vo 75 PUo « AM,, U2)0)

Fuel composition Reactor Burn up Ibolope ILNDI 3 1 NDl/li vnvi KhKINIt/ll 1 > hl MNK/JMi - ht-KINlt/J2 1 I SANIXDKSI SAN1XDK8I SANUÜJKSI l)j| .'1)31)21 ! SANIXÜK2) (Amo 5 Uo 5)0-2 FR 2 A!»Pu l 2593 1 2137 09762 08345 1 2397 oo')Û2 KNK II 1% 2J9pu 1 0023 1 0280 1% 09865 10029 09718 (Uo73Puo2sAm0o>)0/: KNKII 2-tOpu 10371 1 0343 1 0024 09933 1 0169 KNK II 1% 24lpu 1 1843 1 0467 1 1168 1 0058 1 2602 PHENIX 64% 242pu 1 1107 10742 1 0601 09878 1 1821 (Uo74Puo24Amo2>Ol96 PHENIX 64% W 'Am 1 0133 10176 1 052*1 10639 l 0294 (Uo5sNpo4s)O2 PHENIX 4 5% 242mAm 1 4532 1 3674 1 040-4 09078 1 2718 (Uo6Npo2Amo2)Oi93 PHENIX 4 5% (Ne by 3 dm

TABLI EH FABRICA1IO E IRRADIATEN B FUE A DATM 0 1 LF AO N DI PHENIX <

-010 > > (U()74l ll(, 4 (L(17Jl 4 u> u (UofiNpo , 1 ucl compositions 11 US5^Pll4&)O> Np()u)>0, Am0u>>0> Amo 2)0 >

thermal lri.ilmi.n C/20 t 40 h argon 400 C/2h argon C/lhdi|,u0 40 n 400C/lhaigon particle-P ofGS , 400"C/2hair 400C/2hair 400 C/4h ur 400C/4hair 850 C/2han 900C/4hair C/35 n i 86 h 900C/25lian

005 850C/2K Ai5%ll 1 050 C/2h Ai 5%ll. 8h5 C/4h \i5%ll> 900C/3h At5%ll2 pressing I/em4 6 ' 1 5 4 4 SbtAm bill in b 1 t/em- fret density) 1 1 % 6 8 5 58 (W. 1 1) I'll.' 11) 56 4% 1 0 sintering I 162I i U "M i C/bA h l(>2UC/61I «I i A i I()2() h ArS'/fclCh l 1620"C/fah Ai5'illj pellet 0 5 363 ±0(152 mm 5 41710029mm m 7 IU 1 1 1 m S SI 5 543410013mm as 10 pellet density 51087 9 ) VII 968107% II) 95 1 ±0» , II) 959 10%T1 D , FIMAVbu . 0/M I 973 1 957 1 99b 1 928

1 1 Rang G d 242CH f buran eo p mnu formation (includin s daughtegit r product) diffractioy Xra n 5 457815 5 4570 ±5 5 454 ) 15 5 473515 parameter 5 4b95 61 along the radius of a (U() 5 Amu 5)02 pellet irradiated m FR 2

The fabrication dat f fuelo a s irradiate n PHENIi d e summarisear X n i d into pieces From this result t appeari s s doubtfu f fueli l f suco s a hhig h 241Am Tabl I [14 eII e limitin ]Th g paramete oxide th f eo r fuel theis i b r oxygen potential, content can be irradiated in a LWR for the purpose of transmutation adjustee b whieo t sinterin y s d(b hha argon gi n atmosphere containin suco t ^ hH % g5 bure achievep Th nu KNKIe th dm I irradiation gives experimena w n lo s twa a level tha a reactiot a coolann N wit e s minimiseth hi t ratiM 0/ o dSine) w lo ea in Tabl eI hence conclusio lone th gn ntero m material irradiation behaviout no n rca lower e thermath s l conductivity too e irradiatio sAmnth ,o (U .a <>)Of no 2 becomes

be drawn So far no irregularities were observed The main emphasis was given in impossibl othee Th e r fuel parameter elose s ar standare thos o et th f eo d fue xl Worth comparing the experimentally obtained isotopic eomposition with the predicted to mentio l supportege s tha i ne th t d precipitation s use(GSP avoio wa t dy ) an d one which has been presented in 113] The agreement for most of the nuclides is aerosol formation as observed by using mixed oxide powders The fuel specimen sufficient (Table II) reache pointe s W/c dlineaa Ü A m t 40 earlie o dou rt powee e rang0 th rth 30 f n i ero TABL V ESUMMARI MICHOSIUUCF TH F YO ANAE 1UR 1 YSI SPENF SO T REFERENCES MA FUELS IRRADIATED IN PHENIX [I] KOCH , "Formatio L , d Recyclinnan f Minogo r Actinide Nuclean i s r Power (U„74l'u„ 4 (Um Am,, > tUo74l'"0« Stations", Handboo e Physicth n Chemistrd ko san e Actimdei>th f yo 4 l Vo , I'lll Npoo'lO- Npo )O> (Uo;, N Pius»' Am()2)0) (1986)457490 [2] BLOMEKE.J O ,et al .CEC.JRC EUR6929 (1980) 380/325 W/i m I74/273W/CH1 206/283 W/cn i 380/325W/cm [3] SCHMITT.E.et al,CEC,JRC,SA/l Ü5 03 83 13(1984) [4, "Cor ]l a BALZet e Desig , Safetd W , nan y Aspect Largf so e LMFBR's with Cladding Minor Actmide Recycling", Proceeding of the 'Int Fast Reactor Safety ijmml 3294 3293 i i03 3293 Meeting', Snowbird, Utah, August 1990 Ar/rm 0 58 055 1) 84 055 [5] PILATE , "Som S , e Aspect Transmutatioe th f so Minof no r Actinide Fasn si t wait Uueknc-is t m m I 0 448 0449 1)454 0446 Reactors", Proceedings of the Workshop on 'Partitioning and Transmutation of Minor Actinides1, CEC, JRC Karlsruhe, October 89, (KOCH, L , WELLUM, Fuel cladding interaction 1284R 9EU ) (1990Ed , )R A r|um| 19 3 3 19 [6] KOCH , "Statu L , Transmutationf so " IAEA Specialists' Meetin 'Übn go f eo FBR's for Actmide Transmutation' Obninsk, September 1992 Fuel [7J KOCH, L, "Review of Pabt and Present Status of Partitioning and i fuel |min| 2829 2840 2841 2826 Transmutation", IAEA Advisory Group Meeting, Vienna, October 1991 Ar/i |%| 4 1 59 49 43 [8] INOUE, T , et al , "Characterization of Fuel Alloys with Minor Actinides", ANS Winter Meeting, San Francisco November 1991 Kuel restructuring [9) KOCH L , "Evaluation on P & Tin the View of MA Fuels', Proceedings of the Workind 2n g Group Meetin ' Target n g o Fuelb' d san , CECKarlsruheC JR , , /one 1 uulei nngli/i0| 1 0 bl> 1 0(>b 1 058 1 067 Jun4 2 e 3 publishede 2 199b o 2(t ) /une 2 cquiaxial grain 0 <>6 0 5(1i 2l 0 5 5 0 2 2 0 8 (5 I 8 4 0 7 6 0 growth [10] KOCH, L , et al , "Designing a Minor Aetmide Containing Fuel for Trans mutatio a Fas n i tn Reactor WinteS AN " r Meeting n FranciscoSa , , /one 3 columnar or clone, i 0 50 0 1 1 0 0 0 6 2 0 0 0 u 2 2 0 0 48 0 11 t( d gram growth November 1991 /on centra1 e l void 0110 00 0 1 1 0 00 [II] PRUNIER, C , et al, "Transmutation of Minor Actinides Behaviour under Irradiation of Americium and Neptunium Based Fuels", Proceeding of the eiieulareraek (position r/iy 083 0 98 0 l>8 089 'Int Con n Faso f t Reactor d Relatean s d Fuel Cycle R 91'""F s , Kyoto, Secondary phases October 1991 112] KOCH, L.TUSR 36, p 60(1983) metallic precipitates [13] BROEDERS, I, et al , "1 he Measurement of Fast Reactor Irradiated Minor (0=1 urn) /one I 3 /one 1 3 /one 1 3 /on3 1 e Actinide Comparisod san n with Neutron Physics Calculations" t ConIn , n o f pin5 = ) 0 ( Aim ) /one 3 'Nuclear Data for Science and Technology', Juhch, May 1991 [14] RICHTER TUAR, ,(1986K 1 8 ,p ) eeiamic phase [15] COQUERELLE M TUAR n 7(1 n 00n (0-1 urn) A.IK i (0= 1 urn) Ame 2

behaviour did not deviate strongly from the standard mixed oxide except for the He formation stemming from 242Cm deca leadind yan highe a o gl s pressur ga r e inside the pin I he recent results of the post irradiation examinations are summarised in Table IV [15] The fuels with the high MA concentrations did not reach the same linear powe d consequentlan r y were expose loweo dt r bur p Thinu s explaine th s o othedifferenctw re fuelsth o t e , whic e morar h e similae standarth o t r d fuel 05 Further result fuen so l clad interactio isotopid nan c compositio comparison n(i o nt CD predictions) will become available in the future PHYSICS CONSIDERATIONS IN THE DESIGN OF i automaticallthano s i t y recycle n witdca h techniques being developed strippee b , d froe mth initial wast d returne e reactoan e expensth th o t t da r l simultaneouslo e y recyclinn a g LIQUID METAL REACTOR TRANSURANIUR SFO M increased proportio rare-earte th f no h fission products ELEMENT CONSUMPTION The higher radioactivity levels associated with the increased fission product carryover to the transuramc product stream doe t significantlsno y complicate fuel refabncation, whic dons hi e H. KHALIL . HILLR , E FUJITA, WADD , E remotely becaus presence th f eo f higheeo r acumde repeatedle th n i s y recyclee th d d fuean l Argonne National Laboratory, incomplete fission product separation inherent m the basic reprocessing method Argonne, Illinois, The hard neutron spectrum favors destruction of transurarucs by fission over creation of United States of America higher transuraruc neutroa vi s n capture Thus relative th , e concentration f minoso r actimdes do not build up to levels at which their radioactivity becomes intolerable The spectrum Abstract hardness also mitigates the adverse poisoning effect of the recycled portion of the rare earth fission products managemene Th f transuramo t c nuclide liquin si d metal reactors (LM consideres i R) s d based e Integrath f o e l Fasus e otnth Reactor (IFR) concept Uniqu fueR le ER cyclfeature e th e f wito s h respect to transuramc management are identified These features are exploited together with the hard 3. RECYCL SPENR LW TF EFUEO L spectrum of LMR's to demonstrate the neutroiuc feasibility of a wide range of transuramc management options ranging from efficient breedin o purt g e consumption Core physics e aspectth f o s development of a low sodium void worth transuramc burner concept are described Neutrorucs In addition to the main IFR reprocessing and waste treatment steps (i e those concerned with performance parameter d reactivitan s y feedback characteristics estimate r thifo ds core concepe ar t recycle of LMR discharge fuel and blanket assemblies), pyrochemical processes are also being presented developed and demonstrated for the purpose of extracting the transuramc species from LWR spent oxid concentratinr fo e fued an l g these transuraruc metallia n si c form suitabl introductior efo n inte oth IFR fuel cycle The goal of the process development is to recover at least 99 9% of the LWR processee dischargTh s R se beintransurarucLM n i g e studieus r fo sd involv decladdine eth e th f go 1. INTRODUCTION spent LWR fuel pins by suitable mechanical and chemical means, and the subsequent pyrochemical decomposition of the spent fuel into (a) a product stream containing the transuramcs and the major The desire to reduce the long term radiological hazard associated with the disposal of spent portion of rare-earth fission products, (b) a uranium-rich component suitable for storage and potential nuclear fues motivateha l d numerous studie severan si l countrie wayf o ) g destroRefo 6 st e s( 1 s e yth 0 wast( fuelR d ean , LM stream r o R processee source sb th thaLW n s f futura eca to e recoveo dt us e r transuranium products of fuel irradiation Recent efforts m the U S at Argonne National Laboratory residual acumdes and then converted and packaged into forms acceptable for geologic disposal [6-8] have focused on evaluating the potential for accomplishing this with liquid metal reactors (LMR's), particularly metal fuel system s e Integrabaseth n do l Fast Reactor (IFR) concept These It should be noted that the elimination of fission products from the transuramc output of the evaluations have addressed bote recyclth h f self-generateo e d actimdee consumptioth d an s f o n above processes is not required, because this output is designed for introduction into the IFR externally produced inventories of transuranics, e g the transuramc species in spent LWR fuel Core electrorefirung step, which accomplishe e requisitsth e degre f fissioeo n product remova e higTh hl physics aspects of these evaluations will be discussed in this paper radioactivit e transurampuntw th lo f yo d an y c product limit s diversiobot it e riss f it hth o k d nan attractivenes weaponr sfo s application othee th n r O hands effectee ,th d separatioe bule th f th k o f no uranium is important because it helps preserve the compactness of the electrorefirung process by CONCEPR IF E TH T . 2 allowing this process to deal with only about 1 to 2% of the total heavy metal in the spent fuel concepR featureIF y e includ] Ke th [9 t f metallie so eth c fuel form (U-Pu-Zr compace th d an ) t Currently, three different pyrochemical separation concepts ,"sal e referreth ts a transpor o dt , ' t d inexpensivan e techniques being develope r fuefo dl reprocessing (pyrometailurgical processd an ) 'magnesium extraction", and "zinc magnesium ' processes, are being investigated with the objective of refabncation (injection casting) Neutronic feedback mechanisms and other inherent physical identifying and further developing the most promising concept phenomena provid neutroiur efo c shutdow decad nan y heat remova accidenm l t sequences, resultinn gi a high degree of passive safety Fuel burnup levels in excess of 15 at% are being demonstrated for the U-Pu-Zr alloy in EBR-II experiments Current designs [10] are self-sufficient with respect to fissile . TRANSURAM4 C MANAGEMEN SELF-SUFFICIENTN I T LMR's mass, providin r favorablgfo e fuel cycl emodifiee b economic n ca yielo t dd dlarga an s t exces ne e s fissile production if desired Conversely, favorable neutron economy can be sacrificed by reduction of In a self-sufficient LMR transuranic losses by fission are compensated by transuramc ferule conten d coran te geometric spoilin t consumptio permio ne t g e th t f transuramo n c species (Pu 239) breeding, allowing for sustained power production with only fertile material (e g depleted U) instead of their net creation through breeding supplie makeus da p Tabl ecompareI transuraniue sth m isotopi f dischargeo x cmi d fueR l froLW ma (once through) and a 1200 MWt LMR (equilibrium/recycle, based on the IFR concept) The relative The IFR fuel cycle has a number of attractive features in connection with the management of concentration vase th tf majoritso f minoyo r actirude substantialle sar dischargeR y smalleLM e th ,n i r man-made actimdes They stem fro followine mth g while the Pu-239 proportion therein is greater Thus, a self-sufficient LMR using the recycled transuraruc sdischarg R fro mstartus LW it s ea p fissile source would addition i , producino nt g power 1 The main reprocessing step (electrorefirung) directly provides for the separation of the bulk of ove operatins it r g life, resulequilibriun a n i t m fuel composition with reduced long-term radiological transuraruce th s from uraniu fissiod man n product e smalTh sl portion (-1% f transuraruco ) s toxicity owing to the reduced proportion of higher acumdes Of particular importance is the reduced Table I COMPARISON OF LWR AND LMR TRANSURANIC ISOTOPICS Tabl I eI COMPARISO EQUILIBRIUF NO M CYCLE CHARACTERISTIC t 120R SFO 0MW ACTINIDE RECYCLE CONCEPTS Ratio of LWR 1200 MWt 1200 MWt Decayy 2 3 ( ) Fissile Pu and MA MA LWR LWR LMR LMR to LMR Self-SufficienR LM t Pure Burner Pure Burner Discharge (3.2 v Decav) Discharge Decayv (2.0 ) (2.0 y Decay) No of Batches 4 1 1 Fractiof no Cycle Length, days 365 254 453 Transuranic Mass Capacity Factor% , 89 80 80 Np-237 535-2 540-2 6 16-3 619-3 87 Driver Average Bumup, 104 111 241-7 1 12-7 408 8 27 111 Pu-236 6 12-8 MWd/kg Pu-238 9083 1 01-2 6273 6 19-3 16 Pu-239 0503 0508 0747 0747 068 Driver Peak Bumup, 140 150 158 Pu-240 0201 0 199 0200 0200 10 MWd/kg Pu-241 0 157 0 134 2 13-2 1 962 68 Driver Peak Fast 1 30 393-2 388-2 3.51 1 76 Pu-242 849-3 849-3 46 Fluence, ICPn/cm2 863-3 Am-241 344-3 251-2 694-3 29 Driver Peak Linear 130 136 145 Am-242m 1 144 1 11-4 5 11-4 5 07-4 022 Power, kW/ft Am 243 251 2 2482 1 743 1 74-3 140 EOC-BOC Ak Swing, % 027 121 -433 Cm-242 131-3 973-6 3564 2365 041 Cm-243 855-5 786-5 1 93-5 1 85-5 42 Dnver HM Loading, 2770 4040 3040 Cm-244 632-3 5 52-3 796-4 744-4 74 kg/y 5 15 4 C5 m24 5084 1 634 1 63-4 3 1 %U 772 0 0 Cm-246 642-5 631 5 472-5 472-5 1 3 %Pu 224 646 375 MA/Fissilu eP 0 137 0 172 0022 0024 72 «MA 04 354 625 MA/Total Pu 0099 0 124 0017 0018 69 Blanke LoadingM H t , 3910 - - Np-237/MA 0591 0490 0368 0343 1 4 kg/y Am 241/MA 0038 0228 0415 0478 048 Am 243/MA 0278 0225 0 104 0096 23 %U 100 Np chain 0214 0213 0034 0034 63 %Pu 0 - %MA 0 MA = sum of minor actimdes Makeup Feed, kg/y 466 352 346 Fissil Pu-23= u eP Pu-249+ 1 Np chain = Np-237 + Am 241 + Pu-241 %U 100 0 0 %Pu 0 876 0 %MA 0 124 100 fractio f Pu-241o n , whose decay product (Np-237) dominate e long-terth s m hazar e energTh d y productio long-terd nan m waste radiotoxicity reduction benefit realizee coste ar s th t sda associated HM = Heavy Metal with the reprocessing of LWR fuel and the increased amount of fission product waste MA = Minor Actimdes

S TRANSURANIC CONSUMPTIO PURN I E BURNER CONCEPTS Illustrative design and performance characteristics for both cases are summarized in Table II It should be emphasized that the the fuel compositions being employed in such cases are quite exotic in Characteristic conceptR LM f o ss designe t consumptione r fo dexistine pr f no g transuramc containing essentially no uranium, and that the assembly/core configurations using such compositions inventories have also been investigated Taken to an extreme, no ferüle U-238 is utilized in these (and constrained by a beginning of cycle excess reactivity requirement) are nonconventional Thus concepts onld an ,y transuraruc startue useth e r sar dpfo sourc e inventorth s makeuf a e o d yan p material while these concepts are feasible neutromcally, the establishment of their overall viability requires e actimdeaddeth o dt s recovered from self-discharge de neutroru fueTh l c feasibht f suco y h pure extensive technological efforts discussioa , whicf no beyons hi scope dth thif eo s paper burner" concepts, which maximiz transurame th e c consumptio r uni f npe energo t y productions ha , been verifie t n 120LMR'di 0MW s case th whicbotn ei r l hfo transuram hal R c LW specie e th n si consumptioe Th f transitâmeno LWR'n si considerabls si y less attractiv onlt No ye doe e sth discharg case whicth n ei r usee fo h ear d s fue donla an minoe l yth r actimd usee ear d) speciePu o s(n thermal spectrum favor the buildup (in proportion) of the higher actimdes, but it also renders most of --J the minor actirude species as poisons whose accomodation requires greater fissile enrichment, a Table 111 EQUILIBRIUM-CYCLE PERFORMANCE PARAMETERS ro thermal system fueled entirely wit LWR-discharge hth e isotopi f minoo x rcmi actirude t feasiblno s si e neutromcally based on the fundamental requirement of cnticality Moreover, the technological difficulties of adapting current aqueous reprocessing methods to compositions nch m higher actirudes 3 53 0 Breeding Ratio are likely to be severe EOC-BO7 1 C 4 - Reactivity Swing K %A , Average Discharge Burnup MWd/kg 82 6 8 8 Ato m% . CONCEPTUA6 L DESIG TRANSURAN1R NFO C CONSUMPTION Peak Discharge Burnup focue Th f receno s t conceptua s beel e developmenha desig nth ] [7 nL effort AN f cor o t t a se MWd/kg 1186 concepts for net transuramc consumption whose design parameters, unlike those of the pure burner 6 2 1 Ato m% concepts, remain within the bounds of the m-place development program For example, limits on Peak Linear Power, W/cm transuramc heavenrichmenn i % y wt metal) 8 (2 t , linear powe kW/m)0 r(5 , discharge bumu at%)5 p(1 ,

BOC 495 2 2

and peak fast fluence (35xl0 n2 cm) were observed A key additional objective of the core EOC 462 development was to achieve a low (near-zero) sodium void worth Finally, the core concepts were develope that o fissilds ne t e breeding f desiredi , , coul readile db y achieve alternativn a s t da ne o et Power Peaking Factor transuramc consumptioa Such flexibilit r increasefo y d breedin f externas o achieve e gwa us ly db 4 6 1 C BO blankets to minimize the effect on reactor design and to mitigate the inevitable increase in sodium void 3 5 1 C EO worth that would occur with enhanced internal conversion Peak Flux, 10'5cmV A 157t (6050MW MWe) design concept satisfying these objective d constraintan s s wa s 4 48 BOC develope featurey Ke f thi ] o s ds[7 concept transurams it n i , c consumption ) eliminatiomode(a e ar , n 2 0 5 EOC of internal and external blankets to minimize transuramc production by U-238 capture, (b) adoption of Peak Fast Flux, 1015cm 2s ' a pancaked core shape to reduce breeding and void worth, and (c) incorporation of a non fueled BOC 3 68 central regio annulan a n ni r core configuration Detailed evaluation f thio s s core configuration were EOC 3 77 dischargR LW f o e e transuraniuus carrie e t baseth ou dn do m isotopic e reactivth s a s e fuel feed component, limited analyse f thiso s core wit equilibrius hit m (infinite-recycle) fuel composition have Peak6 9 Fas 2 t Fluence, lO^cm2 also been performed but are not described here in detail The use of a flat, annular core shape enabled the achievement of a low void worth ($0 16 for HOC voiding of the enure core and upper plenum Mass Flow, kg/y region) while simultaneously satisfyin assumee gth transuram% d28 c enrichment limi particularn I t , Heavy Metal 5,790 the annular geometry mitigates the central power peaking tendency without recourse to enrichment Transuramcs 1,492 zoning thereby enablin e enrichmength linead an t r power limite satisfieb o t s d concurrently with 8 95 Fissil u eP meeting the criteria of transuramc consumption and near zero void worth Net Loss, kg/y Heavy Metal 508 The breeding ratio of 0 53 permits the net consumption of 234 kg of LWR transuramcs on an 4 23 Transuramcs annual basis The fissile Pu component of this net loss (218 kg/y) would, with repeated recycle, Fissile Pu 218 decrease somewhat as the nuchdes that are less likely to fission increase in proportion Thus some change corn si e performance woul observee db d with repeated recycl transurame th s ea c isotopix cmi gradually equilibriun shifta o t s m recycle distribution characteristi e corth ef co geometr fued yan l cycle parameters Preliminary calculations suggest that the coolant void worth of the equilibrium core is nearly $10 greater because of their presence as dilute species in the reprocessed LWR discharge composition Finally, e flath t core shape resulte radiath n i sl expansion reactivity coefficient being significantly more Additional performance results are summarized in Tables III and IV Of particular note are negative, and the axial expansion coefficient being less negative, than in conventional designs For the the burnup reactivity loss of 4 2% Ak (5 7% Ak without rrudcycle replacement of central absorber same reason, and because of the higher control rod worth dictated by the greater burnup reactivity loss, assemblies), whic s significantlhi y greater thae nominallth n y zero reactivity swing achievabln i e controe th dnvelmd ro l e expansion coefficien s muci t h more negative tha r conventionanfo l cores fissile self-sufficient design e smalTh s l Doppler coefficien e attributeb n hare th ca t d o dt spectru m Dynamics analyses [11] utilizing the computed feedback parameters have demonstrated significant and the low U 238 concentration contributions to the Doppler effect by the transuramcs are minimal passive safety margin unprotecter sfo d los f floreactivitso d wan y insertion accidents VOIW LO DTabl WORTV eI H ACTINIDE BURNER RFACTIV1IY COEFFICIENTS REFERENCES [I] MUKAIYAMA , YOSHIDAT , , GUNJI H , , 'MinoY , r Actinide Transmutation Using Minor BOG' HOC Actimde Burner Reactors ', Proc International Conference on Fast Reactors and Related Fuel Sodium Void Worth$ , Cycles, Vol II, p 19 6, Kyoto, Japan, October, 1991 Core 1 78 285 Plenum 220 -2 70 (2) , BOBROV"Useal t Fasf e so , t B ReactorS , USSe th Radir n Rsi fo o Isotopes Productiod nan Total -043 0 16 Minor Actimdes Burning", Proc International Conference on Fast ReaUors and Related Fuel Cycles, Vol II, p 19 4, Kyoto, Japan, October, 1991 Doppler Coefficient, 10Tdk/dT Flooded Core [3] PRUNIER, C, et al, "Transmutation of Minor Actimdes Behavior of Amencium- and Fuel 073 088 Neptunium-Based Fuels Under Irradiation", Proc International Conference on Fast Reactors Structure1' 030 035 and Related Fuel Cycles, Vol II, p 19 2, Kyoto, Japan, October, 1991 Voided Core Fuel 036 0.46 [4] BUCCAFURNI , LANDEYRO A , , 'MinoA P , r Actimde Burnin Therman gi l Systems", Proc Structure6 021 025 1992 Topical Meetin Advancen go Reacton i s rp 1-301Physics , l l , CharlestonVo , , South Carolina, March 1992 Axial Expansion Coefficient, $/cm Fuel 1 50 -1 19 ] BOWMAN[5 , "Transmutatio D C , Existinf no g Wast e, Specialis' t Meetin Accelerator-Driven go n Fuel and Structure11 1 37 -1 01 Transmutation Technology for Radwaste and Other Applications, Saltsjobaden, Stockholm, Radial Expansion Coefficient, $/cm 1 43 -1 48 Sweden, June 1991 Control Rod Dnvehne [6] HILL, R N , et al, ' Physics Studies of Higher Actimde Consumption in an LMR, ' Proc Intl Expansion Coefficient, $/cm 0778 -0414 Conferenc Physice th n f Reactoreo o s s Operation, Design, 83 Computationd I an , p , l l Vo , Marseille, France, April 1990 Effective Delayed Neutron Fraction 3 50E-3 3 47E-3 Prompt Neutron Lifetime, s 2 17E7 2 47E-7 ] HILL , "LM[7 N R , Concept Transuramr sfo c Managemen Sodiuw Lo n mi t Void Worth Cores", Proc International Conference on Fast Reactors and Related Fuel Cycles, Vol II, p 19 1, 'BOC values are calculated for critical configuration, with primary rods inserted 24 cm Kyoto, Japan, October 1991

'Values reflect perturbatio totaf no l structure, clad estimate e effect b (volum % n 63 ca s s da e fractiof no ] , HILL[8 'EvaluatioN R , f Reactivito n y Coefficient r Transuramfo s c Burning Fast-Reactor Designs", Trans Am Nucl Soc , 65,450(1992) clad in the total structure) of the total structure effect [9] TILL, C E, CHANG, YI, "Progress and Status of the Integral Fast Reactor (IFR) Fuel Cycle Development", Proc International Conference on Fast Reactors and Related Fuel Cycles, Vol I, p l 6, Kyoto, Japan, October, 1991 [10] BERGLUND , Performancal t e , Safetd RC , an e y Desige Advanceth f o n d Liquid Metal 7. SUMMARY Reactor", Proc International Conference on Fast Reactors and Related Fuel Cycles, Vol II, p 111, Kyoto, Japan, October, 1991 The IFR fuel cycle has a number of attractive features for the management of transuramcs These features, when successfully demonstrated e exploiteb n ca , d along wite harth h d neutron [II] CHANG, Y I, et al, "Passive Safety Features of Low Sodium Void Worth Metal Fueled Cores spectrum charactensüc of LMR's to achieve substantial flexibility in transuramc management opüons, iBottona m Supported Reactor Vessel", Proc International Conferenc Fasn o e t Reactord san ranging from efficient breedin destructioo gt burnen ni r concepts Relate , Kyotod2 Fue6 1 ,lp Japan, CyclesII l , OctoberVo , , 1991

CO ENGINEERING ASPECTF SO TRANSMUTATION USING FAST REACTORS (Sessio) n3

Chairmen

L.A. KOCHETKOV Russian Federation H. SZTARK France RADIOWASTE TRANSMUTATIO NUCLEAN I R REACTORS Iß H wafer

A.N. SHMELEV, G.G. KULIKOV, V.A. APSE, V.B. GLEBOV Moscow Institut r Physicefo Engineerind an s g D.F. TSURIKOV, A.G. MOROZOV I.V. Kurchatov Institut f Atomieo c Energy Moscow, Russian Federation

Abstract The burnup of minor actinides may be realized both in fast and 10 thermal reactors e principaTh . l differenc f minoo e r actinides burn-un i p therma d fasan l t reactors concerns wite neutroth h n balance n thermai : l reactor the neutron balance is negative with neutron consumption and iO otherwise in fast reactor the neutron balance is positive with production of excessive neutrons. Applying of fuel containing minor actinides in FBEs leads to degradation of safety characteristics, namely to decreasing of 10s effective fractio f delayeo n higo dt neutronhd positivan f sBe e sodium void iO 10 1000 WO 10iD* effec f reactivito ts supposei t i e o upgradt th yf d e Be valuth ef o e additio e fissilth f o ne nuclides characterize y higb d h fractio f delayeo n d FIG. 1. Index of biological hazard (m water or air/GW(e)-year) as a function of waste decay neutrons in fuel composition. The decreasing of sodium void reactivity may 3 be achieved by use of potassium as a coolant (or potassium alloys Na-K, time (PWR type reactor). K-Pb).

The two important aspects »ay be emphasized in the transmuta- Neutron consumptio n i minon r actinides burn-u n i fasp t tion problem o-f long—lived radioactive waste ( RAM ) produced in and thermal reactors. reprocessing o-f nuclear reactor spent fuel. The -first one is re- The burn-up of minor actinides may be realized both in fast e burn—uf - late minoa f ) s do- pm C r wit e , actinideth hAm , Np ( s and thermal reactors. Fission cross-sections of minor actinides determining RAW potential hazard during a thousand years or more l-l.havthresholde . 0 th V e energ Oe Me th n yi sn regionca t i o s , ( see Fig. la >. The second one concerns with the transmutation o1 e fissioneb fasn i dt reactors n thermaI . l reactor minoe - th sac r long-lived -fission product, mainl) s yFP ( sSr-9 d Cs-137an O , witi. e transmutetinideb n ca sneutroy b d n capture well—fissioth o t e - duration o-f significant radiological hazard up to several centu- nable nuclides ( Pu-239, Cm-245 ) with -following burn-up. As a re- ries (see Fig.la). Here it may be noted the -following point. As it sult the long-term potential hazard of minor actinides ( thousands shows i t Fig.la n e potentiath a l gettinW hazarRA f o dg into organs of year convert) s s intshort-tere th o hund ( s -FP me hazarth f o d of digestion ( IBHwater-Index of Biological Hazard defined as a wa- red f yearo s . ) s ter volume needed for RAW dilution to safe concentration /!/ ) is The principal difference of minor actinides burn—up in ther- determined by the mentioned FPs up to several centuries. During d fasmaan lt reactors concerns wite neutroth h n balance thern i : - the next thousands of years IBH is determined by the actinides. mal reactor the neutron balance is negative with neutron consump- Potential hazard of RAW inhalation is determined mainly by acti- tion and otherwise in fast reactor the neutron balance is positive nides during all time periods ( see Fig.lb ). with production of excessive neutrons. This difference is demons- TABLE 1 ACTINIDE BURNING AND NEUTRON BALANCE balance of actimdes irradiation in FBR is characterised by posi- 00 tive value, the excessive neutrons are produced. Thus, irradiation Np-23O 1O f 7o nucle therman i i l reactor durin e 1OOth gO day> rf s PVft Fßft FBR. suite fission i d f 33.o n 1 nucle daughtes f Np-23o it i d 7an r pro- 4 fldt. Fiuians 33.1 33.* J-f.8 38.0 ducts, but in sum with U-235 fissions for neutron generation the +31.0 '91.0 +23.S total fissions quantit 91.6s i y . Irradiatio Np-23O 1O f 7o n nuclei in FBR during the 100O days resulted in 39.8 fissions. So the ac- 5&S 0 0 tinides burn-up in PWRs leads to the generation of increased quan- (26.9) (°L tity of FPs radioactive decay of Si—9O and Cs-137. trons in thermal reactor it is needed to occur the 53.5 fissions Neglecting the value of FPs neutron capture cross-sections and of U-235 accompanied by the conversion of 24.6 U—235 nuclei to neutro estimatee nb y lossema t i ds that O.142 neutron r fisspe - totae U-236th lo s ,consumptio U-23f o n 83.s 5i 1 nuclei. Neutron sion mus providee tb r thesfo de radionuclides transmutation. This TABLE 2 YIELD OF FISSION PRODUCTS PER FISSION 235U K, f) s includinFP and s w -froFP ne loadin gm e th otheg r reactors, it is important to note that quantity of excessive neutrons gene- IBM, tn^HiO rated in FBR with mentioned breeding parameters may be sufficient t/ucêiJe •for transmutation o-f produced long-lived FPs. now it must be 27.7 0.055 0029 S6 to noted that Sr—9O and Cs—137 have low values o-f neutron capture 30. 0.0« 0.031 407/0 cross—sections e leveneutrof th o- lo s R blanke,n FB -flun s i i xt insufficien effective th r fo et transmutatio f theso n e nuclides. l.sto 0.009 0.009 525 IO Long-live s witFP dh rather high neutron capture cross—sections " 2110' 0062 0062 6A5 catransmutee nb blanken i d e transmutatioFBRf o tTh . f o nSr-9 0 «n 87. OOgfoOOSiï U9 IO and Cs—137 mus realizee tb high—flun i d x reacto Fig.e se , ( r2) 65/0* oooof 135 its concept will be discussed below. It is important to note that separate transmutation in two nuclear reactors may be characte- 65K> 0004 38210 e similath rize y b rd total neutron balanc consideren i s ea d sim- To M 0205 Of 4? 1 90 10" plified schem f long—liveo e s transmutatioFP d Fasn i nt Reactor captute sulstfxitm isikpe. Blanket.

is the minimal estimation because of the supposition of conti- nuous purification of irradiated nuclides from daughter produc1- and neglection of neutron losses due to leakage and useless absor- ption e demanTh . n neutroo d n quantity neede o transmutt d e maith e n long—lived FPs may be compared with rate of neutrons generation in fast reacto e productiorth r usefo f dexcessivo n e fissile isotopes in breeding regime. Suppose that FBR with metal fuel has breeding ratio BR=1. d breedinan 4 g rati f coro e e BRC=1.th r fo O compensation of fuel burn-up reactivity effect. Then the value BR—1=O.characteristice th s i 4 e excessivth f o s e fuel production in the reactor blanket. It means that one fission in core leads to generatio 4 neutron0. f o nn blanke si t with taking into acco- unt of the neutron leakage and useless losses. Comparing this ex- cessive neutron quantity with minimal neutron consumptior fo n transmutation of the main long-lived FPs ( O.142 neutrons per fission after 3O—year deca e concludeb y ma perio t i d ) d that for transmutation of the own long—lived FPs in blanket of such FBR the useless losses of 65X neutrons may be allowed. This re- Refgectot. e used b r servinstance fo n e optimizatio,eca th n i , f blankeo n ' -•4 -fo(pvtGnent*:&t4s) Fu!• O-'JC-,. ?Mt/tmfjf %t«e CWr« Aieom<. 1t&->e (àpùiW l.ttl«$e. tope Pb—2O8e higcauses i th ht y I b threshold. f inelastio d c neut- //«. -OOiSg ns -32.23 SB.il -5 36 -021 4te fr-p*)- ron scattering of Pb-2O8 ( energy of the first excited level is PL -O.OJ9I/ 09i -MM 51.10 -3.o4 -Û.6É, 4. £i ni.tt.tJe 2.62 MeV ) and small neutron capture cross—section. Application of K -o.ookl 0.6 i -21 S9 23 o{ -/.Se -3.Slf 5. Jo Aâ *00619 -Ul -Sin 46.55 -B.n -0.08 tt.lS linear perturbation theor r assessmenfo y f voio t d coolant reacti- t Pi +O.OÏ37 -2.06 -SS.X iojo -S./6 -03£ 3.47 vity effect doesn't permi o analyst t e broath e d range f coolano s t K -0 00091 O.il -16 A M 6 r2.3S -3.33 SS3 density decreasing. The results of direct icalculations of Kef e* 2s -2 IS -60.97 Si.3? -set -o.og f.s- •$:* "i^ -t- O.OiZO variations cause y coolanb d t density decreasrn e presentear g n i d K-XK -os? -K.72 ft 37 -S3S -3.2$ 4.76 KXK -OSl n.os /t/63 -2 H -3.21t S./t Fig. 3. One can see that coolant density decreasing leads to sig- f* - © - -OÛOS1 c Ktifk ^6gZ -JûW 2M3 -sos -0.32 /3 iy nificant non-linear changes of Kef. When alloy 25 wtXNa -t- 75 wtXK PI-20S -0.0233 Mi -IS23 12. gZ -241 -O.IZ 394 is uses coolana de positiv th t e reactivity insertion causey b d coolant density decreasing doesn't exceed the Bef and may be compensated by the slight variation of FBR dimensions. So it mav Safet f o fasy t reactor-burne f o minor r actinides. be concluded that the most preferable coolants for FBR—minor Applyin f fueo g l containing minor actinide n FBRi s s leado t s tinides burne potassiume ar r , sodium-potassium allo d lea- an yen d degradatio f safeto n y characteristics, namel o decreasint y - ef f o g riched by Pb—2O8. As it is shown in Table 4 the effective frac- fective fractio f delayeo n o higt d d h neutronan positiv f Be - s so e tion of delayed neutrons for FBR-actinide burner is comparable dium-void effec f reactivito t yo upgradt t /(,/.i f e valuBe th e f o e is supposed the addition of the fissile nuclides characterized b high fraction of delayed neutrons < U-235, U-233, Tn-232 ) in fuel Keff composition. The decreasing of sodium—void reactivity may be f potassiuo achieve e coolana us s r potassiua my o b d( t m alloys Na-K, K-Pb ). The following coolants in core of fast actinide burner reactor were considered: alloy Na—K, allop y d K—Pan b enriche y isotopb d e Pb—208. Tabl 3 epresent e calcuth s l ational~r suits of void reactivity effect and its components for FBR with u )—nitridP , U ( e fued witan lh metal fuel containin e mino- th g ac r tinides. Void reactivity effec s evaluate e wa formulat th y b f do e linear perturbatio — ne seen) b theory y u P ma tha , R witt U I FB .t( h nitride fues negativha l e void reactivit y coolanty an effec r fo ts R witFB h . meta) investigateK l , fuePb l, containinNa ( d- mi e th g nor actinides is characterized by the significant positive void reactivit b coolantP d an y +18*Be( a seffecN d r -i-12.8*Befan fo tf , respectivel R witFB h. ) potassiuy m coolan s smalha t l negative void FIG 3 KeH as a function of coolant density in a. fast reactor reactivit yo decreaset effec e du t d densit d slowing-dowan y a c n (fuel 235U + MA + Th + 5 wt% Zr) TABLE 4 ß „, l , AND FISSION FRACTION OF MINOR ACTINIDES c Coolant

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208 traditionaf er witf Be h l fbr prompd san t neutron life—tims i p L e less by 3—4 times. In considered variants of actinide burner functio a densit O s cor2 e fasa a H th f B f e to nn o K yi reacto 4 G r FI e contributioth f minoo n r actinide totan i s l fissions quanti• abous i t 4O7- o abous , t 4O7 reactof .o r powe generates i r burnn i d - risk of moderator penetration in core is one of the main argu- up process of these nuclides. Such FBRs are capable to perform ment f opponento s o 2—circuit s t scheme f heaso t remova advann i l - the burn-up of minor actinides produced by 10 LWRs of equivalent ced concept FBRsf so FUR—transmutatan i . s sucFP hf o rcombinatio n electric power. ( fast corneutros a e n source surrounde blankey db t containin- mo g FPs transmutation in Fast Reactor. derator and transmuted FPs ) has the principal sense. In Fig.4 In this sectio wilt i consideree nb l problee th df transmo m u f changeKe shows ie i t th sn when water penetrate coren si n I . tatio f long-liveo n s witFP d h rather high neutron capture cra-ss- the case o-f ( U, Pu )-fuel reactor becomes super-critical. But if -section suc< ss 1-129 a h , Tc-99, Sm-151, Se-7 s d Pd-lOft 9an . 7) e minoth r actinide e insertear s n fuei dl compositio- n ap thee th n mentioned above there is the principal ability to transmute these pearance of hydrogen in core decreases the reactor reactivity. FPs in FBR blanket. If we want to increase the transmutation ra- This fact is determined by the threshold energy dependence of mi- s volusnFP - o reduc blanken t t e i ne dow ecu th d e nan o th e t , te r actinideno s fission cross—sections. Penetratio f moderatoo n n i r utron losses due to leakage then it is necessary to form the sof- core softens the neutron spectrum, multiplying properties of IT tened neutron spectru placiny b m f moderatoro g n I thi. s case nor actinides are degrading and Keff is reducing. So it can bt Fast Reactor intended for reducing of hazardous nuclides beco- concluded that in FBR with sufficiently high fraction of minor hazardouo t s me s nuclear installation itself: fast core surro actinide sprincipae therth s i e l abilit e o forsoft yth tm neutron y blankedeb d t containin moderatore th g . Accidental penetratJ spectru blanken i m y placinb t f moderatoo g r there n ordeI .o t r of moderato corn i y leao positivrt ema d e reactivity inser0 t1 limit the unfavourable influence of softened neutron spectrum in s knowIi t n thamaie th tn f reasono thi e s on o choosfact s wa t e blanket on burn-up of minor actinides in core it was suggested 00 e 3—circuith t schem f heao e t removaw operatinno n i l g FBRse Th . that core is separated from blanket by coolant layer ( see Fig.5 ) O3 - nuclide under transmutatio blanken i n; ) . 7 Tc-9- t& 1 ( 9 — between coolant layer and blanket it is placed the zone thicknes( containe) m c . 3 ) s- d. 7 0 Tc-910 ( 9 The decreasing of coolant density in layer ( 1O 7. ) results in degradatioe blanketh 2 n '/..I t 16 i t . O f o reactivit- n = f Ke y is realized the transmutation of O-72 nuclei of Tc—79 per one fission in the core. Taking into account the data of the Table 2 concludee b n ica t de actinidth tha n i t e burner thera pos s i e- sibilit r transmutatiofo y f Tc—9o n 9 generate reactorO 1 n i f do ( s equivalen n evaluatioa s i tt I powe n . f ) accounrleavio t e ou gth t neutron losses.

Transmutation of Sr-9O and Cs-137 in nuclear reactor. !h moJeïaïo*. Zl£2#ê«» E 4 ! nounons The most dangerous FPs are the isotopes Sr-9O and Cs-137 with half-life 27.7 year d 3O.san O years,- va respectively w lo o t e Du . lue of neutron capture cross—section of these nuclides it is ne- FI G5 Influenc f layeeo r coolant densit neutron yo n spectrum cessar o upgradt ye neutro th e n flux approximatel y 1OOb y O times comparing with achieved ones in modern nuclear installations. On- sucn li y h e transmutatiofluth x f o nSr—9 d Cs-13an O 7 wile b l Such technica firse th l tt solutio a tim s describe/ e wa 7 n / n i d effective. by the author s opinion. The main difficulty of this way is the necessity of transmuted notee b n dI ca t that coolant layer separating core with hard nuclides irradiatio n higi nh neutron flux with simultaneou- sre neutron spectrum and blanket with soft neutron spectrum carries quiremen f vero ty rational neutron utilization. Non-economical out two functions. At first, layer reflects fast neutrons leaking neutron utilization for RAW transmutation in facilities with high the cord preventan e penetratioe th s f resonanco n d thermaean l neutron flux may result in the negative power balance: power de- neutrons from blanket in core. At second, if the coolant density manded for RAW transmutation to harmless form may be higher than in layer decreases then reflection of fast neutrons leaking the power produced by the reactors where RAW is generated. RAW trans- core is reduced and the penetration of resonance and thermal neu- mutationeutrow lo n i t resulnn n flusignificann i ca tx t decre- trons from blanke corn i increaseds tei . These factors resuln i t W amounRA sinf to g compared with natural radioactive decay. degradatio f multiplyino n g propertie f actinido s n i o es fued an l r creatioFo f higo n h neutrgn flud conditionan x s when neutrons decreasing of Kef ( this effect may also be used in fast breeders e mostlar y utilizeW transmutatioRA n i d n with minimal lossen i s by placing the fuel containing minor actinides in the outer part construction materiale nuclideth necessars i e t us si o witt yh of cor. ) e specific properties. These nuclides mus w neutrotlo have eth n n examplea s A reactivite ,th y effec s calculatewa t actin i d n capture cross-section d rathesan r heavy atomic weigh r abilitfo t y burnee Fig.se : < ) 5r to fordesiree th m d spectru e possibl f neutronsth o mf o - e ma e On . ~ core surrounde coolany b d t laye Na—; ( thicknes, r m } Kc 5 1 s— terials is lead enriched by Pb—208. This isotope of lead is the - moderator in blanket - ZrH blanket; thicknesm c O 1O s- 1.85 double-magic nuclide ( neutron and proton shells in Pb—2O8 nucleus are completed ) and so its neutron capture cross—section is very small ( less than O.OOO5 barn even in thermal region ). Heavy ato- 3z- elements mic weight ( A=208 ) determines the small slowing—down capability \ of this nuclide. This fact in turn may be reason of slowing the \ N, \ \ ^ \ ^ chain reaction development in such nuclear system compared with \ r \ \ \ \ \ A A operating nuclear power reactors. It is evaluated that neutron \ \ N A \ \ \ \ \ \ \ 1 slowing-dow Pb-2On i n carries 8i t witou d h time duratio y 10—1Ob n O \ r \ \ times higher than in traditional moderators. \ \ •^ ^ ^ w slowing-dowlo e Th n capabilitw neutrolo d n an yabsorptio n s ^ probability of Pb-2O8 are the prerequisites for creation of fa- 20Br cility characterized by large volume to place the transmuted ra- dioactive y nuclideb hig hd an sneutro n flux. So the rapid and effective transmutation of Si—9O ( or Cs—137 ) 6 Principa G FI l schem f liquieo d metal nuclear reacto r transmutatioS r fo r n requires the high neutron flux and application of materials with low neutron capture cross-sections sucn I .h condition domie sth - nant fraction of captured neutrons will be related with Sr—9O One can see that main -fraction of absorbed neutrons is related r concentratiow Cs-13o Lo ( . 7) f fissilo n e material corn i ss i e with neutron capture in Sr-9O. Appreciable fraction o-f neutrons needed for the rational limitation of heat generation due to high s loss whicwa graphitn FP i t h d contenan e lean i t d allo s supi y - neutron flux. Rapid fissile materials burn-up in such core de- aboue poseb o ppm7 t d . During cycl f irradiatioo e e fractioth n n mands the really continuous refueling. For instance, such co- of neutrons absorbed in Sr—9O decreases due to redistribution of re may consist of the molten Pb-2OS mixed with fissile material neutron absorptions in produced daughter isotopes: Y—91; Zr—91, 92 ( preferentially with U—233 becausw o£>=C~c/C'lo f eo resonann i f - and Zr—93. Factors limiting the FPs exposure time are the reduc- d thermacan e l neutron spectru coolants a bot) m d fues a han l. tiof usefuo n l fractio f neutrono n s absorbee th n Sr—9 i dd an O placee b Sr—9 y graphitn i dma O e tubes arrange n regulai d r lattice graphite integrity after high fluence irradiation. Table 5 con- ( see Fi g.6 ). tain calculatee th s d assessment f neutroso n fluence (En>4e>-5 kev) exampls A core eth e with following parameter s consideredswa : and displacement r atope n sgraphitei m . Averag ecore valuth e f specifieo 0 kWt/1 13 ,- c power Publication n integrito s f differeno y t graphite s functioa s n Braphite tubes: outer diameter/thickness - 7.5 cm/l.O cm, of neutron fluence /8,9/ predict the permissible fluence up to pitch of graphite tube lattice — 15.5 cm, 22 2 Fissile materia U-233/U-234/U-235/U-23- l 60.6/0.21.2/O.O48/O= . .14 5*1 e considereOth r n/(cFo dm. ) reacto t meani r s tha- du t The dependence of K-effective, Fluence ( En > 46.5 KeV ), Dis- ration of irradiation cycle for tubular elements containing tran- placement graphitr spe e ato DPA-graphit( m d Neutroan ) e n Absorp- smuted aboue b 6 Sr-9yeary O. tma O. After exposur tubulae eth r tionuclidey b n s during irradiatio e core presenteth ar e n i n n i d elements are discharged, Sr-90 is separated from daughter nucli- consideres i t TablI . e5 d that reactor operate n stationarso y fuel d graphitan s de e elemen changes e fresi tth hy b d one. Thee th n composition with continuous fuel feeding by U-233. Braphite tubu- cycles of irradiation may be continued. lar elements containing Si—9 e place ar coreOn d irradiatei dan d beginine th e seenb t a n :f o gca irradiatio e Tablt Ii th n 5 e n 00 CO during cycle of predetermined duration ( O.5, l.O and 1.5 year ). K-trans(Sr-9O - 1.O> numbee OK-tran( th actf o neutrof r- so s n œ TABLE 5 NLUTRON ABSORPTION IN NUCLIDLS OF THE CORE* of Sr-9 othed s an witw 0neutroFP r lo h n capture cross-section. Howeve e technicath r l realizatio f suco nh reactor constructio- re n Time, o-f ut.iac/<.atu>^ , y fats quires the comprehensive analysis of technology to purify and feed O OS 1 0 15 the fuel based on Pb-2O8 alloy. Also it must be considered the im- FPs 4 Og 392 3 87 390 portant safety problems related with releas f delayeo e d neutron C-/2 S 26 752 7 85 79S- predecessors from cor circulatinn i e g fuel alloy which serves sa PL, 208 474 If61 8 L,5 461 coolant too n .thiI s concern thercertaia s i e n analogue with Si -30 75-07 5977 4949 U 91 successful experimen n reactoo t wherr/ O ItSR1 e / Esimila r release Y 9i O 648 591 S. 07 of delayed neutron predecessors was performed in circulating fluo— ^^-91 O 8.99 IS SS 2k 61 rides used as fuel and coolant. This liquid-fuel reactor was T.z-92. 0 Ott7 i S3 3 OS characterized by the high negative temperature and power reacti- li- 95 0 0.05 o 35- 0 96 vity effects. The small core specific power ( -~ 13O kWt/1 ) and Leakage. 70 678 671 675- good hydraulic parameters of graphite tube lattice allow to hope ft» OS5 0 S3 083 085 that important role in heat removal belongs to natural coolant K*ff 1.000 0978 0973 0979 circulation. K,„„A) 1.00 0829 0693 0580 Ffue/ici:, J3-, iz Cf->«&5Ä 0 MlO 89,10" 03 /u" On advantages of external neutron source for FPs transmutation. Dft!(Jl7y 36 6 7 S l, 0 U3 s showA n abovquantite th e f excessivo y n eca neutron R FB n i s * Total absorption (excluding fuel) in the core plus e sufficienb o providt te positiv th e e balanc n i transmutatioe n neutro ) reactionn2n leakag n ( sd equaean o 100t l % d generatioan f long-liveo n s witFP d h high biological hazard. Poor neutron balance of thermal reactor doesn t permit to achieve - Ac f fissioe coro . th ) et n i ac n capture e on n Sr-9 i r s pe O analogous situation. However, estimating the princilal ability countin yiele f th Sr—9g o- d fissior Ope abou( n t O.O6 r O.O,o - 3af of fission nuclear reactorW transmutatioRA n i s n into harmless ter 3O—year decay time ) it can be concluded that one fission in forms and considering the use of external neutron sources for such reactor leads to transmutation of Sr—9O produced in 14 ( or thie noteb e sfollowin th y dpurpos ma t i ge points. 28 ) fissions in other reactors. The value of ( 1 - K-trans ) At first e valuth f ,o excessiv e e neutrons suitablr fo e characterizes here the fraction of transmuted Sr—9O. Ff the transmutation purposes is not very large and so FBRs may be used specific power in that reactor is /»» 13O kWt/1 then neutron flux r transmutatio n fo long—liveow s s pluit FP f sdo n perhaps from 16 2 is 2.65*1O n/(cm * sec). Such neutron flux provides the one—two thermal reactors t meanI e .nuclea th s r power system rat f o Sr-9e O transmutatio y 12.nb 5 times higher than- ratna f eo with small fraction of FBRs will suffer from deficit of tural radioactive decay (half-life - 27.7 years). Neutron spectrum neutrons neede r transmutatiofo d f hazardouo n s long-lived FPs. in core is close to Fermi—spectrum since the multiplying medium is It will be necessary to find new sources of neutrons. extremely diluted system. The core is characterized by following At second, amon e long-liveth g s therFP de nuclide ear s with average cross-sections: C" c Isotope separation of Sr-88 At sixth, effective transmutatio f relativelo n y short-lived dangeroue ith s s radiation procedure since radioactive strontium radionuclides w witneutrolo h n capture cross-section Sr-9O( s , has to be converted to gaseous phase. Avoiding o-f isotope separa- Cs—137 ) requires high neutron flux. It is impossible to exclude tion by using o-f the suppleaentary large neutron source is im- the possibility that needed neutron flux can be successfully portant argument -for development o-f such -facilities -for FPs tran- created in non-fission nuclear installations with less expensive smutation without long decay period. and less complicated technology. It may be facilities based on At third, amone long-liveth g s ther FP nuclids i de e Sm-151 nuclear reactions with smaller power output compared with reac- < half—life '»•• 87 years ) with complicated scheme o-f transmutation. tiof fissiono n ; neutron sources base n fusioo d n reaction o r o n This nuclide has rather high neutron capture cross-section for interactio f o intensn e beam f o particles s with heavy targets, effective transmutatio e neutroth n i n fluf nucleao x r reactor. for instance. But Sm—151 is component in the mixture of FPs—samarium isotopes At seventh, one of the important advantages of high neutron ( from Sin-147 to Sm-152 ). Neutrons will be consumed not only in flurelates i x d with possibilita dilute e us d o nuclidt y e sys- Sm—15 t thebu 1y wil uselessle b l y absorbe othen i d r samarium iso- tems. transmutatioe Theth n n process will deal with small quanti- topes. To avoid this problem it is necessary to apply the isoto- ties of hazardous radionuclides. It corresponds with the rational pe separation as mentioned above. Otherwise the mixture of all desire to reduce the amount of radiotoxical materials both in samarium isotopes must be irradiated. Since the most of samarium inned outean r r part f o transmutatios n cycle. isotopes have the long half-lives ( or are stable ) the transmuta- tion of mixture will finish by Sm—153 ( beta—decay with half—life Eu-15o t 46. h 7. Until3) l exhaustio Sm-14f o n 7Sm-15- O isotopes, the precursors of Sm—151, its inventory willn't sufficiently de- References crease. For transmutation of one Sm-147 nucleus to nuclide of another element Eu—15needes i successive t i th d3 e absorptiof o n 1. Bauer S.S., McDonal . A dNuclea r technologie sustainabla n i s e 6 neutrons. Even separated Sm-151 with frequent purification from energy systems Yorkw .Ne , 1983. daughter isotopes require least a s neutront2 transmutatior fo s n . Sol2 . aA Som e preliminary result actiniden o s s incineration of one Sm—151 nucleus. So transmutation of Sm—151 in mixture of and transmutation in a thermal and a fast rectors. - other samarium isotopes is very neutron—expensive process which N.E.A.C.F.R. - A-13OO. Section 4.9, p.281-293, Joint Research may be realized with use of the external source of neutrons. Centre - Ispra, Italy, 1987. At fourth more th ,e completed decontaminatio requireW RA f o n s . Mukiyam3 Takan, aT. , Takizuk oH. . aT Higher actinides transmu- the extensioc of isotopes list to be transmuted. The external so- tation using higher actinide burner reactors .— Proceeding s urce of neutrons will be useful for transmutation. of the International conference on the physics of reactors: At fifth, duration of exposure followed by the procedures of operation, design and computation. April 23—27, 1V9O, Mar- purification and feeding by portion of fresh transmuted nuclides ceille, France, vol.1, pp. 97-107. is determined by the amount of excessive neutrons. Desire to in 4. Sorbachev V.M., Zamiatnin Yu.S., Lbov A.A. Interaction of ra- creas e duratioth e f irradiatioo n no reduc t cycl e quand th ean e - diation with heavy nuclei and fission of nuclei. Handbook. 00 01 titf o outey r cycle operation transmutee on r pe s d nucleus will Moscow, Atomizdat, 1976. oo 5. Bowman C.D., Arthur E.D., Lisowski P.W. et al. Nuclear energy MINOR ACTINIDE TRANSMUTATIO FISSION I N O) generation and transmutation using an accelerate»—driven in- REACTOR FUED SAN L CYCLE CONSIDERATIONS tense neutron source. LA-UR-91-91-26O1, 1991. T MUKAIYAMA, H YOSHIDA, T OGAWA 6. Shmelev A.N., Kulikov E.B. , Glebov V.B. et al. On sa-fety cha- Japan Atomic Energy Research Institute, racteristics o-f -fast actinide burner reactors. - Proceedings Tokai-mura, Naka-gun, Ibaraki-ken, o-f the workshop on nuclear transmutation of long—lived nuc- Japan lear power radiowaste. July 1-5, 1991, Obninsk, USSR, pp. 192-198. Abstract 7. Leipunsky A.I. Selected works. Memoirs. Kiev, "Naukowa Dumka", The concept of minor actinide burner reactor is proposed as an efficient and effective way to 199O, p.64. transmute long-lived minor actmides in order to ease the burden of high level radioactive waste disposal problem Conceptual design studies of minor actinide burner reactors and the fuel cycle . McCo8 y H.E .Beat, y R.U w .Coo, development Ne k . W.Hal t .e n i s facilities were performed to obtain a reactor mode! with very hard neutron spectrum and very high materials -for molten-salt reactors. - Nuclear Applications neutron flu whicn i x h minor actmide e fissioneb n sca d efficientl modelo Tw y f burneso r reactors and Technology, vol.8, Feb. 1970, pp.156-169. were obtained, one with metal fuel core and the other with particle fuel core Minor actinide transmutatio actinide th y nb e burner reactor compares si d with tha powey tb r reactors from bote hth 9. Scott D., Eatherly W.P. Graphite and xenon behavior and their reactor physic fued san l cycle facilities view point in-fluenc n o molten-sale t reactor design - Nuclea. r Applica- tions and Technology, vol.8, Feb. 197O, pp.179-188. 10.Alabyshev A.F., Srachev K.F., Zaretsk y. SodiuS.Aal d t an e .m . Introductio1 n potassium. Moscow, Eoszhiaizdat, 1959. Various methods of minor actinide transmutation have been proposed in order to alleviate the 11.Novikov S.M., Ignatiev V.V. et al. Nuclear power plants with burden of the high-level radioactive waste(HLW) depository problem One of the practical methods is liquid salts: perspectives and problems. Moscow, Energoatom— to recycle the fasmn i t reactors since minor acumdes suc Np-237s ha , Am-241, Am-243 (hereafter fissionable ar ) referreMA s eda with fission threshol rangV capturd ke aboun d i ean 0 70 te cross izdat, 1990. section thesf so e nuchdes rapidly decrease with neutron energy higher than this threshold In the early 1980's, the feasibility of MA transmutation in power reactors such as LWR or FBR was shown In these reactors, however, neutron spectra are rather too soft for MA to directly undergo fissio therman I n l reactors, these nuclides undergo fissio neutroo tw nr o nafte e captureon r , for example, Np 237 undergoes fission as Pu-239 Even in FBR, the fraction of neutron of which energ highes yi r thafissioA nM n threshol rathes di r smal effectivr fo l e fissioA M f no Ispeciaa f l reactor wit hhara d neutron energy spectru higd man h neutron flu technicalls xi y feasible, it will be very efficient and effective for MA transmutation In this context we have been studyin concep e burneA gth M f rto fast reactors(ABR,Actmide Burner Reactor concepe e Th )th f o t double strata fuel cycle consisting of the commercial fuel cycle and the Partitioning-Transmutation cycle is illustrated in Fig 1 The final HLW from this double strata cycle contains only fission products

2. Designing of minor actinide burner reactors

desige Inth n study fuee th , l propert thermae th d yan l hydraulic analyse nucleae wels th sa s a l r analysis wer emode R earnee obtaiguidelineo t s AB t Th a l n dou ne a ar r designin fo R s AB n ga follows, majoe th s ra fueA M l material, very hard core averaged neutron spectrum - very high neutron flux, 1st Stratu Fuef mo l Cycle - bumup reactivity swing less than 3% A k/k per cycle, - long fuel residence cycle length within the maximum allowable neutron irradiation of cladding material.

In Table 1, the composition of MA used in this study is shown. Two types of ABR design Fuel Fabrication Plant 1000MWE Reprocessing Plant were obtained particl, A namelM metad A ey an M fue l R fuel ABRAB l followinge .Th briee th fe sar explanation of ABRs designed. The details of these ABR designing are described elsewhere'Ii2'3'4^ For design study, a computer code system "ABC-SC" was developed to analyze actinide HLW transmutation in fast reactors.5' , j 1 LWR 3410MWT.X13units FPs 14.3t Actinide 1.5t 99.5%U,Pu f0.5%U,Pu 1.2tA 2.1cooledNa alloyMA fuel ABR(M-ABR) i 0.3t/ (^^^^J To design a core with a very hard neutron spectrum, a metal fuel core is the first choice. The ™ ^ ^ pyrochemical reprocessing of metal fuel is also attractive from an economic view points of fuel cycle facilities becaus compactnesse th f eo . Experimental MA fuel property data which are essential for designing fuel and a reactor core are very scarce. Theoretically estimated data, therefore, were used in design study. The followings are the result of the estimation4^; 1) Similar to U-rare earths, Np and Am are not mutually soluble, improvo t ) 2 meltinw elo gmetal A poinM f ,element o te e.g A b M .o t , 640t e Np s ar f ;o alloyed with thermal diluent such as Zr and Y, existenc) 3 woulu P t significantlf eo d no y affec solidue th t alloysf so .

Table 1 Minor actinides generated in a 3410MWI-PWR per year

Cooling tine (year) Nuclide 3 6 OTNp 56.2% 47.6% MAm 26.4 38.3 MAm 12.0 10.1 "'Cm 0.03 0.02 244 Cm 5.11 3.83 243 Cm 0.28 0.23 Final Disposal less than 1000Years Total 100. 100. Total Weight 23.8kg 28.9kg Calculation data : JENDL-2 SRAC-FPG: code S Fig. 1 Flow of high-level radioactive waste per year through double Burnup of fuel : 33.000MWD/MT 00 strata fuel cycle combined with partioning and transmutation (MA burner reactor) cycle : 100% u P Recover d an U f yo CD As the result of these nature of MA metal property, we have to use two alloy systems, namely, GO Np (Pu) Zr and Am Cm-(Pu) Y In these alloys, Pu is added because of two reasons, 1) To reduce cntical mass, 2) To compensate for reactivity gam which is caused as a result of conversion of Np- 237 to Pu 238, Am 241 to Am 242m introduces large reactivity gain and this should be compensated with burnup reactivit , howeverPu yu losP addes f i ,s o dinitiae onlth t ya l loadin afted firse gan th rt 0 loadin converteu gP play7 d 23 fro sp mthiN s role c o cooledHe particle22 MA ABR(Pbed ABR)

Tl thermaw Lo l conductivit meltind yan gmetaA poin limitine M lf th o tfue e ar lg factore th r fo s high MA bumup in M ABR Therefore, the particle bed reactor concept was applied as an alternative ABR, which has the high efficiency in heat transfer since small particle size produces a large heat transfe coatef ro surfacd d be r volum fuee epe lTh particle® e containe doubln di e concentric porous o> 3 fnt directls i s y coole heliuy db microsphere fuea s mi Th l nitridA M f eeo whic coates hi d witha a; refractory matenalN sucTi s ha vT In a cold fuel concept, fuel temperature is to be kept lower than one third of its melting point o reduco t e mass transport Since reduced mass transport result smallen si r swellinreleases ga d g,an thicknes coatinf so gminimizee layeb n rca givo dt e large heavy-metal densit orden yi obtaio rt n hard neutron spectru core kernea fuee mn th Th i el f particlo l homogeneous ei A M d s mixturan u P f eo > Cd nitnde The fuel concepts of M-ABR and P ABR are shown in Fig 2

Designing23 plants fuelofABR and cycle facilities

Conceptual design studies of ABR plants and their fuel cycle facilities were also earned out to assesheae plan th R feasibilite t, 3 balancconceptsth R AB t consisting M AB Fi f f eo n yo I s6 f go modules of M ABR is shown In this plant, 300kg of MA generated in about 13 units of 3410MWt undergR PW o fission yearl 353Md yan W electricit generates yi d heae th plan, tR 4 balanc shows AB i tg In P Fi thi n f I neo s scheme , fro355kA 5 mM 1 f go units of PWR is fissioned yearly Because of the cold fuel concept adopted for designing a P-ABR, the fuel temperature is restncted to lower than 1000K and the coolant output temperature is only 340 coolee H °s Ca dwhic w reactolo s hi r Under temperaturthiw slo e condition applicabln a , y ewa generatioe oth f electricitf no workinf o e us ygy wilb flui e lb d witmeltinw hlo g point, suc freos ha n To avoi matenae dth l corrosion proble thermamy b l decompositio freonf no temperature th , freof eo n shoul lowee db r than 140°C Ownin theso g t temperature w elo electricite th , y generation efficiency is only 15% to In the designing of a fuel cycle, pyrochemical reprocessing of spent MA nitnde fuel was studied becaus compactness it f eo e estimateTh s fuee dth lmasn i ) s Pu flo f actimded wo an A s(M cycle facilitie s onli s y e uni 19kf f actimdeson o o t f fissio gr o fo g y n2k da 3 products, r pe c et , 1200MW1P ABR plant when the plant load factor is 30% (Fig 5)

3 Characteristics of ABRs

Reactor1 3 performance

The reactor core design parameters of M ABR and P ABR at their equilibrium state are shown in Table 2 respectively Comparison of core averaged neutron spectra is shown in Fig 6 In this Fig.3 Heat balance of M-AJBR plant A plant which cosist unit6 f M-ABf so so R bums 300kf go MA(minor actinides) per year and generates 353MW electricity

0 D

Fig 4 Heat balance of P-ABR plant

89 CD Table 2 Reactor design parameters of Actmide Burner Reactors O Power Reactor Fuel Cycle PWR(3410MWt) 13umts M ABR" P ABR2' InventorM H y 85tonxl5 Fuel concept pin bundle coated particle Batch Reloading 1/3 of Inventory/year 3 material 1C > Np 22Pu 20Zr (66NpAmCm-34Pu),0N,0 Burnup 33GWD/HMT OC AmCm 35Pu 5Y Minor Acorude 357kg/year MA initial loading,g 4k ' 666 2065 Np/Am,Cm/Pu 255/199/212 765/598/702 Reactor power, MWth 170 1200 Partitioning Coolant material Sodium Helium Np 201 kg/year velocity, m/s 8 total flow kg/s 1088 Am 137 kg/year Fuel Fabrication- inlet pressure, MPa 10 Cm 19 kg/year 3 pressur1 e drop a kP , MA 357kg/year inlet temperature.'C 300 127 outlet temperature(core max),^ 1C 484 OC 440 340 Fuel temperatur maxeÇ 3' 9 80 C O 4 1C83 722 P-ABR Clad temperamre.'C max6' OC487 1 C5 4 0 t temperature56 Fn x ma , Output 1200MWt 15 2 InventorU TR Ito2 ny Reprocessing Neutron flux, 10 n/cmc se 1C 4 1 OC 3 4 84 2 Batch Reloading Whole core/year Neutron fluence (E>0 IMeV7 1 C )O ICFn/cm 2 2 C 1 22 MA Fuel Cycle Facility Core averaged mean neutron energy,5 ke78 VC O 6 1C76 743 TRU 2 1 ton/year Reacuvity(% ak/k) u P d an A M U TR Plant Load Factor 30% TRU 19kg/day Na void reacQvity/core 252 —— Doppler reacQvity/cor t=300Ta e( C) 001 001

Kineuc parameters Waste 3 fa 1 55X10 1 72X103 P 357kg/yeaF r 8 8 /,p, sec 684XJ0 108X10 ^Others Cycle length, full power days7' 730 300 MA transmutation, %/cycle 260 25 3 Fig 5 Mass Flow in P T Fuel Cycle with P-ABR MA bumup, %/cycle 178 17 3 1) M-AB metalliA RM c fuel actimde burner reactor 2) P-AB particlA RM e fuel actimde burner reactor InneC 31 )r Core OuteC O , r Core figure, neutron spectru f MOX-LMFBmo Ralss i o show compansor nfo n Significantly hard neutron 4) After 1st cycle.only Np.Am.Cm are added spectra of ABRs are obvious In the M ABR, the neutron flux is not so high as opposed to the initial 5) Predicted melting poin f fueR to lAB 900TM r :fo attempt to design a very high neutron flux reactor because of low melting point and low thermal Max allowable temp of fuel 727t; (1/3 of M P 3000K) for P-ABR conductivitABRP e neutroe th fue,th A n I lM f n yo burnu fluA vers M xi r yea e yppe th s ri hig d han allowablx 6Ma ) e tem cladding/fnf o p 650X) 9 T ;(H t large owninrR efficienthae AB th no g t thaM f t to heat removal characteristic particlf so e fuel 7) Fuel irradiation time significane th f o e tOn differences between ABR powed san r reactor tha s sfuei e tth l residence cyclf eo ABR limites si neutroy db n fluence while thos f poweeo r reactor limitee sar bumuy db p reactivity e facth t los o that sABRn e i t Thidu burnue s th si s p reactivity swin e resuls smalgi th f s o ta l compensation for bumup reactivity loss by reactivity gam from the conversion of MA to fissionable Am242/Ao t 1 24 m mA , 242mPu o t materia )p N g (e l 10° Tabl e3 Compariso f JENDL-no 2 based nuclear calculatiod nan ENDF/B-V M-ABr basefo e dRon

10' Library Item JENDL-2 ENDF/B-V

10' One group cross section(bam)*) ^^Np capture 0711 0678 Fast neutron fraction (.%) fission 0 575 0 577 10" 241Am capture 0 96 0 4 10 >100kev >lMev fission 0587 0531

Core averaged neutron mean reaction 790 757 M-ABR 80 23 energy in the inner core(keV) P-ABR 76 22 cyclt 1s t ea C BO f o f k^ 0551 0318 3 10' IMFBR 56 10 MA bunup per cycle in the inner core at 1291 1308 10th cycle(%) 10' *) collapsed usin core gth e averaged neutron spectrum

10' l 111 t 1 l I 1 1t II l 6 10 4 10 3 10 2 10 ' 10 10° 10' . Compariso4 transmutatioA M f no n powei ABRn i nd ran s reactors Neutron Energy (eV) In Table 4, the transmutation characteristics are compared between two types of ABRs Fi g6 Compariso f corno e averaged neutron spectra together with therma transmutatioA d fasM an tl e reactorth r Fo n powei ns r reactorse th , of minor actmide burner reactors and MOX-FBR concentration of MA is assumed as 0 2% and 5% of heavy metal for U-PWR and fast reactors, respectivel thao additio e ys wilt affecA th t no M l f tn o major reactor design parameters sucs ha enrichment, coolant void coefficienc et t currene Inth t desig ABRsf no , Doppler reactivity coefficien delayee th d an dt neutron fraction vere ar positivye smalth d an le sodium void coefficien largs i tcora d n ei ean owin lavJe U th f co o gt 4 I MA burnup rate vere alsth y o hart d neutron spectrum transmutatioe Th n rat usualls e i cyclf o d thao et y en weigh tdefineA e ratie M th th t f oa ts o da Effect2 3 of Nuclear Fueland Property Data Uncertainty R DesignAB the on beginnine oth f cyclf go thin I e s definition nucleay an , r reaction suc s fissionha , neutron capture, (n,2n), etc can be used for transmutation and the conversion of Np 237 into Pu is the transmutation Nuclea fued ran l propert stronglyA datM f ao y affec desige th t ABRf no s because their major Pu in the transmutation chain of Np-237 is mostly Pu-238 and these duly Pu is not favorable for the fuel material is MA Also, the composition of MA will be full of variety depending on the spent fuel reactor physics and the ex-core fuel handling Therefore, the MA burnup rate defined as the weight of different irradiation history The effect of nuclear data uncertainty was checked by two sets of ratio of MA fissioned during the irradiation to that at the beginning of cycle is the real index of calculatio M-ABRr nfo usine ,on g JENDL- othee 2th datrd usinaan g ENDF/ summare Th BV f yo transmutation effectiveness and efficiency because only fission is a real transmutation reaction to this compariso shows ni differencee Tabln i Th e3 eacf o s h parameter quite sar e larg t thaf ebu o t solv problee eth long-livef mo A dM MA bumup is rather small. In Table 4, the burnup rate per cycle (transmutation effectiveness index) is 15 to 18% in all fuee th l r Apropertsfo y theoreticar dataou f i , p N l n predictioi m n solubhtC , no f no Am f yo reactors excep MOX-FBtm bumue RTh p ratyear(transmutatior epe P-ABn ni 4 speed% d R17 an s )i t correctno s fuei e th , l desig M-ABf no R wil vere b l y differen thermae Th t l conductivit metaA M lf yo othen i % r 7 reactor o t s High bumu hars pit ratdo t P-ABn neutroei e du Rs i n spectru verd man y high assumes fue halwa e l f thath f o f U-Ps d o a tallor Z u thif I y s numbebumuA thirde M on e s ps i ri th , neutron flux In the MA bumup calculation for power reactors, MA generation from their fuel should smalle% e 26 lOOtTh r ; lower meltin gfueA poinM l f tha o estimatee t nth d 90CTC result% 18 m s be accounte generatioA M r dfo larges ni r thabumuA nM U-PWn pi aboud Rhalan e f bumuth tfo n pi Jowe bumuA rM p fast reactors The net MA burnup per IGWt a year of ABRs is significantly larger than that of power reactors CO Table 4 Comparison of MA transmutation in various reactors Table 5 Effect of MA addition in PWR fuel ro ) HM f o n (valueto r pe s MA Burner Reactors Power Reactors R M-ABAB P R U PWR MOX-FBR LMR Item Reference PWR MA-PWR (MAPWR)/(RefPWR) Output (MWt170 ) 12003410 2600 2632 Nuclide (g) Cycle length» (FPD730 ) 300 850 750 900 Np 469 918 20 Core averaged Am 162 276 17 Neutron flux (xlO's) 30 65 3 3 84 7 3 0 Cm 38 296 77 Mean neutron energy (keV78)0 750 thermal 480 490 3 4X17 0 1 7X104 510 MA loaded (kg 6 66 ) 2065 180« 1450« 1200« Bk MA generated" (kg/cycle) 0 9 5 83 4 71 Cf 3 5xlO'7 24X104 690 MA transmutation ratio4' (%/cycIe3 5 2 ) 7 2 68 0 3 6 7 2 1 4 5 o -activity MA bumup ratio" (%/cycle) 178 8 16 172 4 9 150 029 1 3 46 (105Ci) transmutatioNeA tM n (kg/lGW year) 407 435 25 62 47 bumuNeA tM p (kg/lGW year) 279 296 -43 8 1 14 6 63 36 58 1) Fuel irradiation tune (10 n/s) Spnt fission U-PWRr ) MOX-FBr fo Concentratio2 fo R % 2 % fue n LM 5 ,i 0 , d l A Ran M f no 049 35 7 1 generatioA M u ) P 3 fuel.id n i an eU (109n/s) transmutanoA 4M ) n rano=( MA(BOC) MA(EOC MA(BOC/ ) ) bumuA 5M ) p rauo= fissioneA (M / MA(BOC d) ) MA addition 02%ofHM U enrichment 32% Burnup 33000MWD/T Irradiation 847 days 4 2 Influence of MA transmutation in power reactors on fuel cycle Cooling day0 15 s

For the MA transmutation using power reactors, not only reactor performance and fuel manufacturing but also the influence of transmutation on the fuel cycle facilities should be taken into accoun large Th te difference betwee transmutatioe nth burnue nth ratd pean rat powef eo r reactors show Tabln ni indicatee4 largee sth r conversio heavieo t A M r f nuchdeno s thao T n fissioA M f no case I nth ABRsf e o shieldine decath ,e th yd gheaan t remova muce lar h severer problem than evaluate the effect of MA addition to the fuel the analysis was made to calculate the increase of decay transmutanoA theM powenm r reactors sinc concentratioe eth vers i yA higM s i f ABR n ht o i i s A s heat, neutron emission, and y -ray intensity In Table 5, the increase of neutron emission m spent mentioned in the section 2 3 of this report, the fuel cycle facilities for ABRs are very compact and the fuel causeincrease th y d b neutrof eo whef addeC s ni d 2wtemittern0 A o an dt %M k B , s sucCm s ha number of these facilities is also limited Therefore, the economy of MA transmutation may be PWR fuel is shown In Table 6, the effect of MA addition to power reactors of U-PWR, MOX-PWR favorable for the ABRs even if the resources required to develop ABRs is larger than that for MA summarizes i R andFB increase Th d decaf o e y hea causes generatiote i th y db Cm-24f no e th d 4an transmutation in power reactors ^ increase of neutron emission is caused by the generation of higher Cm isotopes and Cf-252 case I nth f fas eo built8 reactors wil20 p alsde u l proble e b lT o th e th , m becaus higs it f heo energy t-ray (2 6MeV) 7) Tl 208 builds up in the Th decay chain starting from Np-236 which is . 5 Conclusions formed by (n,2n) reaction of Np-237 When 5wt% MA (composition shown m Table 1) is added to the MOX-FBR fuel, the Tl 208 build up is about 50 urnes as much as that of without MA addition e designeTwar oR thesn typeI dAB f eso burner reactorsburnuA M e p th , rat r cyclepe s ei At this level of the concentration, Tl 208 build up will pose a shielding problem significantly higher than those in power reactors The small Doppler reactivity coefficient, delayed A resuls a transmutatio A increasf strond M o t e an f th g f C -emitterf thesf eo , o l nal eCm r fo , neutron fraction and the large positive sodium void coefficient of the present ABR designs is less scheme in power reactors shown in Table 4, the shielding design change will be needed not only for favorable from the reactor safety point of view MA transmutation in power reactors (LWRs and the fresh fuel handlmg(manufactunng and transportation) but also for the spent fuel handling FBRs) will require the design change of the radiation shielding in the whole fuel cycle facilities (transportation and reprocessing) This may cause the increase of cost in the generation of electricity because of the increase of strong neutron and y emitting nuchdes Cost evaluation is required to Tabl e 6 addtio A EffecM f poweo o nt r reactor fuel handling 5) YGunji.et al A computer code system for actimde transmutation calculation in fast reactors -ABC-SC-",JAERI-M 92-032(1992) Rati adde A valuf oo M df e o fue thao t lnormaf to l fuel Hom.eL F ) 6 l -"Compacta t Nuclear Power Systems base Particln do Reactors"d eBe , BNL- Reactor/Fuel Decay heat Neutron emission Photon intensity 38379(1986) LOBRJ ) 7 Y "Transmutatio Np-23f no fasn 7i t breeder reactor Tl-20d san 8 build-up", Pro Cont cIn f 2wt%)0 ( R UaPW on Fast Reactors and related Fuel Cycles FR91 ,Vol 11,19-3 (Kyoto,1991) fresh fuel(U235 4wt%)b 36X103 8 3X104 1 3X103 8) T.MUKAIYAMA.e l "Minota r actmide transmutation using minor acumde burner reactors",ibid Vbl 11,19-6 spent hiel(45GWD/t)c 1 5 4 1 MOX-PWR (0 5wt%) fresh fuel(Pu 6 5wt%) 14 48 12 spent fuel(45GWDA) 1 5 17 1 MOX FBR (5wt%) fresh fuel(Pu 30wt%) 22 1 OX102 21 spent fueI(80GWDA) 28 19 1 a. minor actinides(MA) fraction in fuel (HM weight %) b fuel enrichment c fuel bumup (cooling time 10 years)

choose the cost effective transmutation system based on the reliable data base which is not yet available concepR AB e t wilTh l enabl confinemene eth f troublesom o closete on e Fron di su mA eM the economic safetd san y view point desirable b confinemen e y th , ma e A M f o t

ACKNOWLEDGEMENTS

authore Th s gratefully acknowledg significane eth t contribution YGunjf so Nucleaf o i r Fuel numericae th n i c lIn analysis

REFERENCES

1) TMUKAIYAMA.et al "Conceptual study of acarude burner reactors , Proc Intn'l Reactor Physics Conf (Jackson Hole, 1988),Vol IV,p369 2) HTAKANO,etal "Nuclear Characteristic Analyse BurneU TR f rs o Reactor s JAERI-M 89-072 (1989) 3) T TAKIZUKA,et al ' Thermal-Hydraulics of Actimde Burner Reactors ', J AERI- M 89 091 (1989) CO 4) T OGAWA ,et al "Fuel Elements and Fuel Cycle Concepts of Actimde Burner Reactors", JAERI- CO M 89-123(1989) g HOMOGENEOUS RECYCLING OF MINOR ACTINIDES tion is paid to the evolution of mmor-actinides contents during recycling, and to IN AN EFR TYPE FAST REACTOR Pu238 as a transmutation product of Np237 and Am241

H.W WIESE, B. KRIEG 2 EFR AND LWR SPECIFICATIONS Institut fur Neutronenphysik und Reaktortechnik, Kernforschungszentrum Karlsruhe, EFRe th r cor,a Fo e mas3tHM0 4 f averagn so a , e core burn-u f 130GWd/tHpo Mn i Karlsruhe, Germany 3 cycles, a cycle length of 2 years (640 fpd, 90 d for discharge and reloading), a cooling timyeayears3 1 r reprocessinf d eo fo r an , g plus refabncatio- as e nar sumed Abstract Fuel management model The core is subdivided in 3 equal-mass load regions The capability of the European Fast Reactor (EFR) as a typical large fast reactor t BOC1A 3 l threR al , d ean R1loa2 R ,d region chargee sar d with fresh t fueA l to incinerat minoe eth , producer actmideAm d LWRan dm bees p sN ha s n inves- EOC1,2,3 regione th , s R1,2, dischargede 3ar recharged an , d with fresh fued an l tigated for the case of homogeneous recycling of Np, Am and Pu Np, Am from external LWR waste, respectively, assuming that Np and Am can be Detailed accoun takes i threa t r nfo e region reload schem whicn ei h evero tw y separated from the waste The irradiated fuels from R1 and R2 are only partly core th refuelles ef i year o fractio3 e s1/ Th d f admixeno d minor actmider va s si withdrawe ar burn d an t n fro procese mth f recyclinso t EOC4,5,6A g , again R1,2,3 led from 2 8 to 7 5% dischargede ar , wit wit2 hR h reloadin d fres an from 1 h A R mfue wit d , f go an hlNp Results are given in terms of the number of clean-up LWRs by one burner, the external LWRs For R3, the reload now consists of fresh fuel, of recycled Pu, Np nuclide inventories during recycling, and the risk potential of the waste from the Am from the reprocessed fuel discharged at EOC3 and of Np, Am from external incineration system compare non-incineratioe th o dt n case LWRs This process is continued, always recycling in cycle n the Pu, Np and Am concludes i It d tha enforcen a t d research shoul done db chemistrn ei provo t y e f cycl o 2 dischargeReprocessind en- en e th t da assumee gar losse% r 1 dfo f so tha satisfactora t y separatio nfro m rarespeciallalse d C m th f e an oo earthm A f yo s Pu and Np Since Am is closely bound to the rare earths and hence can be re- in the waste is possible on large scale covere onlo dt y 5-40% calculatione th n i , e s ar estimate % 70 f o d lossem A f o s used As a parameter, the amount of admixed Np and Am is set to 2 8, 5 0 and 7 5% of the total reload mass per cycle Reactivity control is performed by an appropriate chang reloaPu/( e e ) rati3 th th f f Pu Ud eo o d + an fue t eac2 a l C h BO cycl t A e 1 INTRODUCTION Pu/(U+Pu) is increased according to the reactivity loss due to the Np and Am charge from LWRs only KfK and the French CEA are investigating in close cooperation the transmutation The burn-up reactivity los thin si s phas assumes ei compensatee b o dt withy db - long-livee th d of an hazardou dm fissioC , nAm s , productnucleNp s a i s Tc9d 9an drawing absorber rod keps i s C t FroconstanBO t ma cycl» thao K t tf cycl, o et4 e 1129 in fast and thermal reactors 3 Due to the deterioration of the recycled Pu, for 28% admixture Pu/(U + Pu) in Sinc fasn ei t reactors fission products canno destroyee b t d efficientl y- excep o t t creases from 20 5% at BOC 1 to 27 5% at BOC 19 Since the equilibrium core in- a certain degree in moderated subassemblies in outer core or blanket regions - ventory depend e additioth n o sf fres no h plutonium, refine d reactod2 r calcu- this paper is restricted to the incineration of actmides lations to determine the needed Pu from k« for the full reactor will be used for The results presented here are part of a common CEA/KFK paper [1] prepared for n further investigations the ANP Conference m Tokyo in October 1992 However, here also the risk po- tentials in terms of the ingestion hazard from the minor actmide waste of the in- The Np and Am to be burnt in the EFR originates from a reference 1300MW«,, PWR cineratio withouR LW n e tsyste th actmid d man e burnin compares gi d Sincr efo of 40GWd/tHM averag r yeare pe burn-up coolina ,M IH 4 4 greloaa , 4 tim2 f f deo o storage timeabouo t p su t 1000 year fissioe sth n products dominat ris- e eth kpo 3 years before reprocessing, and 1 year for fuel element fabrication tential, this comparison mainly is performed for medium- and long-term storage The flow of the actmides during recycling can be seen from Fig 1 times larger 1000 years In orde accouno t r realistia r fo t c 5-40e recover i , %Am f durinyo g reprocessing using the PUREX process, as a parameter a loss of Am of 70% is assumed in this 3 CALCULATIONAL MODEL work The incineration capability of the core of a large fast reactor in terms of the num- recycle On e period comprise cycleR years0 EF s(1 s5 t eac A ) h cycle from cycle ber of LWRs which may be cleaned-up by one FBR is determined Special atten- 6 on the reload fuel consists of several streams of fresh U, Pu fuel and of por- U Pu 800 0

7000-

FP and Losses of 600 0- U Pu, Np, Am

Np, Am o . / fldmcxturRr 8 , 2. Np f o e / 0 7 m fl , /. 1 Losse p N : s 4000- ———— Np237 frou P dep m, U LWRl s ---- Pu238 3000- S/A Fabrication - —> Losses Am241 remote Am243 200 0-

Pu, Np, Am (Pu,U,Np,Am)-MOX 100 0- homogeneous

50 10 0 15 0 20 0 25 0 30 0 350 d LosseFPan f so Time (Years) of EFR Operation m A , Np , UPu , Fig 2 In—Pile BOG Nuclide Masses During Recycling

4 RESULTS ) % 1 ( % 0 7 Reprocessin m A , % 1 p N g , LossePu , U s 4IN-PIL1 E CHARACTERISTICS Np, Am Loading in the EFR 2 8 - 7 5 % From Fig 2, showing the evolution of the masses of Pu238, Np237, Am241 and Am243 during recycling, a tendency to reach equilibrium concentrations after 1 Schem . Fig f Minoeo r Actmide Incineration Including Recyclinu P f go about year6 3 cycles 8 s(1 firse yearn states i th )te t n I sd whe n therselfo n -s ei generated Np, Am available, LWR Np, Am, mainly consisting of Np237, is domi- whicm A hd hav an througn r morp o eru N e , e tionhon Pu recycl f so e periods nating In the course of recycling when, because of incomplete burning, less LWR Sinc burnine eth g behaviou homogeneousle th f o r relativeln i y y small amounts material can be charged to the EFR, the content of Np decreases whereas the admixed materials approximately is determined by the neutron spectrum of the self-generated part is gaming importance The strongly reduced external load af- dominating fresh-fuel part, the relative variation of the isotopes of Np, Am and Pu seee yearb n te n wher3 r frosca te g numbee mFi eth f cleaned-uo r p LWRs i s from LWRs and the EFR during recycling can be pre-calculated by using only two depicted versus the time of EFR operation sets of effective EFR neutron cross sections for the inner and outer core, namely equilibriue th , I b mTa n m-pilI e masse minoe th f o sr actmide nuclidef o d an s frese thath hf o tfuel s Pu238 maximue th , m Pu238/Pu numbee th d f referencan ,ro e LWRse b whic y hma To account (a) for the different compositions of the external LWR waste and the mmor-ac% 75 d equilibriumn i an - R cleaned-u0 5 EF , 8 e listee 2 th ar ,r y dpfo b EFR self-generated Np, Am and Pu, and (b) for the different spectral conditions in % tmide70 f o sd admixturean p r reprocessinN fo d d an san u P r gfo losse% 1 f so the inner and outer EFR core for both wastes and for the inner and outer core for Am separate pre-calculations for 1 IHM LWR and EFR reference fuels were performed minor-actimde% Fo8 2 r s r largeadmixturFo u r P admixture f eo Pu23% 6 s 8i s wit KORIGEe hth N cod] e[2 Pu238 increase moru o st P f eo tha% n10 f recyclino Tht ese g equations space base timth d n ean edo fuel mass balances The number of cleaned-up LWRs significantly increases with the admixture of Np is solved with respec nuclide th o t t e concentration t begisa f eacno h cycle using and Am for 5% admixture already 11 LWRs could be cleaned up Then howev- CO Oï actual EFR and LWR fuel masses and the above given recycling scheme er a Pu238 content of 9% would have to be dealt with 10% CO 0>

30.0- X Losse 0 7 R s , X N1 p Ct a r 3; ö of Np f fltn : ——% 8 —2. ---- 5.0% ...... 7.5 o % (X

10'S 2.8 X fldmlxture of Np, flm 40 Yeors of EFR Operotlon

JO - - - - Losses : Np 1 %, Am 1 % ...... i_osses : Np 1 %, Am 70 Referenc— —— R eLW

Kf 10.0 15.0 20.0 ° 10 25.0 s 30.10 0 * 10 3 10 ? 3*101C ' Time (Years) of EFR Operation Time (Years) after Begin of EFR Operation

: Incineratio Fig3 . nR CapabilitEF e th f yo Fig. 4 : Waste Ingest. Hazard without Equilibr. Contribution

TAB. I : EFR CORE EQUILIBRIUM CHARACTERISTICS DEPENDENT ON NP, AM In a recent work of Küsters and Wiese [4], former toxicity indices had been used ADMIXTURE giving rise to a marked contribution of Ra226 after about 105 years of storage. This pea decreases ki toxicite th f di y values fro appliedme ar Ref ] [3 .. Qantity Unit Parameters and Calculated Values NpAdmixturm A , e % 2.8 5.0 7.5 sho5 Figure rise d wth k an potentias4 waste th f eo l electrica a actinideGW r lspe Np237 kg 720. 1329. 2020. output dependent on the time after begin of the EFR operation in the case of 2.8% wastreference e th Th f comprisee o . R minoee Am LW admixturth d l sran al p N f eo Am241 kg 168. 236. 315. actinides produced during LWR operation and the reprocessing losses of U and Am243 Kg 109. 148. 193. Pu. The other curves show the hazards of the waste actinides from the LWR-EFR system for 70 and 1 % losses of Am. These waste actinides consist of the accu- 613. 1036. 1518. Pu238 kg mulated losses from reprocessing Pu238/Pu % 6.0 9.0 11.6 • of the recycle burner fuel, including U and Pu losses Cleaned-Up LWRs 6.0 11.3 f whico 17. m 4insertes hi A d an dp intEFR N e fuel R oth e ,th LW includin e th f o gU • and Pu losses

U recovered from burner fuel reprocessing, and U and Pu recovered from LWR 4.2. MEDIUM LONG-TERD AN - M ACTINIDE RISK POTENTIALS fuel reprocessing assumee ar , usee b r late do dfo t r fabricatio f fueno Fig- l . .1 thin I s work rise th ,k potentia r hazard(o l determines i ) maximue unitn di th f so m Excluding the burner inventory from the hazard calculation - as done in the cal- allowed annual radioactivity ingestion as given in Ref. [3] from 1989. Probabilities culation r Fig s- meanfo 4 . s that burner reactor continousle sar y installee ar d dan of migratio hazardoue th f no s nuclei fro finae mth l depositor biosphere th o yt e ear actin interis ga m storag hazardoue th r efo s nuclei thin I . s case onl reproce yth - not accounted for if the risk potential is measured in this way. essing losses finallhave b o et y dispose. dof I 2 8 / flaWxture of Np, Rm o 40 Years of EFR Operation l. v a o 10" % ---1 m -A % Losse 1 p N s -o o c Losses Np 1 %, Am 70 % ü ——— Reference LWR < o •g 10" TD

o o < "o / m floWxtur fl 28 , Np f eo -H10'0 40 Years of EFR Operation o N O os I % ---1 m -A % Losse 1 p N s Losses Np 1 % Am 70 % 10° , Referenc— —— R eLW * 10 5 10 " 10 f it f it 3*10' 10 2*10' Time (Years) afte Operatior R BegiEF f no n s 10 5 10 3*10' * 1010 3 f it 10 l WastRe e6 Inges g Fi t Hazoro without Equilibr Contr Time (Years) after Begin of EFR Operation

Fig 5 Waste Ingestion Hazard with Equilibr Contribution

O 25- m RdmLxtur. R X , 8 Np 2 f o e If however, on the other hand the operation of burners is stopped, then the equi- OperatioR 0 Year4 EF f so n librium inventory will have to be finally disposed of and will then contribute to the waste hazardhazarde Th o thin si ,to s 5 cas showe g ear Fi nm --— Losses Np 1 %. Am 1 % From Fig 4, the strong effect of the losses of Am with respect to the hazards can 20- X| Losses Np 1 %, Am 70 % be seen In the case of probably unrealistic Am losses of 1%, a reduction of up C. — Referenc—— R eLW ordee t oon f magnititud o r calculates ei comparison di hazare refe th th o n-f t d o 15- erenc exceptime R th e eLW n i regiot n around 5000 year r estimateFo s d losses o reductio a f 70%o , hazare m oA th f f r nstorago d fo e times larger tha yearss n10 , -o long-tere th r fo me i hazard states i f abou o , % d50 t o The increase of the hazard around 5000 years is mainly due to Pu240 and, less pronounced, to Pu240 originating from reprocessing losses and from decay of Am243 and Cm244 Fig 5 shows that in the case of stopping the operation of burner reactors a smaller reductio long-tere th f no m hazar achieves di d A more drastic representation of the incineration ability is given in Figures 6 and 7 showing the ratios of the actmide hazards of the LWR-EFR system and the ha zards of the reference LWR 3*10' 10" 105 10* 1Cf 10* 10 In the case of large losses of Am the generation of Cm244 from Am243 and hence Time (Years) after Begin of EFR Operation CD the build-up of Pu240 is reduced This leads to a decrease of the peak at 5000 years compared to the case of low losses of Am In the long term range where l WastRe Fi e7 g Ingestio n Hazard with Equilibr Contnb CO Np237 and its daughters dominate the hazard the relatively large amount of re- REFERENCES 00 cycled Np in the case of large Am losses causes a worsening of the burning abil- ity [1] H Küsters, M Salvatores, R Corcuera, B Krieg, J Tommasi, H W Wiese, To improve (he incineration capability in the medium-term range, a separation of A Zaetta "Recent Investigations in France and Germany on the Transmuta- Cm during reprocessin subsequen a daughtee d th fues f gan a o lu e woulP r us t d tion of Actmide and Fission Products in Reactors", to be presented on the be helpful! This is also concluded by Zaetta in Ref [1] However, also Cm is International Conferenc "Design eo Safetd nan f Advanceyo d Nuclear closely boun raree PUREe th th o dt n earth i X d processan separatios it s s a s ni Power Plants (ANP'92)", October 25-29, 1992, Tokyo difficuln thaAr s f a t o t [2] U Fischer, H W Wiese "Verbesserte konsistente Berechnung des nuklearen The plutonium peaks around 5000 years were also found by Corcuera [1] who in- Inventars abgebrannter DWR-Brennstoffe auf der Basis von Zell-Abbrand- vestigated the incineration in LWRs as burners However, since LWRs contain less Verfahre t KORIGEN"nmi , KFK-Report 3014 (1983), ORNL-tr-5043 hencd Pan u e less transplutonium elements, this pea s lesi k s pronouncen i d [3] K Topfer, Bundesminister für Umwelt, Naturschutz und Reaktorsicherheit "Neufassung der Strahlenschutzverordnung", Bundesgesetzblatt Teil 1, LWR-LWR burner systems 2 Jul1 , i N19834 r 9 KüstersH ] Wies[3 W H , e "Burnin f Actinidego Fissiod san n Productn si 5 SUMMARY Reactor Accelerator-Drivem d san n Spallation Sources", ProcS AN . Winter Meetin Franciscon gSa , Tran 641-754,O sA 1991 The investigations show tha r estimatefo t dcomn i realisti% - 70 f co lossem A f so WehmanK U ] [4 n "Some Physics Aspect f Minoso r Actmide Recyclin Fasn gi t parison to the non-incineration case Reactors", this Meeting

long-tere th • m actmide hazards hazarde th r storage l ,sfo e times larger than 10 years, can be reduced to about 40-50%, s • for the medium-term range between 1000 and 50000 years of storage, the Pu generateburnePu the in the r decafuedvia and Cm24l of y 4 gives risan eto increase of the actmide hazard of 50%

The incineration ability could be improved long-tere th n i m • rang reductioa y eb reprocessine th f no g , losseAm f so medium-tere th n i • m subsequenranga separatioe a d th y f an eb o m e C f us tn o daughte fues a lu P r

It is concluded that an enforced research should be done in chemistry to prove that a satisfactory separation especially of Am and also of Cm is possible on large scale

One large fast reactor of type EFR is able to clean-up six 1300MWe, PWRs in equilibrium in case of 2 8% LWR and recycled Np and Am homogeneously ad- mixed to the core fuel Then the fraction of Pu238 in the core Pu is limited to 6% Larger admixtures possiblseee b mo t n smalleei r cores with respec safeto t t y parameters [4] t the ,e contenbu th n f Pu23o t 8 will strongly increase causing problems with heat and radiation during fabrication of recycle fuel

ACKNOWLEDGEMENT

The authors gratefully acknowledge the numerous fruitfull discussions with KüsterH r D s CHARACTERISTIC TRANSMUTATIOTRU SOF N I. INTRODUCTION LMFBN INA R Ther strona s ei g social requiremen leavo t radiologicae t eth no t l hazardous material form the use of nuclear power to future generations. Some of the T. WAKABAYASHI, M. YAMAOKA Oarai Engineering Center, transuranic (TRU) nuclides containe residuan di l waste from reprocessing have Power Reactor and Nuclear Fuel extremely long-term radiotoxicityU). Ther e somar e e mean f reducino s g Development Corporation, radiotoxicity of the TRU nuclides under investigation(2),(3),(4). The TRU nuclides Oarai-machi, Ibaraki-ken, Japan produce useful energy when converted into short-lived fission products by neutron bombardment. From this standpoint nucleaa , r reactor provide extremeln a s y Abstract rational means for transmutation of TRU nuclides. Among the various nuclear reactor sodium-coolee th s, d liquid metal fast breeder reactor (LMFBR) undew no , r Feasibility studies of TRU transmutation in an LMFBR has been performed to establish TRU (exactly minor actinides) transmutation technology by in an LMFBR. development, can be used for transmutation of many TRU nuclides, because of the Systematic parameter survey calculations were performe a conventiona r fo d l possible nuclear fission generated by high-energy neutrons. lOOOMWe LMFBR core to investigate basic characteristics of TRU transmutation in The following studie e implementear s establiso dt (exactlU hTR y minor an LMFBR core and also to establish TRU loading method which has no serious actinides) transmutation technolog LMFBRy yb PNC-Japan si n: influence on core design. TRU loading resulted in a decrease of the burnup reactivity loss, which is a desirable feature for the extension of reactor operation • Feasibility studies of TRU transmutation by LMFBR and evaluation of period. TRU transmutation rate reaches approximately 11% per cycle and the TRU material balance amount of the TRU transmutation with TRU ratio of 5% is almost six times as much • Nuclear data evaluation of TRU nuclides in sample irradiation experiments as that of the TRU from a lOOOMWe-class LWR. Homogeneous TRU-loading has •Measuremen evaluatiod an t physicaf no chemicad an l l propertieU TR f so almost no serious influence on core performances. However, it is desirable to limit compounds loadinU TR e g th rati severao ot l percent fro e aspecmth reactof o t r operational •Development of fabrication technology of TRU fuel safety. Heterogeneous TRU-loading method resulte largea n di r power swing. •Evaluation of fuel behavior by TRU pin irradiation experiments Further optimizatio necessarns i clarifyo t feasibilite yth y method e ofth . This paper shows the results of the following two feasibility studies of TRU Study on an innovative core concept for TRU transmutation was also carried out transmutation in an LMFBR : taking into accoun resulte th t f systematiso c parameter survey calculations. Neutronic feasibility was studied on a lOOOMWe Super Long-Life Core (SLLC) with (1) Study on core characteristics of an LMFBR loaded with TRU fuel. no nee fuef do l exchange during plan t SLLe lifeattractivn a Th .Cs i e optior nfo In this study, systematic parameter survey calculations have been transmuting TRU nuclides since it can transmute them with confining them into performe conventionaa r dfo l lOOOMWe-class LMFBR purposcoree Th . s ei the reactor during plant life. It was found that remarkable reduction of both (i)to investigate basic characteristics of TRU transmutation in an LMFBR reactivity change and power variation during burnup is possible by optimizing the cor(iid establiso )et an loadinU hTR g method seriouwhico n s hha s influence amoun loadedzonind U reactivite an t TR Th f go . y powee changth d r ean swin g on core design. during 30 years of the optimized core are less than half of those of the SLLC with no TRU loaded. The amount of TRU transmuted is about lOton which corresponds to (2) Stud feasibilitn yo Super-Long-Life-Cora f yo e (SLLC) loaded witfueU lhTR amoune produceU th LWRsTR 3 1 f SLLe o t y dTh b C. loaded s fuewitU wa lh TR which has no need of fuel exchange during plant life found feasible fro neutronie mth c point of view. Taking into accoun advantage th t TRU-loadee th f eo d core lOOOMWa , e R&D programs are being pursued in PNC-Japan to establish the SLLC wit neeo hn f fue do l exchange during plant lif s proposei e n a s a d CO CO transmutation technolog nuclideyU ofTR LMFBRsn si . LMFBR core for TRU transmutation. The SLLC is an attractive option transmutin nuclideU gTR s transmutn sincca t ei e them with confining them o o into the reactor during plant life. A feasibility of the SLLC was studied form i 200 neutronie th c view point. \ -^ Inner core ICCO 1f* Outer core The subsequent two section describes the mam results of the studies. The last _ Axia1 l-i blankel t 2ÖÖ e establishmenth sectio r fo prograU D nC TR showR& PN f e n o mti th s 1 transmutation technology in an LMFBR. r-——————— 3680 ——————— ——————— 4000 ————————— r H. STUD CORN YO E CHARACTERISTIC LMFBN A F SRO LOADED WTTU HTR r.-un FUEL

. Calculationa1 l method First a lOOOMWe-clas, s LMFBR cor s define efuee X witth wa l s hMO a d reference whose main parameters are shown in Table 1 and Fig. 1. The nuclear characteristic TRU-loadef so d cors calculateewa two-dimensionay db l diffusion

Tabl eDesig1 n Parameter Conventionaf so l lOOOMWe LMFBR Core

Parameters Data

Thermal Power 2520(MW) Operation Cycle Length 456(EFPD*)

Fuel Material Pu02/UO2 SOitSlSlSlSlSlS@S@§ISli©SliI*I* Number of Subassemblies . 8 . 79 Inner Core/Outer Core 175/180 Radial Blanket 72 Control Rods 24 Core Height 100(cm) O Inner core fuel 175 Equivalent Core Diameter 368(cm) (o) Outer core fuel 180 Axial Blanket Thickness 30(cm) (?) Primary shutdown control rods 18 Numbe f Fueo r l Batch ) Backu(a 6 p control rods (Core/Blanket) 3/4 0 Radial blanket 72 Fue Diametern Pi l 0.83(cm) (S) Neutron shield (Stainless Steel) 78 Assembly Lattice Pitch 17.98(cm) ) Neutrog 7 n27 shield (B C) Volume Fraction 4 (total 799) Fuel/Coolant/Structure 41.6/37.5/20 9(%) *EfFective Full Power Days Fig 1 Core Layout of Conventional 1000 MWe LMFBR theory with depletion chain. The cross sections used were effective seven group 1,000 constants condensed fro Japanese mth e standard 70-group constant set, JFS-3-J2(5), »Values in parentheses denote TRU transmutation rate per cycle whic baseevaluatehs n i a n do d nuclear data library, JENDL-2(6). 800 assumes fueU wa l TR comdo t espen R fromtLW fuel with five-year cooling time before reprocessing. The isotopic composition of the TRU shown in Table 2 was (12.5%) 600 calculated by the ORIGEN2 codeCT. . Surve2 loadinyU ofTR g ratio 400 Influence of TRU loading ratio was analyzed with TRU homogeneously CJ (12.6%) disperse maicoree e th Th n.n d i result s were shown below. 200 Figur showe2 relationshie sth p betwee amoune nth transmuteU t ofTR d dan core R t least A percenloadine FB . U e on , th TR tha f fueo n g to loadet i f to l U dTR S (11.5%) < woul needee db eliminato dt e amoune transmute U TRUth d TR f an ,o t d increases linearly with that of loaded TRU. Since a lOOOMWe-class LWR produces about 26 kg of TRU per year, an LMFBR with 5% TRU-loading can transmute the TRU mass _L JL fro LWRx msi rougn si h estimation. -200 5 2 0 2 5 1 0 1 5 0 AloadinU s seeTR Fign corno i gt , 3 .e result significana n si t decreasf eo Ratio of TRU amount loaded in fuel ( % ) burnup reactivity loss mainl238po t e uydu build-up. However e burnuth , p reactivity loss becomes positive whe e amounnth loadef o texceed U dTR s 10%. Fig. 2 Relationship between Amount of TRU Transmuted and Ratio of TRU Amount loade Fuen di l

Therefore maximue th , m amount of loade woulU limitede TR d b severado t l percent from the aspect of reactor operational safety. A proper amount of TRU might be Table 2 Composition of TRU Fuel * advantageou extensioe th so t n of reactor operation period.

Weight Fraction founs wa dt loadin I thaU tTR g tend increaso st burnue eth p change globa e ofth l Nuclide (%) power distribution. However, it could be possible to minimize the power change by optimize the ratio of TRU amount between the inner core and the outer core. Np-237 49.1 Am-241 30.0 3. Survey of TRU loading methods Am-242m 0.08 We consider here two typical TRU-loading methods in the LMFBR core. One is Am-243 15.5 the method dispersing TRU homogeneously throughout the entire core Cm-243 0.05 (homogeneous TRU-loading method), which is expected not to affect the core Cm-244 5.0 characteristics seriously. The other possible method is the use of small number of Cm-245 0.26 subassemblies (target S/As) which concentrate TRU in MOX fuel (heterogeneous * Discharged from PWR (35GWd/t) and TRU-loading method), whic havn he advantagca eth fuee th lf eo cycl e becausf eo Cooled for 5 Years Before Reprocessing handling the small number of TRU-included fuels in fabrication factory. (93%)

0 5 10 15 20 Ratio of TRU amount loaded m fuel (%)

Fig 3 Relationship between burnup reactivity loss and ratio of TRU amount loade fuedm l

9 Target S/A including TRU A typica loadinU TR l g cors assumeewa e homogeneou eacr th dfo f ho d an s 4 Loading Fi g pattern of target S/As heterogeneous TRU loading methods The total mass of loaded TRU is set to be ( Heterogeneou loadeU TR sd core) approximately identical betwee whole th n eni corthemU e TR fueth , % than i 5 l , tis case of the homogeneous method, and 37 target S/As which have 50% of TRU in fuel equilibrium cycle This power swing woul a grea e db t obstacl r thermafo e l e heterogeneou case th th f eo m s method. Figur showe4 loadine sth g patterf no characteristics Slight decreas e maximuth f o e m linear heae t th rat n i e TRU-loaded heterogeneouS/Ae th sm s metho cor loadeA dU e als s witTR dowa o hn heterogeneous TRU-loaded cor attributes ei loadin U regioe th TR n go i dnt where analyze reference th s da e case peae th k powe loades i U r doccurTR o n f si Table 3 compares the calculated results about the nuclear characteristics of the The control rod worth of the TRU loaded cores decreases from the reference core two TRU-loade referencde coreth d san e core by 10—20% cause e hardenine b Thi th y y dsb ma g of neutron spectrum burnue Th ploade U reactivit TR smalled e % coreth 40 n ys i sri thane thath f o t The maximum linear heat rat loadeeacf U eo muco hTR s dt hcorno differen s ei t reference core because of the production of 238p from 237]\fp m TRU fuel with that of the reference core The radial power distribution of the heterogeneous u TRU-loaded cor ehowever, quits ,i e different from tha tothe e ofth r core shows sa nm The Doppler coefficients of the TRU loaded cores are 20-30% smaller in Fig 5 At beginning of the equilibrium cycle, the power of TRU-loaded region is absolut esodiue valueth d man , density reactivity coefficient large% 50 r e thaar s n very depressed compared wit reference hth e core, e whilth f eo thed t closen y ge t ea the reference core because of the the spectrum hardening Power density (MW/cm3) Power density (MW/cm3) Table 3 Comparison of Nuclear Characteristics of TRU-Loading Cores

Reference Core Homogeneous Heterogeneous Item (No TRU- TRU-Loading TRU-Loading loaded) Core Core

Pu Enrichment 15.3/19.3 wt% 16.2/19.% 6wt 15.7/19.% 7wt (Inner Core/Outer Core) Max.Linear Heat Rate (BOEC/EOEC) (BOEC/EOEC) (BOEC/EOEC) Inner Core 380/419 w/cm 376/431 w/cm 386/428 w/cm Outer Core 420/357 w/cm 416/355 w/cm 439/342 w/cm

-3 3.3%Ak/kk' 1.9%Ak/kk' 1.8%Ak/kk' » Burnup Reactivity Loss c Control Rod Worth e- p (BOEC 33cm Insertiof no 1.7%Ak/kk' 1.5%Ak/kk' 1.5%Ak/kk' Primary Rods) (1.00)* (0.88)* (0.88)* ra as -1.1X10-2 -7.1X10-3 -7.4X10-3 o o Doppler Coefficient H O Tdk/dT Tdk/dT Tdk/dT Coolant Density Coefficient Power density (MW/cm3) •o g" Power density (MW/cm3) - . ~ û. (Ap /p/100%Density Change) -1.7X10-2 -2.5X10-2 -2.7X10-2 £ 3 5 3 o — ' ßeff 3.7X10-3 3.5X10-3 3.3X10-3 Amount of TRU -46kg 184kg 176kg s-1 transmute r cycldpe e (11.7%) (10.6%) T3 S" *) Value Parenthesen si s Denote Relative Contro Wortd lRo h

valuee Th prompf so t neutron life timeffectivd ean e delayed neutron yieln di the TRU-loaded 10%d corean , smalles si % respectively 20 y rb , than e thosth f eo reference core. Both TRU-loaded core n transmut y ca ll~12%/cycls b U TR e e (about G C 9- 180kg/cycle) ther differencd o ,n an es i loadinU e betweeTR o g methodstw e nth . maximue Th m cladding temperatur eacf eo h TRU-loaded corevaluates ewa n do condition tha coolane th t t flow distributio same core e th s thath th s ea e i f n ni o t B o reference core. There is no thermal problems about the homogeneous TRU-loaded O nO CO o core, sinc powee eth r distribution hardly changes froreference mth e e th core n O . other hand, the position where maximum temperature of cladding occurs in the Tabl e4 Desig n Parameter lOOOMWf so e heterogeneous TRU-loading core moves from the reference core, and the hot spot Super long-Life LMFBR Core temperatur claddine th f degrees-eo 0 4 gs i C higher tha nreference thath f to e core. Although there might be rooms for optimization of the flow distribution, the Parameters Data significant power swing is a disadvantage to thermal characteristics. Thermal Power 2520(MW) However, this problem may be overcome by using target S/As with a smaller Core life 30(EFPY*) amount of TRU or loading them in a more scattered manner. Further optimization Fuel Material Pu02-UO2 necessars i clarifyo t feasibilite yth y method e ofth . -(Np,Am,Cm)O2 4. Conclusion Number of Subassemblies TRU loading results in a decrease of the burnup reactivity loss, which is a Core/Blanket/Control Rods 462/84/49 desirable featur extensioe th r efo n of reactor operation periodtransmutatioU TR . n Core Height 180(cm) rate reaches approximately 11% per cycle and the amount of the TRU transmutation Equivalent Core Diameter 494(cm) with TRU ratio of 5% is almost six times as much as that of the TRU from a Axial Blanket Thickness 15(cm) lOOOMWe-class LWR. Homogeneous TRU-loadin almoss gha seriouo tn s influence Fuel Pin Diameter 1.22(cm) on core performances. However, it is desirable to limit the TRU loading ratio to several percent from the aspect of reactor operational safety. Heterogeneous TRU- Assembly Lattice Pitch 20.8(cm) loading method resulted in a larger power swing. Further optimization is necessary Volume Fraction to clarif feasibilite yth methode th f yo . FueVCoolant/Structure 53.0/24.7/22.3(%) Pu Isotopic Ratio 58/24/14/4(w/o) . STUDm POSSIBILITN YO SUPER-LONG-LIFE-CORA F YO E (SLLC) »Effective Full Power Year LOADED WITH TRU FUEL

Calculationa. 1 l model Core neutronics calculations wer ee two-dimensiona donth y eb burnuZ R l p e feasibilitTh Super-Long-Life-Core 100e th f 0yo MW e (SLLC) loaded with code based on the diffusion theory using 7-group effective cross sections collapsed TRU studieds fue wa desige l Th .n parameter showe core sar Tabln Th eni . e4 from JFS-3-J2 library(S) based on JENDL-2C6). The isotopic composition of the reactivity life should be 30 years (effectively) in order to accomplish reactor e samTRth s thaUs ea i stude t useth sectiof ye corn o di Th e(TablnH . e2) operation without fuel exchange during plant life time thin I .s study amoune th , t performance e optimizeth f o s d core were compared with e SLLthosth f Ceo with and the region of TRU loading were surveyed so that burnup change of reactivity mixed oxide fuel. powed an r distribution coul minimizee db d assuming tha dimensione tth core th ef so 2. Results of parametric survey same SLLth n cor those s a fuee a e s th Ceth a ar f l e o d wit an h mixed oxide fuel(8) The core layout is shown in Fig. 6. A large fuel pin of 12mm diameter is employed Figure 7 illustrates the effect of the TRU loading ratio on the variation of the in order to enhance the internal breeding, which resulted in a fuel volume ratio of effective multiplication factor (keff) during burnup with uniform TRU-loading for 50%. The core volume is chosen to be about 4 times larger than that of the loadeds wholi U reactivite eth TR , coro n ef I Figf (Sey o . chang) .e7) (1 e durin0 g3 conventional core with the same power output so that the average fuel burnup effective full-power years (EFPY) is over 10%Ak and the reactivity life is about 25 during life tim abous ei t 200GWd/t EFPY. (The reactivity life coul extendee db d mor increasiny b e e plutoniugth m Axial Blanket Effective Multiplication Factor Outer CoreI p L __ ___ vo o H- « i_ o 4) *—< o

c OJ D OJ C o O 0 GQ u u m c u Inner Core «3 c ro Q> Q> «1 T3 •o o ID O 8 a: •o I O ro S Outer Core I 5

Axial Blanket

za n> 4940

5330 O (Control Rods not Shown) n ro Fig. 6 Core Layout of Super Long Life Core (Vertical View) tu ro

c enrichment. However t wouli , d increase reactivity change. loadinU e TR th )gn i •o tt> core resulted in significant reduction of reactivity change. The figure shows that o Effective Multiplication Factor TRU loading ratio of about 10% makes reactivity change of about 5%Ak possible. 3 ta p o Figur illustratee8 variatioe sth n powee ofth r distribution during burnur pfo in o core loadeth U e thawitd TR dan o t hn wit h wholr uniforfo loadinU e% m10 TR f go O core. (See cases 1 and 2 of Fig. 8) The figure shows that uniform TRU-loading for o whole core resulte significandn i t variation powe e ofth r distribution during burnup. a. ~uO c n> This means that the uniform TRU-loading for whole core accelerates the power shift due to burnup from the outer core to the inner core. It seems that this effect became CO c conspicuous due to the very long core operation time. §• i I -H . • Loading TRU in the outer core subassembly more than in the inner core sub- C ai m T3 - assembly is considered one of the effective method for suppressing variation of the S- c

power distribution during burnup. Figure 8 illustrates the variation of the power -o O o Cu distribution during burnup for the core with TRU loading ratio of 7.5% and 15% for Q. D the inner core and the outer core, respectively.(See case 3 of Fig. 8). This non- (Q n n o uniform loading of TRU fuel resulted in very constant power distribution during tu burnup. Figure 7 illustrates the reactivity change of the core with non-uniform o Ol TRU-loading core (See (2) of Fig. 7). From the results shown above, we finally o Tabl 5 e Core Performance f Supeo s r Long Life Core en LoadeU NoTR d TRU Loaded Core Core

Loading ratioU ofTR — 75/150* [w/o]

Reactivity Change [%Ak] 133 52

Average Power Density 79/75** 80/75** [w/cc]

120 Max Linear Heat Rate 268 266 5 Case 2 TRU Loading Ratio [w/cm] 1. 100 En l Lifdo e = 100% for Whole Core Power Swing*** 21 4 472 5 so o [%] O- _>, 60 Fuel Burnup [GWd/t] 188/262 Mtddk of Lift 2 29 / 7 18 JD E 40 (Averag) x Ma e / (Average / Max ) 69X1023 66X1023 < 20 Max Fast Neutron Fluence [nvt]

34 567 89 10 11 12 Max Cladding Temp not calculated 655 Radial Region Numbe* r [°C]

12 0 Sodium Void Reactivity 0014/0030** 0026/0031**

LoadinU TR Cas ge3 /15% Rati5 7 % o= [Ak] for Inner Core / Outer Core Subassemblies Doppler Coefficient 83/47** 46/27** E-lXl03Tdk/dt] *Inner core assembly / Outer core assembly **Begmnmgoflife /End of life *** Average of maximum power change of each assembly during life time

4 3 5678 2 1 1 1 0 1 9 selected the TRU loading ratio of 7 5% and 15% for the inner core and the outer core, Radial Region Number* respectively The reactivity change during 30 EFPY is about 5% ('Control rod regions not included) The results of parametric survey show that remarkable reduction of both

Fig 8 Variation of Assembly Power Distribution of reactivity change and power variation during burnup is possible by optimizing the Super Long Life Core due to Burnup amoun loadezonind U tan TR df go 3. Analysis of nuclear characteristics Table 6 TRU Mass Balance of Super Long-Life Core Nuclear characteristics for the optimized core were analyzed and compared (unit: ton) with those of the core with no TRU fuel. The results are summarized in Table 5. In Loaded Discharged this study controo n , strategd ro l y during burnu consideres reactoe pwa th d rdan Difference amount amount (B-A) calculations were performe dcontroe witth l hal l rods withdrawn fro coree mth . (A) (B) The reactivity change of the TRU-loaded SLLC during 30 years (EFPY) is u 159 133 -26 about 5% and it is less than half of that of the SLLC with no TRU loaded. The Pu 20.4 25.5 5.1 values of the maximum linear heat rate for both of the SLLCs are about 270w/cm. 238p ( u) (0) (3.5) (3.5) They would increase to some extent if control rods are inserted in order to suppress (Pu Fissile)* (14.5) (14.8) (0.3) the excess reactivity e averagTh . e fuel burnu s aboupi t 190GWd/e th d an t maximum fast neutron fluence is about 7X 1023nvt. This irradiation condition is Np 8.4 2.6 -5.8 almost the same as the target of a commercial FBR core. Am 7.8 3.5 -4.3 The maximum cladding temperatur evaluates ewa condition do n thae th t Cm 0.9 1.0 0.1 reactor outle respectively, inled °C tan 5 t temperature37 d an . 0 Althoug53 e sar t hi Np+Am+Cm 17.2 7.1 -10.1 preliminaris y sinc controeno insertiolrod n effec considereis t calculatiothe din nof *239Pu 2 + the power distribution e maximuth , m cladding temperatur enougw lo s i eh compared wit limis hit t value (typically °C)0 . 70 , loadingU Dusodiue TR th o et , m void reactivity increase absolute th d dan e LWRsb3 y1 SLLe Th .C loaded foun s fuewitU wa l hdTR feasible fro neutronie mth c value Doppleth f eo r coefficient decreased safete Th y. aspece SLLth f Co t loaded point of view. with TRU is one of the problem to be investigated, although a lot of parameters other than these parameters also contribute to the core safety. IV. R&D ISSUES FOR TRU TRANSMUTATION IN AN LMFBR Table 6 shows mass balance of TRU. Total loading amount of Np, Am, and Cm is about 17ton and lOton of it (about 60% of loading amount) is transmuted programD R& PNC-Japae n si th Show s i e transmutatio 9 Fign th ni . o nt n during the 30-year reactor operation. Transmuted amount of Np, Am, and Cm is technology of TRU nuclides in LMFBRs. yearr pe g ,abou k whic0 t34 nearls hi y equaamountoe th o lt f those from ISLWRf so Through design study of an LMFBR transmuting TRU and evaluation of TRU the same power output. Total amount of 239p and 24lp i slightly increased u u s mass balance, it is planned to establish appropriate core design concepts for a large- during operation. Thus, the SLLC has an advantage of transmuting large amount of size LMFBR and strategy of their introduction. Basic researches including nuclear TRU with preserving plutonium fissile material. data evaluation d fuean ,l property measuremen evaluatiod an t e alsar no 4. Conclusion programmed. It was found that remarkable reduction of both reactivity change and power "Joyo" an experimental FBR, is planned to be used both for initial variation during burnu s possiblpi optimiziny eb amoune gth zoninU d an tTR f go transmutation experiments irradiating fuel samples, and for oxide fuel irradiating loaded reactivite Th . y powechange th d re an swin g durin yearoptimizee 0 g3 th f so d testpost-irradiatiod an s n tests addition I . , n basic test e plannear s fuen do l cor e lesear s than hal SLLe thosf fo th loadedf U Ceo amoune witTR Th o .h n f o t preparation of TRU obtained from the partitioning of nuclides , the results of which TRU transmuted is about lOton which corresponds to the amount of TRU produced will be reflected in fuel fabrication technology o REFERENCES 00 1. L. KOCH, Formation and Recycling of Minor Actinides in Nuclear Power a -3 Stations, Handboo Physice th Chemistrn d ko san Actinidese th f yo , Chapte, 9 r 5 North-Holland Physics Publishing, 1986. . WATCHE . CROFFW G . . 2J .A d , ActinidRan e Partitionin g- Transmutatio n Program Final Report ffl. Transmutation Studies k RidgOa , e National Laboratory report, ORNL/TM-6983,1980. pi fü g. g. . Mukaiyam3T . , Higheal t ae r Actinide Transmutation Using Higher Actinide -© -e -e Burner Reactors, Presente Minof o t T Workshoda r P- Actinidese th n po , Mito, Ibaraki, Japan, Oct. 16-18,1990. 80° CD 4. T. Inoue et al, Transmutation of Transuranium Elements by a Metallic Fuel CO FBR Using Higher Actinide Burner Reactors. Presented at Workshop on the P-T a. G of Minor Actinides, Mito, Ibaraki, Japan, Oct. 16-18,1990. 5. H. TAKANO, et al., Revision of Fast Reactor Group Constant Set JFS-3-J2. Japan Atomic Energy Research Institute report, JAERI-M 89-141,1989. ra CD . NAKAGAWA6T . , Summar JENDL-f yo 2 General Purpose File. Japan Atomic o CD to Energy Research Institute report, JAERI-M 84-103 ,1984. . CROFFG . 7A . User'A , s Manua ORIGENe th r fo l 2 Computer Code k RidgOa , e _Q -e National Laboratory report, ORNL/TM-7175,1980. a 8. M. KAWASHIMA, et al., Neutronic Feasibility of an LMFBR Super Long-Life CD Ct O CD Core (SLLC). Nucl., Engng., Design, 106,1988. f CO S -e

ö o CD CO co •a 0

CD CD o 01 -e- EXPERIMENTAL INVESTIGATIONS CONCERNING THE fission cross-sections e neutro(~5th n %i n energy rang f o e0.5 - PROBLEM OF MINOR ACTINIDE TRANSMUTATION -10.0 Mev). But the central reactivity worth (CRW) measurements wart adequatno e o identificatt e e capture cross-section- re a o t s S.M. BEDNYAKOV, V.A. DOULIN, I.P. MATVEENKO, quired accuracy. A.M. TSIBULYA, A.V. ZVONAREV A systematic investigatio transmutatioA M f o n n problem Institut f Physiceo Powed san r Engineering, started at 5*C assemblies in 1991. The first experiments were Obninsk, Russian Federation carrie t usinou d g absolute calibrated fission chamberd an s reactivity perturbations by a small MA samples. Abstract A systematic investigation of MA transmutation started at BFS ANALYSI A C (JAPAN(RUSSIA6« FC D F O SAN ! ! EXPERIMENTS assemblie n 1991i s e firsTh . t experiments were carrie t usinou d g absolute calibrated fission chambers and reactivity perturbations by a small MA sample. f o Pu-24W resultA CR 0ThFC n e waro s e compared with similar 5*C ones. Parameters of assemblies are given in Table I.

INTRODUCTION TABLE I. PARAMETERS OP ASSEMBLIES The proble minof mo r actinides (MA) transmutatioe b n ca n

solved by recycling of spent nuclear fuel in fast reactors. The Assembly Fuel Enrichment Neutron pio * evaluatio f periono d parameteran d recyclf so e requires validation % percentage |2ï fission and capture cross-sections of UA in a wide range of below 10 keV U 1-n spectrum neutron energies e achieve.Th d requiredan parenthesisn (i d - ac ) curacies of neutron cross-sections are: KKM5A-I U0? 21 4,7 -I.IO B4C-49-2 (Pu-U>02 12 13,1 -0.92 &cr t ,% av BSC-49-4 (Pu-U)Oj 12 6,7 -0.72 f' *vo * BW-5I-I U 36 0,4 -0.72 EÏC-55-I Pu-U 10 2,4 -0.60 Np-237 7(5) 15(5) 30(10) KBP-IO uo2 90 16,4 -1.54 Pu-238 10(5) 25(10) 40 FCA-IX-I U 93 17,6 - U - Pu-240 10(10) 20(4) 20(15) F CA-IX-2 93 9,1 FCA-IX-3 U 93 5,0 - 20 Pu-241 8(5) 15(7) FCA-IX-4 U 93 0,9 - Am-241 10(5) 10(5) 30(10) FCA-IX-5 u 93 0,4 - Am-243 10(5) 30(5) 30 FCA-H-6 u 93 0,1 - FCA-IX-7 u 20 0,04 - Cm-244 10(5) 30 30 B4C-6I Pu-U 15 4,0 -0.73 HDC-54-4 uo? 22 4.6 -1,08 e firsth On f f to o experimente wid t . criticaa se e n o s l BK-56-IA (Pu-U)02 16 6,6 -0.76 assemblie thin o s s proble carries FCn i mwa At dou (JAERI ) /!/. KW-56-IE (Pu-U)02 14 5,5 -0.70 Analisy f theso s e experiment / showe/2 s agreemenn a d f o microt - O

d macroexparimentan r fo Np-237s , Pu-238, Am-24 d Am-24an 1 3 a) Celcul*tion*l CR',7 rnt'os oBf tUo ?35 CO 1O TABL . ASSESSMENEII EXPERIMENTP TO NP-?3N SO PU-?3O ?T 9 CRW RATIO R SOMSFO E ASSEMBLIES t) «/ A. FCA.:

Assembly Experi- Calculations, BHAB-90 (C) C-E ment HI-" N !,„ -.0077 b

(E) N1M,-0 Ml«, = =.0059 0.05 - v1 horn het het het

FCA-IX-I -0.865 -0.896 -1.05 -0.8II -0.744 O.I2I FCA-IX-2 -0.242 -0.184 -0.248 -0.196 -0.183 0.059 FCA-IX-7 O.II7 0.188 0.187 0.194 0.187 0.070 -0.05 o -PCA (e***o) KB, P&K - • B. BSC-54-4: -0.10

Correc t i ons ÄW35 Self-shielding for '1Kb"1) f"echun hp- -0.15 teroge'-e- (1=0) 0.00220 0.00283 0.00456 'ty

Eiper-'nent -0.220Î -0.004 -0.005 -0.017 - -0.20 ^.007

Calculet^on - -0.005 -0.006 -0.010 -0.027 0 10 20 30 1

Pig.1.Central reactivity worth f 240, o KBPs Pn C FCi u E$ .A , Horizontal axis - percentage of neutrons below 46,5 keV The experiments carried out at BOC are shown in Table IV. in. assembly spectrum The compariso f o experimentn d calculationan s s i spresente d in Table V. The result CRn d fissioo Wsan n cross-sections ratior fo s Np-237 to Pu-239 (o-,, /a- , p /p ) of 64>C are close to simi- Set f absoluto a e discrepancies between calculationd an s r resultla f FCAo s . Ther e alsar e o contradiction r Pu-23fo s d an 8 experiments on these assemblies contradict each other (see Fig.I). Am-24 o Pu-23t 3 9 fission cross-sections ratios (cr^/a. ) ^ Two causes could clarif e discrepancyth y a e firs s Th i . t hydrogen inside FCA dioxide sample. The second is an uncertainty EXPERIMENTAL RESULT 5P-N O 5H-35SD 1AN 0 of self-shielding corrections for thicker 54C metallic sample. An illustrative information on this analysis is g-iven in A set of MA cross-sections ratios to U-235 fission cross- Tables I-III and Figures. sections was measured in the center of BP-1 plutonium critical as- TABLE III. ASSESSMENT OF EXPERIMENTS ON PU-240 TO PU-239 TABLE IV. EXPERIMENTS AT WC ASSEMBLIES NORMALIZED BY CRÏÏ RATIO R SOMSÏO E ASSEMBLIES U-23 PU-23D AN 5 9

A. FCA : Assembly E$C-54-4 B«C-56-IA K&C-56-IE KW-6 Kuolide

Assembly Experiment CnlculRtlons, HAE_go (C) C-E K 238 Pu a _ N1^A Nl«o •* -0052 b~ ' frr frr frr N1 «0 Nl^-.OO?' (E) JJ} 240 Pu w cr r fr frr, crw crw frr b"1 horn. het. het. het. 241 Pu frr,erw orw orw orw F CA-LX-I 0.253 -O.II2 -0.400 0.071 0.065 -0.188 7 23 242 Pu frr _ _ frr FCA-II-2 0.472 0.271 0.189 0.236 0.221 -0.251 HP îrr.crw w frrcr . crw frr.cj FCA-IX-7 0.307 0.338 0.331 0.346 0.333 0.026 241 3 24 Am frr crw crw - m i frr frr _ frr . BK-54-3 d BK3-45A-»n 4 = I 244 Cm frr frr - frr £L Corrections 1 w5» Self-shielding for îlltb" ) Mediu- mhe a) Reaction rateö /f rati J 6 o .00261 .00297 .00548 .0223 terogenei- (t-o) ty h) Central reactivity worth . rati D o

BSC-54-4: E 0.209* -0.023 -0.029 -0.046 — _ ^.ooe

C - -0.015 -0.016 -0.020 - -0.008 high enrichment zone, plutonium subassembly betwee- me n the d man HXM5A-I tallic uranium subasaembl enrichmenw lo n i y t zonee résultTh . a E 0.232* e presentear n Tabli d. VI e — — — — — ±0.010 Other sample f o Am-24s 1, Cm-244, Pu-230, Pu-24e ar 0

C - - - - -0.058 -0.015 irradiated at EH-350 now.

THE PROGRAMME OF EXPERIMENTS PLANNED FOR NEAR FUTURE

sembly (average neutron energ e experimenta Mev)1 yTh . l techniques f measuremento ¥e t plase na s directe o t validatiod f o n used glasses, activation detectors and layers. Some results of fissio d capturan n e cross-section t a A criticaM f o ls these experiment comparisod an s n with Monte-Carlo calculation. or s assemblies f smalo t lse A samples. , fission chamber d othean s r e basith f 5HA5-7o a 8 librar e giveyar n Tabli n. VI e detectors with MA will be used. A wide range of neutron spectra There were many measurement t a 5H-35s 0 aime o chect d k neut- froe hardesmth t (5P-1 o mort ) e softer than BH-35 s i 0planne o t d n cross-sectionro n i spectr w enrichmenA lo M f o af o s t zone, be arranged. TABLE V . COMPARISON OP EXPERIMENTS AND CALCULATIONS v c - t •7 i I I I il ro X f< f< K XX H H H M M M M 9I PI PI SI 5I I S C/E Ratio BK-54-4 RDC-6I t y - t t t Cole. - Exp. t£*» .w 0.05 /€ 0,67 0,62 ^ifo^MS- 1,02 1,01 &/***'/ <3fJîf' 1,00 0,98 2-MA , / ,- ,™ K/Ji" 0,98 0,95

-0.05 ^7^^ 1,02 1,01 6?~**'/&D:ar 1,20 - @*-'n/6'ta*f 0,69 0,65 -0.10 - <5ÇM"/6£"* 0,96 0,96

p /p 0,084 _ -0.15 - BSC• , KE= ( P Pnî

f „J^fwJ O- FCA ( ! * o) 0,126 0,074 -0.20 P*"'//'*"*

a,b) For the marked ratios the difference C-E is given i/ > -0.25 l/ l O O -0.1 0.0 3 0.0- 1 2 0. Yo)

Some experiments carrie t witou d h these chamberd an s P1g 2_Correlàt1on of Pu central reactivity worths with samples at E)/P*?'>)- calculated (RBP. mock-up of faat reactor with Pb coolant) assemblies are analysed. and measure BK( d . !) A next step of research la connected with active cores containin ga larg e quantit f MAyo t .I will permi o t validatt e capture and inelastic cross-sections. It will also give a possibility to found parameters of cores of reactors for burning MA in homogeneous and heterogeneous compositions. We plan to have at BJ>C 10-15 kg of Np-237 dioxide at the e yearth enf .o d This amoun s i expectet o increast d o 50-10t e . kg 0 TABLE VI . COMPARISON OP CALCULATED AND MEASURED AVERAGE CROSS-SECTIONS NORMALIZED BY u-?35

BP-I EH-350 Reaction Zone ol low enrich- zone or high enrtcn-i Wi-sudasem- | uepietea rate ment (ZLE) ment (ZHE) y (ZLbl E center) metal U.ZLE E C/E E C/E E C/E E C/E E C/E

240Pu(n.f) 0.827*0.026 0.91*003 0.200*0.010 1.02 0.242*0.012 1.04 241Pu(n,f) 1.29*0.04 1.02*0.3 1.11*0.05 1. 16 1.12*0.05 1. 15 242Pu(n,f) 0.66*0.02 0.98*0.03 237Np(n,f) 0.77*0.023 1.04*0.04 0.182-0.006 0.994 0.249*0.012 0.93 0.227* 0.978 0,190* 1. 01 ±O.OII *0.07

241Am(n,f) 0.825*0.025 1.00*0.04 ?37Np(n,tf) 0.240*0.012 1.08*0.07 0.73*0.04 0.95 0.68* 1.09 Î0.03 241Ani(n,$)" 3.51*0.28 1. 018 4.02± 0.92 *0.20

a) Normalized by 23q?u(n,jf)

REFERENCES

/!/ T.Mukayama et al. Actlnide Integral Measurementa on FCA for Evaluating and Improving Their Cross-Sections Data.-NEACRP-A- -684, 1985. X2/ C.U. BeflH«KOB M «p. FIpoBepKa HeKTpoHHbix flaHHbix paaa aKTM- HMÄOB B HHTerpajibHbix 3KcnepnijeHTax. - Bonpocu aroMHoa HayKM M TBX- HMKM, Cep.: HaepHue KOHCTaHru, a.4, c.71, 1991. A CONCEPT OF SPECIALIZED FAST REACTOR Here we shall confine ourselves to considering physics aspects FOR MINOR ACTINIDE BURNING of this problems. One of the main inpedirnents is associated with the drastic rise in a posiive value oF sodium yoid reactivity effect ( SVE ) compared with conventional cores. V I MATVEYEV, A.P. IVANOV, E M EFIMENKO Note, tha " Nucleat r Safety Guide n Reactoo s r Installatiof o n Institute of Physics and Power Engineering, Atomic Power Plants "(NS P AS-3GR , vali) 9 d withi e borderth n s the Russian Federation contain the requirement to ensure a Obninsk, Russian Federation negative value of SVE. Another problem is connected with an essential reductio e valuth f ß

INTRODUCTION

Use of fast reactors for minor actinide transmutation I. MODELS AND INITIAL DATA INVOLVED is likely to be an effective way of solving the general problem of long-lived nuclear waste activity reduction. The o e tw Th reactormodel f o s s were selecter fo d analysis of fast reactor capabilities for this purpose calculâtlonal investigations a :relativel y small-power reactor demonstrate e expediencsth consideo t y o trendstw r : (modular-type)—17 e a BN-?OreactoMW d 0 Oan r type reactor. -us f eseriao l fast power reactors whose fuel contains minor bote Th h reactors adopte e similath d x FUel subassernbles (face actimdes (3-47. weight) in the form of minor admixture; to face dimension 96mm) and fuel elements (d=6.9 rnm ) with the maximum heat loa q d =45" w/cm. -adoption of specialized cores ( or specialized Fast 1 ritax reactors), whose fuel contain e maximuth s m feasible amounf o t The fuel represents a zirconium matrix with minor actimdes minor actimdes. (neptunium, americiur d curiuan n m mixed with uranium—235) doped It is assumed that in the first case no further serious into it.For the BN-300 type reactor consideration is also given problems in fuel cycle organizaion or in reactor design are n talternativa o e compositio f fissilo n e rnatenal-a mixturf o e assumed to emerge, except the necessity of the spent Fuel deep minor actimdes with plutonium froWER-luOe th m O reactor spent purification from minor actimdes. The latter circumstance fuel. With the notion of "enrichment" inapplicable in the event represent e maith s n conditio f transmutatioo n n expediencn i y of the core of the type involved, £he critical parameters in fast reactors in general. In the first case the efficiency e characterizear densite th y b dg/crn"( y f o )Fissil e materials of minor actimdes transmutation is attained at the expense amount homogenized over the core volume. of fairly large amount of fast reactors in nuclear power t» The First reactor core represents a single-zone version, the estimates made their portion should be not less than Z07.. the second reactor is Furnished with 3 sub—zones of dissimilar The second case is peculiar in the ability of composition ( analogous to three enrichments ) to Flatten the neutronics properties of minor actinides—arnericium, neptunium, power density field.A sodium cavit e bot th s use i hyn i dreactor s curium-(in principle) to establish critical load on each oF the r additionafo l reductio n i nSVEs locatei ,t i d betweee th n above actinides including their mixtures n thiI . s case their core and the upper axial blanket-similar to /l/;the lateral and more complet d rapian e d burn-up coul e b attained d witn a h bottom axial blanlet e mad ar f steeo ed s sodiu an l m (the efficiency 5—lO-fold higher tha e firsn th bha n ti t case.Evident Tables I, 11 ) . The calculations were made in the traditional are the difficulties encounting in designing such a reacor. (dlffusional> technique e two-dimensionath n i s l (r,z) geometry. TAÎLE I.SPECIFICATION REACTOE TH F SO R Phi-300. TABLE II-SPECIFICATION REACTOE TH F SO R EN-170.

VERSION VER.l VER. 2 Type of fuel U+MA Compositiof o n Np/Am/Cm Type of fuel U+MA Pu+MA fresh fuel 59.26/37.41/3.33 Compositiof o n Np/Am/Cm Pu/Np/fim/Cm fresh fuel •39. 26/37. 41/3. 33° S3. 82/6. 67/4. 13/0.37' Fuel matrix density (g/cm ),accommodated 0.37 Fuel matrix density ovecore th re (9/cm > , accommodated 0.4S Volume fraction of fuel (7.) 30.00 over the core Numbe f cor1 o r e sub—zones Volume fraction of fuel ("/.) 33.23 Core height (cm) 100.00 Numbe corf o r e sub-nones 3 Core radius (cm) 57.OO Core height (cm) 100.00 Sodium cavity thicness(cm) 3O.OO Core sub-zone radius (cm) 32.GO/ 105.24/ 123.34 Compositions of lateral and Sodium cavity thicness(cm) 30. ' >0 bottom axial b lanl-et ( st. /Na) 0.36/0.64 Composition f laterao s d an l Thicness of lateral and bottom axial blanket (st. /Na> 0.36/O.64 bottom blanket (cm) 43.0/6O. Thicness of lateral and Number of burn-up compensators 6 bottom blant-et (crn) 100.0/ 100.O C B4 Burn—up compensator Number of burn-up compensators ir material (boron carbide 6O7. enrichment) Burn-up compensator B4C material (boron carbide 807. enrichment) Spent fue f WWER-10o l yea5 " r( 0 Decay)

Spent fuel of WWER-1OO ( 5 year Decay)

2. RESULTS OF CALCULATIONS The values of ß . , reactivity variation resulting from e calculationTh f o criticas l parameters were made proceeding from the requirement of the SVE zero value; for the fuel burn-up (withi 0 3 dayn , ) sefficienc f burn-uo y p BN—300-type reacto e additionath r e provisioth lr demanfo s n wa d compensators, fuel composition unde burns it r o — u0 p9 ove, 3O r of power density field flattenning ovee corth re radiusn I . 150 days were also calculated. The principal calculât lonal this case SVE was determined by the space incorporating the results are given in Tables III-VI and in fig. I. core, sodium cavity and upper axial blanket. In this case two e resultTh s presented testife feasibilitth E o t ySV f o y critical parameters were determined: homogenized densitf o y valu d acceptablan e e 3 valu( f o e eff ia specializen d core fissile material d weighan s t e totaratith lf o amoun f minoo t r of fast reactor witn a h adequately nigh amoun f minoo t r actinides to the general amount of fissile isotopes. actinide n fueli s - 45-50 "/.. TABLE III .CALCULATI ONAL VALUES OF MAIN CHARACTERISTICS TABLE IV.VARIAT ION OF FUEL COMPOSITION (KG) IN O) THE BN-17O REACTOR DURING BURN-UP TIME (DAYS)

TYP REACTOF O E BN-17 : R O : BN-3OO ISOTOPES 0.0 3O.O 90.0 150.0 r VE2 R. VE: l R.

am41O O . 1 0 2 193.00 192.00 136.00 critical density of :zonel 0.36 zarte: 0.3l 1 am2m 0. 15 0.51 1.13 1.3O fuel material 1.75 :zone2 1.10 :zone2 0.40 am43 70.20 69.50 63. 2O 66. 5O (g/cm core) :zone3 1.25 :zone3 O.45 cm42 0. 13 1.57 4.01 5.31 cm44 0 25 2. 22.70 23.00 22. 9O weight fraction of MA cm45 1.51 1.56 1.66 1.76

in fuel material (7C) 53.0 47.0 : 11.: nP37 430. OO 423. OO 412O .O 397.00 u235 65 1 . 00 64O.OO 617.0O 59 3O. O 0. 434 0.526 301 u233 73.30 73. 1O 72.70 72.30 u236 0.00 2,3O 6.35 11.36 efficienc f fueo y l pu3S 0.00 3.63 12.20 21. 12 burn-up compensators 2.3 7.7 PU39 O.OO O. 16 O.53 1. 13 fp35 O.OO 12.2O 36.50 60.40

continuous reactor TABLE V.VARIATION OF FUEL COMPOSITION (KG) IN operation (days) 16O 110 THE BN-30O REACTOR DURING BURN-UP TIME (DAYS) /FUEL:U+«A VER. I/

reactor reactivity : ISOTOPES O.O 30.0 9O.O 150.0 variatio n: -0.5durin e 5th g -0.7* -3.33 1-st mont operationf o h : am41 753.00 743.OO 711.00 673. OO am2m 0.57 2.54 6.15 9.33 am43 264 . 00 261.00 254. OO 243. OO Plutoniu o minot m r actimdes rati r thifo o s versios i n cm42 0.43 3.54 21.90 31.90 unchanged ,therefor s -7.i E 5eSV X(l-k)/k cm44 34.30 36. 2O 33.30 91. 10 cm45 5.69 6.00 6.60 7.22 np37 1620. OO 1590.00 1530. OO 147O.OO u235 246O.OO 2390. OO 2270.00 2150.00 u233 276.00 276.00 273.00 270.00 Pu33 0.00 20.30 63.00 116. OO PUÖ9 O.OO 0.39 3.44 7.02 Fp35 O.OO 63.20 133.00 314. OO TABLE VI.VARIATIO F FUEO N L COMPOSITION (KGN I ) 1000 THE BN-300 REACTOR DURING BURN-UP TIHE (DAYS) /FUEL:PU+MA VER.2/

ISOTOPES 0 . 0 30. 0 90.0 150.0

am41 60 . 90 •53. 3O 53 . OO 43. 10 am2ra 0. 50 0.50 1,23 1.74 am43 21.30 22.30 24. 4O 26.90 cm42 0 . 04 1.96 5. 14 7.36 6.31 7.39 3.7O 10.13 CFor Core) cm44 * -SVE 1 O.57 0.32 1. 12 (For Core+Sodium cm45 O.46 * -SVE 2 Cavity) -30.00 26.00 35. 9O 44.60 =SV3 E (For Core-*-Sodlum pu38 0 6 . 1 2 Cavity* pu39 93 1 . 00 92 1 . 00 303.00 690 0 3. Lateral Blanket pu40 42.3 . 00 434.00 430 0 3. 434 . 00 -40.00 PU41 212.00 205 . OO 193.00 131.OO 0 00 20 00 40 00 60.00 80 00 10000 Pu42 0 IO0 3. 104.0O 105.00 11O.OO Loading Rati f o Minoo r Actinides (SB) np37 0 130 O. 123.OO 110.0O 0 90 3. 62.30 133.00 316. 4O Fig 1 Relationship between loading ratio of minor fp39 0. 0 actmide d sodiuan s m void effect (SVE)

REFERENCE

I. The core concept of fast power reactor with sodium void reactivity close to zero. Matveyev V.l., Chebeskov A.N., Krivitsky I.Y. CONCLUSION The specialists meeting on passive and active safety e propertieth Fast f o reactorss . Orai-Mac , hi Japan, e probleTh f o specializem d core developmenr o < t november, 1991. specialised r reactofo efficien ) r t burnin f o minog r actmides directly involves the problem of required safety parameters ensurance, primarily SVE and ß . This problem caf uranium—233-free o nb solve e us d e witth h e fuel with inert diluent e discusseTh . d capabilities will evidentle b y essentially determined by the specific features of the external fuel cycle, primarily by the feasibility to Fabricate fuel containing a great amount of actimdes. This needs a special consideraion. In the BN-17O and EN-300 reactors over a year 150 and 450 kg of minor actmides ( Np, Am, Ctn ), respectively, are burnt. ± THE PROPOSED FUEL CYCLE OF THE ACTINIDE Engineering basi DOVITr fo s A program demonstration realizatiot na BURNING FAST REACTOR DOVITA EIAR: 00 (Summary) complex of facilities on pyroelectrochemical reprocessing and production of high activity nuclear fuel in the hot cells; A.V. BYCHKOV, A.A. MAYORISHIN, O.V. SKIBA automated facilit vibropar fo y c fuel elements production from high active fuel; V.I. Lenin Research Institute of Atomic Reactors, BOR-60 fast reactor; Dimitrovgrad, Russian Federation - complex of hot cells for postirradiation investigation; analytica d radiochemicalan l laboratories.

e obtaineTh d experienc e e appliewhicb n realization hca i d e th f no The existing conceptions and programmes on transmutation of program: long-lived minor actinide considet no o d r) s practicallCm (Np , ,Am e th y question of fuel cycle development of actinide burner fast reactors (ABFE) BOR-60, BN-35 BN-60d 0an 0 uranium-plutonium fued subassembllan y or accelerator. production; an experience of FBR irradiated uranium and uranium-plutonium fuel Our program of fuel cycle for ABFE is based on the following reprocessin y "drygb " methods; requirements: n experienca developmenf eo d operatioan t f fueno l elementd san assemblies automated-production facilities. wastU T8 e content doe exceet sno d 10~^; minimum wastes quantity is mainly in the solid form; The possible dates and stages for DOVITA program demonstration under minimum operations under fuel reprocessing; financing : l manufacturinal earrinf o t ou g g process e remotelstageth n i s y controlled hot cells facility. . 1 Productio compositionf no s containin minoe th g r actinides; o realizr T progra ABFe ou th Rf o en closeo m d fuel cycle (DOVITA) experimental fuel elements productio d theinan r irradiatioe th n i n some inconventional and familiar approaches are used; BOR-60 reactor - during 3 years. (D) Dry reprocessing and preparation technologies of fuel 2. Postirradiation investigation of fuel elements; development of minor composition, containing actinides will provide the minimum actinides contained fuel reprocessing; investigations on quantities of technological stages and wastes; optimization of the closed fuel cycle - during 3 years. (0) Oxide fuel application as the most studied will permit to use 3. Creatio demonstratiof no n compleBOE-6e th n 0xo minor actinides the familiar engineering decision R involvinFB r fo s g promising fuel transmutation on EIAR basis; preparation of the suggestion for element d stee san achievemenr fo l maximuf o t m burn-up; commercial facility desig durinn- years3 g . (V) Vibropacking technology of fuel element production will permit to introduce the different compositions containing minor actinides Considering the flexibility of pyrochemical technology of actinide into fuel and automate completely the remotely controlled production fuel reprocessing and vibropac fuel technology the program can be changed process; taking into account the requirement for physics of definite FEE type and (I) Integral disposition of fuel reprocessing and fuel elements base fuel whatever uranium-thorium fuel. refabrication facilities together with reactor will permit in solvin f ecologicao g l safety problems since such complex will need only suppl f actinido y x froemi m other reactors; (TA e whol)Th e comple approachef xo s will permi creato t e th e compact plan Transmutatior fo t Actinidesf no .

The proposed program is based on the research devleopments:

pyroelectrochemical production and reprocessing of fuel with high neptunium content; partial m containereprocessinC d an m dA compositiof o g n befora e repeat irradiation; Am and Cm pyroelectrochemical separation from the main fusion products involving REE; introduction of the different additions into vibropac fuel elements involving composition minoe th r actinidfo s e burn-up; high burn-up of vibropac fuel elements. REDUCTION OF MINOR ACTINIDES IN NUCLEAR WASTE 2. REACTORS AND SCHEMES VIA MULTIPLE RECYCLIN FASGN I T REACTORS The initial fuel irradiation is assumed to take place in a 900-MWe, U02-fuelled PWR, analogous or close to 34 reactors running in France and 5 in Belgium. The U235 fuel enrichment is A.F. RENARD . PILATES , FUENTA L . A , E o s tha , burnut% a f 7 abouo p3. t 43,000 MWd/tonne b n ca e Belgonucléaire, achieved ove operation ra n years4 tim f eo . Brussels, Belgium Whil totae fueR eth lPW l inventory amount 2 tonne7 o st heavf so y metal, the fate of one tonne of uranium initially loaded is J. JOURNET, G. VAMBENEPE, J. VERGNES followed in a multiple recycling scheme. Electricit Francee éd , Paris and Lyon, The reactor r userecyclinfo da 1450-MW s i g e fast reactor, similar to the first consistent design of the European Fast France Reactor (EFR), already considere sucr fo hd studie [2]n i ss . It average plutoniu wile on l d m an enrichmen , % 1 s clos2 i t o et Abstract associat twelfte eon R sucabovef r ho LW fa reactoha e s ,th a o t r as fuel quantitie e concernedsar othen I . r terms actinidee ,th s Seven successive re-irradiation a fasR typn EF ti s e e reactoth f o r togetherm C , Am , ,(eithe Np accordin , alonee u rPu P th r o o g,t have been explicitly represented, with realistic out-of-pile times. This case), produced from 4 years of operation of 12 PWRs can be cover a perios f tim o df abou o ee century on t . Minor Actinide e assumear s d recycle core on en di loa sucf do fasha t reacto. ) R r(F homogeneousle tb o y recycled, i.e ) .oxid mixePu e, d(U fuel wite e th Th h. + H.A u P . recyclinadvantage th f o ge stratego t reduce s th i ey Seven successiv minod an e u rrecycleP actinide th f so e (MA) flows radio-toxicit e actinidesth f o y . are explicitely represented, covering nearl 0 a perioy10 f o d years detailee .Th d time scheme givee sar Tabln ni beloweI .

Table I 1. INTRODUCTION Recycling time schemes The problem of long-ranging toxicity in nuclear waste due to Steps Duration (y ears) alpha-emitting actinides is now clearly identified. While bêta- and gamma-emitting fission products dominate the activity of Initia4 irradiatioR lPW n nuclear waste over some hundred years, beyond that time eth 3 Storage+cooling residual activity is nearly entirely due to the actinides. Reprocessing+fabrication 2 5 5. FR irradiatio l nn° The idea has thus emerged to recycle the quantities of generated Subsequent Storage+cooling 5 actinide n nucleai s r reactor woule on d s: recycl t onle no yth Reprocessing+fabrication 2 plutonium, as in the present MOX recycling programmes, but also FR irradiations n"2 to n*7 5.5 neptunium, americiu curiumd an so-callee m th , d minor actinides. Many recen exampletr fo studies ] ,[2 d hav,an lik e] consideree[1 d as well therma fass la t neutron reactor thao st t aim. As the fissile zones alone of the FR reduce progressively the Pu In the studies presented below, fast reactors are considered for quantities, because they have a negative breeding gain, use is recycling. Realistic scheme e in-pile th assumear sr d fo ean d u generateP blanke• e madth n f o i o par t restordetf o t e out-of-pile times d successivan , e recycling operatione ar s criticality at reloading ; the hypothesis of a FR just explicitely represented over a total period of about 100 years. self-sufficien thus i su P made n ti . Minor actinides are assumed to be mixed homogeneously with the In view to establish the benefits of FR recycling on waste usual mixed oxide (U,PuO2) same th e n stoechiometri,i c fractios na toxicity comparisoe ,th mads ni e with respec referenca o t e case at the preceding discharge. without recycling, corresponding to the once-through strategy : irradiationR aftePW e rth spenl al ,t fuel assemblie storee ar s d Reactor schemed san definee sar section i d . 2 n and constitute waste. The method of calculation is mentioned under section 3. The impact on the toxicity of nuclear waste from recycling, either A first recycling strategy consists of re-irradiating plutonium only, or plutonium and minor actinides, is compared and successively in FR the Pu alone, supposed to be recovered with discussed under sectio. n4 99. efficienc% 5 reprocessint ya g; mino r actinide rejectee sar d

The computer programme used for this multiple recycle scheme is PWR ORIGEN-2 [3]. It treats neutron flux irradiation periods as well FRn'l FRn'2 FRn*3 FRn'7 Sum as natural decay. It allows to calculate the evolution of of activity for a very large variety of actinides (130 isotopes) and Waste of fission products (850 isotopes). Among the actinides, one follow n particulai s isotopel al r f intereso s uraniumr fo t , 627/297 300/155 158/93 51/46 M 237 627 neptunium, plutonium, americium (10 isotopes) and curium (11 Np isotopes). The version of the programme used was released in 1990. 3 1.5 M OS 0.24 9

The cross-section library retaine r thermafo d l reactor 77 429/310 465/258 370/214 318/199 irradiation is that referred to as '50,000 MWd/t' ; the fast Am241 reactor library correspond e 'advancedth o t s ' one. Parallel calculations with more detailed methods have served to check the LJ> 2 LI LI 1.46 26 validity of the burnup results. The LWR calculations were backed by a comparison with results produced in an OECD expert group 169 168/229 228/243 241/230 [4], while the FR burnup calculations had been checked with Am 243 169/156 respec desigR EF o nt method n [2]si . 0.8 u L2 1.1S_ 0.8 8 The activities have been converted in relative toxicities, by dividing them by the values of Annual Limits of Intake (ALI), for 51/89 68/107 81/112 69/82 an ingestion, recommende workerInternationae e th th r y dsfo b l (Cm 244) 62 Committee of Radiological Protection (ICRP) in 1990 [5]. These limits, define Becquerelsn i d , takbiologicae eth l effece th f to 0.3 0.37 0.44 0.46 0.34 1 radiations into account. 2,8/16 16/23 23/27 24/22 s worti It h mentioning 199thae th t0 ICRP recommendatios nha Cm 245 2,8 witm C h, respecAm e th , o Np t modifie , valueI Pu AL r e sfo dth former 1986 recommendation : the ALI are now lower (thus more 0.01 OJ_ OJ 2J.

A potential toxicity value can be calculated in the same way for 4. RESULTS e quantitth f naturao y l uraniu (7.e mor ) whic6t uses hi o dt prepare one tonne of enriched uranium feeding the PWR. 4.1 Variation of M.A. quantities over 100 years of recycling- Dividin toxicite toxicit e wastge th e initiath th th y eb f f yo yo l H.A.d recycle an evolutio e casFoe th u ,th P rth f e o f eo M.Af no . uranium ore allows to know after how long a time waste will quantities 100-yeaovee th r r period explicitely consideres di become as harmless as the uranium ore is ; this is analogous to give n Tabli nr . tonnMasseU pe eII f e o eg showes ar n i d the e 'rismethoth kf do factor' . 1] [ uses ,a n di initially loade four fo dr important M.A. isotope Np237s: , Am241, Am243 and Cm245. One observes the following : Maximum potential hazards onlconsideree yar d attemphero n : e t is made to evaluate the possible release rates of fission - Np23 reduces 7i abouy db tfactoa after2 rinitiae eacth f ho l products and actinides out of waste repositories. FR irradiations ; afterwards the reduction factor becomes progressively smaller, leading to a minimum fractio f abouo n t It appears from this comparison that, as far as waste is initiae l/15tth f ho l amoun; t concerned, theradvantago n s ei o recyclt e u onleP y (excepf to the energy production). There is a large incentive to recycle Am241 is growing in the two first recycles, because of the thao s t , H.AadditiomassePu n i e . th s o aftent year0 10 rn sca natural deca Pu24f yo 1 durin out-of-pile th g e time the; s n comes be reduced, e followinwit th R caseh PW y b respece ,g th o t a slow, progressive decrease, which derives from the reduction factors : of its predecessor Pu241 ; Np23r fo 70 9 firss i Am24o t to growing3 thed an ,n slowly decreasin thi; g s 60 for Am241 comes froreductioe mth irradiatioy nb predecessos it f no r Pu242 Am24r fo 30 2 4 for Cm245. - Cm245 requires some 7 cycles to become under control ; this isotope gives a small, long-term contribution to the build-up of It shoul e stresseb d d tharecycline tth g programme cannoe tb Pu24 Am241d 1an . stopped after this period, otherwis reductioe eth n factors above would fall to about 14 for Np237 and 3 for Am241 ; there would be 3 case recycleo e Tabln giveth I s: r II erecycleu sfo .P d an u .P no reduction at all for the 2 last isotopes. M.A. recycle e quantitie th ,e sam th 4 isotopee f o s s which accumulat n wastei e afte year0 r10 recycling f so . Such a recycling programme needs to be pursued. othee Onth r hand M.A e obviousln th ,ca . concentratee yb d prior to recycling. Table HI 4.2 Long-term evolution of waste toxicity Quantitie f Minoso r Actinides Accumulatin Waste th n gi e In term toxicityf so firss i t ti , usefu noto e t leth whae ar t after one Century, actinide isotopes contributing mostly to the total toxicity. This (in g/t of Uranium initially loaded in PWR) is done in Fig.l for the waste from the initial PWR irradiation. The major contributors are, respectivel: y PWR PWR - Am241 100d betweean 00 year10 n ; s PWR +7 Recycles +7 Recycles - Pu23 Pu24d 9an 0 between 100 10,00d 0an 0 years; once-through inFR inFR - Pu239 around 100,000 years ; (Pu) (P M.A.u+ ) - Np237 and its successor Th229 beyond 100,000 years. One notes that the importance of Np237 is smaller than in earlier N£237 work, Thi . likessentiall] s si [1 e re-evaluatioe resulya th f to n PWR 629 629 3 of the ALI values for Np in the last ICRP recommendation. Now, FRItoT 169 4 successive re-evaluations in either sense let suggest that the results should be given a relatively large uncertainty. Transmutations 170 125 2 Total 800 || 922 || 9 Fig. 2 compares the time evolution of waste toxicity for the different cases considered : PWR alone, PWR plus 7 cycles in fast A 241 reactor, recycling Pu only, or M.A. in addition to Pu. ^ PWR 299 299 1.5 FRlloT 1340 11 The total toxicity contained in waste is their maximum potential Transmutations 1200 -112 14 hazard, remaining trapped in packaged waste, should anyone Total suddenly ingest them. The effect of possible migration is not | 1506 2 0 | | 1527 accounted for. 243 Fission products dominate firswaste th te toxicity. w Aftefe a r ^ PWR 169 169 0.85 572 hundred years their contribution is rapidly falling to a very low FR Ho 7 7.15 level, corresponding to the presence of Tc99 and 1129. From that Transmutations 1 -3 time on, the actinides dominate waste toxicity. Recycling M.A. in Total | _ 168 738 || 8 additio decreaseu P o nt s their toxicit thato ys , with respeco tt caseR th reductioe ePW ,th follows a s ni s: 24S Cf m PWR 2.8 2,8 0.014 - at 1000 years, by a factor 40, ro - at 10,000 years, by a factor 25, FRHo7 10 0.7 - at 100,000 years, by a factor 30. 4 Radtotoxicily ro - Radiotoxicily io" ro i (related to initial U ore) = (related to initial U ore)

-^_ Actinides: PWR. once-through \ "^V ^^^PWR + Pu in FR l to 7 \ \ \ /PWR + Pu and MA. inFI

- • /7^ \

I01 k, ^ \ v\\"s >> \\ -

10° \ : Initia Uore^ x - — _ o S^ ; FÎSSion produc4ts / \ \ \ : P once-th. WR r ough \ \ \\ .o'2

m I n i / i lin i i i nul ni i i l n nu l il l 10 s ,n- Mil1 1 1l n in i t i Mil1 1 1l 1 1 1 1 Illl Mil1 1 1l 1 1 1 (till 10 10 10 10 10 10 to' 10" Time (years) Time (years)

FI ToxicitG1 actmidee timth s yv r efo s containe wastn di e (PWR, once-through) FIG 2 Toxicity vs time for the actmides and fission products in waste These reduction factors are significant, although they do not The latter time rang stiln e ca e coverelb y humadb n memoryo ,s reac e factohth r 200, theoretically associated with residuef so tha surveillanca t e programme makes n contrassensi ; ee tth 0.5 %. former time ranges, much longer, escape from this scope. 4.3 First results in terms of acceptable toxicity levels (Simple criteria based on maximum potential toxicities are not meant here to be preferrable to more elaborate criteria covering toxicitiel al , O2 n relatee Fig. sar quantite thao th d t f to f yo possible geological transfers by migration up to the groundwater. uranium ore needed for the fabrication of the initial tonne of The poin s ratheti o deduct r e orientation d prioritiesan r sfo enriched uranium feeding the whole scheme. A first, simple further activities, based on parametric studies.) criterion would thus be to declare acceptable (i.e. requiring no surveillance anymore) nuclear waste when their toxicity b y ingestio coms nha e dow thio nt s level. 5. CONCLUSIONS Starting fro initian ma l fuel irradiatio e long-terth PWRa , n ni m Accordin thiso t g , fission product e acceptablsar e after about potential radio-toxicity of the spent fuel, which becomes waste 300 years. In contrast, actinides would need much longer times : once-througe inth h option bees ha n, compare toxicite th o dt f yo 300,000 years (PWR, once-through), nearly 100,000 yearu s(P residual waste from multiple recyclin a fasn gi t reactor. Either recycle) 20,00r ,o 0M.Ad . yearan u recycle)s(P . the plutonium alone is recycled, or the plutonium and the minor actinides (Np , Cm) ,Am l supposee al b recovere, o t dy b d 4.4 Toxicity versus energy production reprocessing with a 99.5 % yield. recyclinR F e Inth g schemes considered here ove ra century e ,th Seven successive re-irradiation fasa R tn s i EF reactoe th f ro total productio electricaf no l energ abou s time5 i y2. t s higher type have been explicitely represented, with realistic irradiatioR PW e thath n ni n alone appeart I . s thus justifieo dt out-of-pile times. This cover perioa s timf o df eabouo e ton scale down by such a factor 2.5 the curves giving the toxicity of century. Minor Actinides are assumed to be homogeneously the actinides with recycles on Fig. 2. recycled, i.e. mixed wit(U,Pue hth ) oxide fuel. 4.5 Sensitivity studies + M.A u e advantagP Th .e recyclinth f eo g strateg reduco t s yi e the radio-toxicity of the actinides by the following factors with Two types of variants have been considered. In the first one the once-througR PW respece th o t h cas: e out-of-pile times were changed secone th separatioe n dth .I one, n t reprocessina M.Ad . an yieldu P f sgo were varied. - at 1000 years, by a factor 40, 10,00t a - 0 years factoa y , b r25 out-of-pile Ifth e e timehalvedb n sca , this favourably affects - at 100,000 years, by a factor 30. the quantities of Am241 during the cycling operations themselves. For example ,e firs th reducethee o thire tr ar tw on y fo d y db In addition, relating the actinide toxicity to the energy refabrication campaigns othee th rn . O hand activite ,th Am24f yo 1 production scales factodowa y nb r (e.g. 2.5 e curvee th )th r sfo in waste afte r100r o 100 00 ,year20 s hardlsi y influenced. actinides in the recycle cases. Assuming 80 % yield for the M.A. (what is already a hard The actinide isotope whose contribution is dominant from 200 to challenge for Am) results in an increase of the actinide 1000 years is Am241. toxicity, given above for the 99.5 % assumption, by a factor 10 in the time range from 100 to 2,000 years. This stresses how Halvin out-of-pile th g e times would reduc e quantitieth e f so develoo important s i pt ti hig h purification separation methods Am241 effectively recycled, wha favourabls ti refabricatioo et n; Americiumr fo l d firsal an f .to M.A.foe , rth toxicite wastth e th ef yo itsel hardls fi y influenced. Improvin separatioe gth isotopeu P ne th yiel sr frodfo 99.5e % th m With respect to waste toxicity, recycling Pu + A.M. in fast assumed abov 99.9o et beneficials %i furthet i s a ,r reducee sth reactors wilsuccessa e lb , provided tharecovere tth y yielt da actinide toxicity by a factor 3 around 1000 years and 4 between separatio e broughb n nca t high enough isotopeu P wels r a ,lfo s 10,000 and 1000,000 years. as for M.A. isotopes and first of all for Americium : respective target f so 99.9 % (Pu d )99.5an % (Am) shoul e n db aimei t a d 4.6 Acceptable toxicity levels : recapitulation research work. When one combines the findings of par. 4.4 and 4.5, assuming in Under such conditions indeed, the actinide toxicity will be particular recovery yields of 99.9% for Pu and 99.5% for M.A., reduced to the level of toxicity which corresponds to the natural the specific actinide toxicit reduces yi recycly db e e th dow o t n uraniu initialle mor y used, after about 1,000 year contrasn i ; s t level of the natural uranium ore toxicity after, respectively, to the PWR, once-through strategy (for which 300,000 years are IV) 15,000 years (Pu recycle) or about 1,000 years (Pu + M.A. needed) and also to recycling Pu alone (15,000 years are still 00 recycle). required), this period of time makes sensible a surveillance programme for waste repositories. r-o* —Not—e f-v The present results and orientations should be checked with e regarlong-terth o t d m waste disposal assessments.

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Cecilie, L. Commission of the European Communities Rabotnov, H.C. Institute of Physics and Power Engineering 0 Loia l 20 Ru,e d e 249020, Bondarenkl . oSq B-1049 Brussels, Belgium Obninsk, Kaluga Region, Russian Federation Cocksy, C.A. General Electric Company Sasahara, A. Nuclear Engineering Department Nuclear Energy Komae Research Laboratory 6835 Via de Oro, M/C S-72 Central Research Institute of San Jose, CA 95153-5354, United States of America Electric Power Industry (CRIEPI) 11-1, Iwato Kita 2-Chome, Decressin, A. Commissio Europeae th f no n Communities Komae-shi, Tokyo - 201, Japan 0 Loia l 20 Ru,e ed B-1049 Brussels, Belgium Shmelev, A.N. Moscow Institute for Physics and Engineering 31, Kashirskoe Shosse Ivanov, A.P. Institut Physicf eo Powed san r Engineering 115409, Moscow, Russian Federation 249020, Bondarenkl . Sq o Obninsk, Kaluga Region, Russian Federation Sztark. ,H Framatome/Direction Novatome Service NVTPC Khalil, H. Argonne National Laboratory 10 rue Juliette Récamier Building 208, Argonne BP 3087 Illinois 60439, United State Americf o s a F-69398 Lyon Cedex 03, France

Koch, L. Commission of the European Communities Vasiliev, B.A. Experimental Machine Building Joint Research Centre Design Bureau Institute for Transuranium Elements 603603, Nizhny Novgorod - 74 Postfach 2340 Burnakovsky Proyezd, 15, Russian Federation D-7500 Karlsruhe, Germany Wakabayashi. ,T Oarai Engineering Center Kochetkov, L.A. Institute of Physics and Power Engineering Power Reactor & Nuclear Fuel 249020, Bondarenkl . oSq Development Corporation Obninsk, Kaluga Region, Russian Federation 4002 Narita, Oarai-machi Ibaraki-ken, 311-13, Japan Ledergerber, G. Paul Scherrer Institute CH-5232 Villigen, Switzerland Wehmann, U.K. Siemens AG Power Generation Group KWU Hatveyev, V.I. Institute of Physics and Power Engineering Friedrich-Ebert-Strasse 249020, Bondarenko Sq. l D-5060 Bergisch Gladbach 1, Germany Obninsk, Kaluga Region, Russian Federation Wiese, H.W. Institut fuer Neutronenphysik und Reaktortechnik Matveenko, I.P. Institut Physicf o e d Powesan r Engineering Kernforschungszentrum Karlsruhe 249020, Bondarenkl . Sq o Weberstrasse 5 Obninsk, Kaluga Region, Russian Federation Postfach 3640 0r1o D-7500 Karlsruh German, l e y Zcykunov, A.G. Institute of Physics and Power Engineering 249020, Bondarenko Sq. l QbninsK, Kaluga Region, Russian Federation

Arkhipov. ,V Scientific Secretary International Atomic Energy Agency Wagramerstrasse5 P.O. Box 100 A-1400 Vienna, Austria