IAEA-TECDOC-840

Unconventional options for plutonium disposition

Proceedings of a Technical Committee meeting held Obninsk,in Russian Federation, 7-11 November 1994

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UNCONVENTIONAL OPTIONS FOR PLUTONIUM DISPOSITION IAEA, VIENNA, 1995 IAEA-TECDOC-840 ISSN 1011-4289 IAEA© , 1995 Printed by the IAEA in Austria November 1995 FOREWORD

This publication summarizes discussions and presents selected papers from a Technical Committee meeting tha IAEe tth A convene Obninskn di , near Moscow, Russia, 7-11 November 1994 invitatioe ath t Ministre th Russia e f no th f yo n Federatio Atomin no c Energy whicd hostes an , hwa d by the Institute of Physics and Power Engineering. The meeting focused on the disposition of plutonium produced from the operation of nuclear power plants in areas related to the nuclear fuel cycle. Plutoniu mformes i l existin al n di g nuclear powe runconsumee plantth d san d part remaining in spent fuel is a generic by-product of nuclear power generation. Over the next 15 to 20 years, a significant amoun f plutoniuo t m wil producee b l nuclean di r power stations worldwide, addino gt amounts already in storage. Additionally, the world's plutonium stocks are being affected by decisions concerning the management and utilization of plutonium recovered from nuclear weapons which are being dismantled. In this context, national strategies are directed at reducing the stockpiles of separated plutonium worldwide furthen i d an ,r developing technologies capabl safelf e o secureld yan y using and handling plutonium. The purpose of the IAEA's Technical Committee meeting was to consider unconventional approache plutoniur sfo m disposition, both fro pointe mth e vief sth o f wo fuel e specificycl whola th s d ea ecan type f nucleao s r fuel being used aime Th .s wer obtaio et n technical description thesf so e approaches, engineering judgement thein so r technological statud san development, and reports on national experience in this field. The meeting's results and conclusions are providing valuable guidance for future activities in this subject area. EDITORIAL NOTE

In preparing this publication for press, staff of the IAEA have made up the pages from the original manuscripts as submitted by the authors. The views expressed do not necessarily reflect those governmentsofthe nominating ofthe Member nominating the States of or organizations. Throughout textthe names Memberof States retainedare theyas were when textthe was compiled. The use of particular designations of countries or territories does not imply any judgement by publisher,the legalthe IAEA, to statusthe as of such countries territories,or of their authoritiesand institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement recommendationor partthe IAEA. ofon the The authors responsibleare havingfor obtained necessarythe permission IAEAthe to for reproduce, translate materialuse or from sources already protected copyrights.by CONTENTS

Summary of the Technical Committee Meeting ...... 7

GENERAL OVERVIEW (Session 1)

Management of plutonium in Russia ...... 17 N.N. Egorov, V.M. Murogov, V.S. Kagramanian, M.F. Troyanov, V.M. Poplavski, V.I. Matveev An overvie activitiee th OECe f wth o f Dso nuclear energy agenc plutoniumn yo ...... 9 2 . ZaribasN. Waste managements aspect 2 fuel f O (Ths o ) s ,Pu ...... 5 3 . W.H. Hocking, ToyP. lor, L.H. Johnson, R.J. McEachem, SunderS. Rational Pu-transmutatio U ...... ^ r nfo 9 4 . A. Lecocq, K. Furukawa

GAS COOLED REACTORS AND THORIUM ASPECTS (Session 2)

Plutonium destruction with pebbl typd ebe e HTGRs usin burneu gP r balld san breeder balls ...... 63 K. Yamashita, K. Tokuhara, R. Shindo The plutonium consumption modular helium reactor (PC-MHR) ...... 69 Alberstein,D. Baxter,AM. W.A. Simon Advantages and limitations of thorium fuelled energy amplifiers ...... 81 Magill,J. C.O. Carrol, Gerontopoulos,P. Richter, GeelK. van J. Mixed plutonium-thoriu fasmn i tfuee breedinus l g reactor sEcologica- y wa l of accumulated plutonium burning...... 97 E.Ya. Smetanin, V.B. Pavlovien, G.N. Kazantsev, I.Ya. Ovchinnikov, B.Ya. Zilberman, L.V. Sytnik

CANDU AND LWR (Session 3)

Annihilatio plutoniuf no CANDmn i U reactors ...... 3 10 . D.A. Menelay, A.R. Dastur, R.A. Verroll Plutonium burnin thermaa gvi l fissio unconventian ni l matrices...... 5 11 . C. Lombardi, MazzolaA. Reactor physics characteristics of possible fuel materials for plutonium-incinerating LWRs . . 135 R. Chawla, U. Kasemeyer, J.M. Paratte Physical and technological aspects of cermet fuel. Application of Pu burning in WWER reactors ...... 149 V.M. Dekusar, A.G. Kalashnikov, E.N. Kapranova, A.D. Karpin, I.S. Kurina, V.V. Popov Fuel with carbon matri burninr xfo g plutoniu thermamn i l reactors ...... 5 15 . E.A. Ivanov, G.N. Kazantsev, I.Ya. Ovchinnikov, P.P. Raskach

MOLTEN SALT (Session 4)

Some ideas about hydrid system concepts ...... 165 P.A. Landeyro, A. Buccafurni, A. Santilli Rational Pu-disposition for ^U production by THORIMS-NES (thorium molten-salt nuclear energy synergetics) ...... 169 K. Furukawa, K. Mitachi, S.E. Chigrinov, Y. Koto, A. Lecocq, L. Berrin Erbay Neutronic examinatio plutonium-transmutation no smala y nb l molten-salt fission power station ...... 183 Mitachi,K. Furukawa,K. Yamana,Y. Suzuki,T. KatoY. possibilite Studth f yo f usinyo g molten salt r plutoniusfo m utilizatiod nan actiniae transmutation...... 197 V.S. Naymov, A.V. Bychkov, O.V. Skiba, P.T. Porodnov

FAST REACTORS (Session 5)

Studu burneP n yo r fast reactor cores without uranium...... 7 20 . M. Ishikawa, A. Shono, T. Wakabayashi Pu burning in fast reactor cores using unconventional fuel without ^U ...... 219 I.Y. Krivitski, G.G. Byburin, A.P. Ivanov, V.I. Matveev, E.V. Matveeva Mononitride mixed fuel for fast reactors...... 229 B.D. Ragozkin, N.M. Stepennova, Yu. E. Fedorov, M.G. Shishkov, O.N. Dubrovin, L.V. Arseenkov Plutonium disposal and burning in lead cooled fast reactor ...... 237 y.V. Orlov Physic fasf so t reacto converter- U ...... ^ o t u rP 7 24 . D.N. Ziabletsev, V.G. Ilynin, M.S. Kolesnikova, A.V. Magaev

WEAPONS GRADE PLUTONIUM (Session 6)

Plutonium rock-like fuel integrate (PROFITD dR& ) ...... 3 25 . T. Muromura, N. Nitani, H. Aide, H. Takano Combinin acceleraton gturbina s ga a ed modulaan r r helium reacto near fo r r total destruction of weapon grade plutonium ...... 263 A.M. Baxter, AlbersteinD. Physical and technical-economic aspects of weapon grade Pu utilization in HTGRs ...... 275 Yu.P. Sukharev, A.I. Kiryushin, N.G. Zuzavkov, N.G. Kodochigov, A.S. Kudryashov, N.N. Ponomarev-Stepnoi, N.E. Kukharkin, E.S. Glushkov, V.N. Grebennik

Lis Participantf o t s ...... 5 28 . SUMMAR TECHNICAE TH F YO L COMMITTEE MEETING

Introduction

recenn I t years IAEe th ,bee s Aha n workin colleco gt analysd an t e informatio overale th n no l developments and trends in spent fuel processing and plutonium utilization. In its 1993-1994 programme Nucleae th , r Fuel Cycl Wastd ean e Management Division, followin currene gth t world situation, has been giving more attention to the collection, analysis and exchange of information on both plutonium accumulation and its utilization.

Presentl limiteya d numbe countrief ro involvee sar plutoniun di m disposition issues, although subjece th gainins i t g greater attentio lighte th globaf o n i l development 1990se th n si . This meeting focused on unconventional concepts of plutonium disposition, both from the point of view of the fuel cycle and from the point of view of fuel type, since they have not been reviewed and discussed in depth elsewhere. These unconventional concepts include options whic evolutiobasen e a h ar n do f no technology (e.g. modified fuel designs, metallic fuel, advanced LWRs, reconsidered fast reactors, etc. requird )an e further developmen validatiod tan wels n a option s la concept w basee ne s ar n do s still to be further elaborated by appropriate R&D (e.g. PuO2 dispersion in non-fuel oxides, nitride fuels, combination cyclewith T ha , HTRs, etc.) maie Th .n emphasi colleco t s technicaswa a t l description of those concepts and to be informed on their status and the schedule for further development and potential implementation.

Plutonium is formed in all existing nuclear power plants and contributes to a large extent to the power generated by any nuclear fuel. The unconsumed part remaining in spent fuel is a generic by-produc f nucleao t r power generation generalls i t I . y designate s "civilda r "reactor-grade"o u "P wild referree an b l s "RPu"ia o dt n this document envisagee .b Disposition ca u d RP alon f no g three scenarios:

1) spene th leav n ti fuet long-terr ei fo l m storag finald ean disposal source Th . e ter f mspeno t fuel radiotoxicity is dominated during the first 30 to 70 years (depending on fuel type and burnup) by the fission products, thereafter by americium resulting from the decay of Pu, and itselfu P e .theth Thiy nb s scenari beinn oi g develope numbea n di countriesf ro ;

2) strip the spent fuel from most of its fission products and refabricate the resulting mix of U intu P od fuele synergistian Th . c fuel cycles involving light water (coole moderatedd dan ) reactors (LWRs) and pressurized heavy water reactors (CANDUs), being studied in Canada and the Republic of Korea, and the recycle based on pyroelectrochemical technology for liquid metal (cooled) fast reactor (LMFRs) mixed oxid , Pu)Oe(U 2 (MOX) fuel, already demonstrated on a pilot-scale in Russia, are typical examples;

) 3 separate uranium, plutoniu fissiod man n product reprocessiny sb . Thi Pu recycl d sd gan an eU scenari beins oi g industrially pursue numbea y db f countrieso r . Further developmentm sai at minimizatio improved nan d managemen dispositiod an t minoe th f no r actinides (MA)o t further reduc radiotoxicite eth nucleae th f yo r cycle wastes.

Whatever the scenario, with storage time, the radioactivity of the stored material increases, an technicas dit l valu fuea s ela decreases rate dependageinf u e o Th . type s RP lowes it i f gn t o i : so r fofrou burnuP w r mlo s coolepga d reactor (OCR) highesd fuefrou P an l m r fo thig h burnur p(o MOX necessarfuelf i R n . )LW ca Age cleaneu e yb dRP (whicdm froA s mhit must the disposede nb ) r diluteo freshey db . Pu r

In April 1993, an IAEA Advisory Group Meeting on Problems Concerning the Accumulation of Separated Pu evaluated inventories of separated RPu, stating that about 120 t would accumulate early in the next century. A more recent update by the IAEA projected 160 t, and various other evaluation samse havth en i eperiod t range0 20 . do Thit fro0 sm 11 inventor a y o t buildu e du s pi shortag f availableo fabricatioX eMO n capacity. National strategies beinw no ,g pursued, will result i nsubsequena t reductio stockpile th f no separatef eo worldwideu dRP d technologiean , s capablf eo safel secureld yan y usin handlind beinge an ar gu g P furthe r developed.

The decision to dismantle nuclear warheads and eliminate military Pu stockpiles will add to the RPu stockpile more than 100 t weapon-grade plutonium ("WPu"). It should be mentioned that WPu has a negligible ageing rate and could therefore be stored almost indefinitely without technical inconvenience. However, nontechnical considerations calr earlfo l y dispositio f WPuno t leas,a t sufficiently advanced to make further military usage unattractive. This could for instance be done without dela blendiny yb g WPu soos a , available,s na ageo t d RPu, thereby rejuvenatin lattere gth . Different locations and ownership of the two types of Pu make such simple and straightforward solutions inapplicable, except for restricted amounts of RPu and WPu.

Objectives of plutonium disposition

All Pu utilization options pursue multiple objectives. New concepts of plutonium disposition place followin e r moremphasio th e f eo on gn s o objectives:

eliminatio stockpilesu P e th f no ; maximum consumption of Pu (sometimes even without production of new Pu); reduction of radioactive wastes; reduction of personnel exposure throughout the reactor cycle; recover f energyo y resources.

When several objectives are to be considered, compromise solutions usually have to be adopted. As a result, several alternatives are presently being developed.

No single scenario can be universal. The selection is dictated by national policy, bilateral (or multilateral) agreements, availability and/or accessibility of the technology, specific sensitivities of public opinion, etctherefors i t I . e recognized that different scenarios wil selectee b l considerer do d varioue inth s countries.

In assessing options and developing scenarios for Pu disposition, several criteria must be considered, some of technical nature, some based on political arguments. They include:

existenc potentiar eo l developmen technologe th f to y fue: l processing (conversion, fabrication, reprocessing and/or waste vitrification e materialth f o ) , fuel performance, reactod an r auxiliary facilities, transportation, etc.;

licensing consideration, including the required database;

timeliness, taking into accoun onlt annuae tno y th l capacit dispositiou P f yo t als timne bu o th e schedule for development, demonstration and industrial implementation;

safeguards, proliferation resistance and f addresse,i scenarioe th y db , irreversibilite th f yo demilitarization;

energy resource utilization;

radiotoxicit hige producte th hf th o levef yo d lsan waste arising fro scenarioe mth ;

amount of radioactive waste from the fuel cycle and reactor operation and from dismantling of those nuclear installations;

environmental impact; economics, includin developmente gth , demonstratio first-of-a-kind nan d costs;

public acceptance or acceptability.

Technical options

The alternative options to MOX fuel in LWRs, LMFRs and advanced thermal reactors (ATRs) are based on Pu in inert matrixes or on fuel type incorporating Th. The first group usually necessitates the simultaneous use of burnable or removable absorbers, which deteriorate the neutron economy. The utilization of Th results in the production of ^U and thereby conserves the energy resources to a variable degree, depending on the nuclear characteristics of the option. The danger of creating thereby a new weapon-grade material can be intrinsically avoided by tuning the generation

oUf towards significant contents Uof, whose daughter products are high energy gamma

232 emitters 233 . While this contributes to proliferation-resistance, it also leads to inconveniences in the subsequent utilization of this fissile material: heavy investment in reprocessing and fabrication plant development and construction and increased exposure level of the fuel cycle and reactor personnel. Some options consider the flexibility to utilize either type of fuel and even to return to conventional MOX utilization. Amongs fuee th l l optionsal t onee th , s base oxiden do probable sar y applicabln eo the shortest time-scale: foreseeing 10 years development and data acquisition (including a representative performance data-base), followed by 10 years for establishing an industrial fabrication capacit t unreasonableno s yi .

Whil time-scale eth fuer efo l cycle accommodation relatee th d dsan investment importane sar t to take into consideration, the time-scale and investments necessary to develop new reactor systems are potentially many times larger. But simultaneously, it is the reactor system that plays a predominan annuate rol th transmutatio u n eP i dispositio lu e P rat th f eo n i d nan level. Depending on wher emphasie eth placeds si larga , e variet optionf consideredye o b n sca . Various aspecte th f so LWRf o e , us CANDU, LMFR, high temperatur cooles ega d reactor (HTGR molted )an n salt (cooled and fueled) reactor (MSR) NPPs were described at the meeting, as well as concepts based on accelerator-driven subcritical reactors e utilizeb o t , d eithe a basi s a rc installatio a "pos r no - t combustion" installation.

Conclusion recommendationd san s

severae Th l objectives being pursuemultiple th d dan e criteria being considere difficule dar t to reconcile in any single option proposed or considered for Pu disposition.

Conventional options (MOX fuel in LWRs, LMFRs and ATRs) already applied industrially able copar o et e wit dispositioe hth separatef no stockpilesu dP , provided sufficien fabricatioX MO t n capacit deployeds yi . othee Somth f re o objective criterit r onlmetd o sno an , y e partiallaar y met.

unconventionae Th l options being suggeste developer do addressint a m dai g objectives and/or presen e utilizatioX th y b MO t t criterinme routinest ano turnn I . , non f thoseo e unconventional options meet all objectives and criteria. Certain systems place emphasis on flexibility to switch the option(s) to meet changing emphasis with time on particular objectives or criteria. Even so, the technical experts evaluating or developing unconventional options feel uneasy about the absence of general consensus on the ranking of the objectives and criteria. It is a concern worthy of being addressed, since some of those options required considerable time and budget for development, demonstratio industriad nan l implementation l thesal r e Fo reasons. effectivenesse th , , cost-benefit balanc timelinesd ean presentee mosf th so f o t d option difficule sar evaluateo t t .

A rather long list of proposals and recommendations were discussed at the working groups and plenary session, and the main ones are presented below: 1. It is important to evaluate unconventional alternatives being considered for Pu management. This meeting dealt with many technical aspects of various potential alternatives, enabling each of perspectiven i t pu thee b mo t . Regular review thif o s s fiel f developmendo t were considered highly valuable at the international level.

2. While all the presentations provided a good insight to technical aspects, most of the papers providdit dno ecleaa r pictur time-scalee th f eo considerecoste d b san o sdevelopment e th r dfo t stage, demonstratioe th ne industria stageth d an , l implementation stage (afte e evaluatiodu r f lessonno s learned fro deme mth o period). These aspects addresseneee b o futurede t th n di .

. 3 Future working group f thio s s type should additionally discusspecifiw fe a s c topics, including, for example, the use of Th in conjunction with Pu disposition, selection of the inert diluent and radiotoxicity issues.

4. Quantitative criteria should be developed to assess radiotoxicity and proliferation-resistance requirements.

. 5 International collaboratio recommendes ni near-terme th n di r developmenfo , materiaa f o t l technology databas r candidatefo e inert-matrices pursuo t d an suitablea , y formulated, numerical benchmark exercise for neutronics calculations relevant to particular reactor systems.

6. Further work should be conducted to prove the acceptability of inert carrier materials for fast reactor service. Experimental investigation necessare sar verifo yt characteristice yth conceptuaf so l core designs with different compositions in-corn o addition D I . e performancR& no t e characteristics, recommendes i t i d that wor continuee kb ensuro dt technicae eth l ability somt a , efuturee timth n ei , to reprocess used inert-carrier fuel.

. 7 Considerin inadequate gth e experimental data bas accelerator-driven eo n componentse th , performance testing and development of targets should proceed to some confirmation stage.

The meeting featured six sessions, which are summarized below:

Sessio . n1 General Overview, chaire . MurogoV . Bairioy dH b d van t

All presentations pointed towards the accumulation of separated Pu from the reprocessing of thermal reacto smallea o rt fueld r an ,extent , fast reactor fueldecisioe Th . dismantlo nt nucleae eth r warhead separateo t sequivalenn a wild u ad l dRP t quantit f separateo y d weapoe n Th grad . Pu e concerns abou tdecas radiotoxicitit f yo itsel u productd P desirabilite e an f th th d f yo san preveno yt t misuse of weapon grade Pu have promoted initiatives for considering several options for Pu disposition fuer LWRX fo lLMFRd MO an s.s Whila s developei se us eindustrian a n do l scale, alternative options are being considered in order to decrease separated Pu accumulation or to reduce Pu production even more.

Exploratio e lonth gf no time s alternativy needean r fo d e option strongly suggests that, whatever long-term path is taken, safe, secure and verifiable storage of separated plutonium is essential. The choice of Pu management is not only related to technological, safety and economic consideration alss i t o sinfluencebu non-proliferationy db , national policy publid an , c acceptability implications resulta s A . , international organizations have reviewe e currentlar d dan y reviewing several related aspects IAEe Th . Amaintainins i updatind gan gdatabasa forecasd ean civiliaf o t u nP generation, separation and utilization rates, as well as fuel performance updates. In this regard, the IAEA is in the process of publishing topical Pu-related safety standards. The OECD/NEA has been involved traditionall reviewinn yi economice gth logisticd fuesan e th l f cycleso generaln s i , ha d an , pursued and is pursuing activities in these fields on MOX fuels. Currently, an expert group is reviewin evaluatind gan broae gth d technical questions concernin managementu gP anothed an , r group is reviewing the physics of Pu recycling.

10 exampln a constrainte s th A f eo optiond san s typica eacr lfo h individual country Russiae th , n managemenu P t perspectives were presented availabilite th : LMFRf yo theid san r good performance, on the one side, and, on the other side, the quite large development program to be considered for MOX utilizatio WWERn ni s favour firss a , t priority e implementatioth , industrian a f no l capacity of LMFR MOX fabrication and the construction of additional LMFR plants, to utilize the Pu in the most expeditions time-scale.

e utilization conjunctioi Th h T f o n n with burnins beeha nu P gconsidere y severab d l organizations, sinc excese e th evolutios it scontrollee d U brereactivitb ^ an n e du nca th P y f db yo from Th without producing further Pu. The use in CANDUs of Pu fuel with either ThO2 or inert carrier matrice undes si r active investigation presentatioe On . n focuse assessmenn do f (Th-Pu)Oo t

fuel fro perspective mth f loneo g term waste disposa comparisoy b l n wit e existinhth g repository 2 design for used UO2 fuel: the only issues requiring further investigation are segregation during fabricatio build-ud nan excesf po s oxygen during irradiation THORIMS-NEw ne e Th . S stratega s yi candidate for a Th-Pu cycle maintaining the world energy inventory and effectively consuming Pu.

Cooles Ga Sessiod Reactor . n2 Thoriud san m Aspects, chaire . ChawlR y db a an . KudriavtsedE v

foue Th r paper thin si s session (two HTGR-relate eace on h d concernindan g fast reactord san

the energy amplifier concept, respectively) considered achieving high levels of Pu elimination by discontinuing the use Uof.238 There was general consensus that, for both reactor-grade and weapon- grade plutonium higa , h burnup capability need aimee b o srelativt t d a conventionao et l methodse Th . feeling was expressed that effective plutonium disposition will take several decades to complete and that, as such, certain novel (unconventional) methods would not necessarily extend the overall disposition tune-frame.

Three of the four papers considered the control of excess reactivity for the fresh core through e replacementh thoriumy b s recognize a8 f wa U o tt I . d thalatter'e th t s e contexusagth n f ei o t plutonium reduction would rais technicaw ene l issues with regar non-proliferatioo dt spent-fued nan l radiotoxicity. Differences in viewpoints were linked, on the one hand, to the desirability of generating ^U as a valuable thermal reactor fuel and, on the other hand, to the perceived replacement of one typ fissilf eo e inventor anothery yb . Proliferation concerns associated with U were expectee b o dt mitigate easies it y drb detectabilit high-energe yth (duo t e y radiation frome th 232 y U)wels b a ,s a l 233 denaturing possibility with U. Regarding long term radiotoxicity, one of the points made was that the Th/233!! cycle, considere 238once-througn di h operation t necessarilno s wa , y advantageous relative ^U/Pe toth u cycle.

Sessio . n3 CAND LWRd Uan , . chaireZabudk L . Pag R d y ean db o

With LWRs likel o continuyt dominato et e nucleath e r energy scene s importani t i , o t t consider their potentia reducinr fo l g plutonium inventories worldwide currene LWRn i Th . e f us tso MOX e corefuelth f ,so whilconstitutin% 40 e effectivelo t 0 3 g y decreasin e separateth g u P d inventory, does not, however, represent a net reduction of plutonium inventories as approximately same th e amounplutoniuw ne f o t mgenerates i thas da t consumed.

Design coreX beine r 100ar sMO s fo % g developed effectivd an , reductiou P e n shoule db achievable with them build-ue plutoniuw Th . ne f po m fro presenU m^ MOXn i t , however, remains disadvantageous in the perspective of eliminating further Pu production and consequently designs of plutonium burning fuels are being considered without uranium. The fuel would consist of plutonium

dispersed within an inert matrix, e.g. ZrO, A1O, MgO, zirconium alloys, or graphite. The use of 2 3

suc hfuea waten i l r reactors would overaln alloa r wfo l total plutoniu2 m reductio 60-70f no % with a fissile ^Pu reduction of the order of 90 - 95 % from a single reactor cycle. The absence of uranium (or Th) results in core physics effects which need to be investigated in detail in order to ensure that controllability and safety are not significantly affected. Scoping investigations indicate that the use of

11 appropriate burnable poisons, possibly combined with advanced fuel concepts developed in other contexts (suc s cermeha t fuel containing small amount fertilf so e material), could provid esuitabla e route towards implementatio LWRn ni f current-daso y design.

The CANDU reactor system, with its high neutron economy and well-thermalized neutron spectrum, reduces ^Pu formation by increasing the fission-to-capture cross-section ratio of 239Pu and eliminates the need for reactivity suppression through rapid on-line refueling. Using inert matrix fuels initiae th absence f th ano l burfissiln dn i % 238 f ca eo n U90 t e i , inventor once-througa n yi h mode cycl achievd ean e further reductions throug hmultipla e limitea shuffling s ha absencde U Th ^ . f eo dynamice effecth addressee n b o thu td n san scurrenf ca o de witus t e reactohth r technology. Certain aspects such as the refueling schedule require further clarification.

The development of LWR and CANDU inert-matrix, Pu-incinerating fuels is still at a relatively early stage, with work on candidate matrix materials having just started. It will be necessary implemeno t majoa t r fuel desig testind nan g programm considerablf eo e duration befor fuee eth l will e availablb r full-scalfo e e application. Ther s somwa e e indication that possibly unrecognized technological difficultie exisy ma st with candidate somth f eo e materials being considered. Valuable test data of relevance could already be available in individual laboratories, and it is highly desirable that this informatio poolee nb d before majo effortw undertakenne e r ar s .

Ther s somewa e differenc f opinioo e n regardin virtue gth f once-througeo h schemes. Inert matrices not amenable to reprocessing (intended to provide a spent-fuel form for direct disposal as waste) would, on the one hand, offer greater proliferation resistance but, on the other, make further reductio radiotoxicitf no y (e.g recycla .vi fasn ei t reactors) more difficult.

Session 4. Molten Salt, chaired by D. Alberstein and G. Pshakin

All papers presented in this session addressed approaches to Pu elimination that involved use of molten salt systems - some in conjunction with an accelerator-driven component, others without. All systems presented f non-aqueous o mak e eus , molten salt processin maximizo gt e plutonium reduction and minimize waste volume production. In all but one of the papers use of thorium for productio fissilw ne ef n o materia s described wa remaininle th n I . g paper focu e s onlth , n wa syo plutonium consumption, without productio fissilw ne ef n o materials .

The design and overall Pu elimination and B3U production capability of an accelerator-driven molten salt breeder reactor and a molten salt fission power station were presented to provide more explanatio THORIMS-NEe th f no S concept presente Session di thi n I s . systen1 m rather than simply burning out excess plutonium inventories, the plutonium can also be used to produce new fissile ^U. The applicability of THORIMS-NES in a scenario for plutonium elimination and initiation of a larga thoriu n o e a scal mevaluateds er ewa required an , d installed capacitie f eacso h reactor were quantified.

With regard to the accelerator-driven component of these systems, evaluations to date have addressed feasibility t thee supportebu ,yar minimay db l experimental data. Proposal r targefo s t performance testing have been prepared but not yet funded.

With regar molteo dt n salt processing systems, previous experience with these systems sha been accumulate Ridgk d Oa fro ee mMolteth n Salt Reactor developmental activities. Systems under consideration today differ in some details from the Oak Ridge system. The system evaluated and presented by the RIAR was derived from thirty years of investigations of fast breeder reactor MOX fuel reprocessin chlorida gvi e molten salts t indicateI . possibilitsa applyinf yo g pyroelectrochemical and pyrochemical techniques. The scientific bases of this approach have been formulated, and the initial aspect closee th f dso fuel cycle have been developed indicates wa t I . d that, afteadaptes i t i r d molteo t n fluoride salt processes, this proces includee b overale n th sca n di l process systee th r mfo molten salt reactor. It was also agreed that molten salt processing systems should proceed to the

12 experimental confirmation stage. A performance parameter that was particularly subject to discussion was the level of Pu contamination in the processing waste stream.

The molten salt systems with thorium fertile material can effectively reduce excess plutonium inventorie t alssbu o offe opportunitn a r redireco yt focue th tnucleaf o s r energy programs froe mth uranium/plutonium cycl closea o et d thorium/uranium cycle. Effort beginnine ar s organizo gt d ean increase worldwide cooperation in the evaluation and development of molten salt reactor technology. To the extent that fluoride systems are being considered for use in molten salt reactor systems, attention is being focussed on properties of fluoride molten salts such as actinide electrochemical propertie behavioud san f rarro e earth element thesn si e melts.

Session 5. Fast Reactors, chaired by D. Menelay and V. Kagramanjan

Several countries have developed LMFR technologies, and about 20 liquid metal cooled reactors have accumulated more than 200 reactor-years of operating experience over the past 25 years. The special characteristics of these reactors can be utilized to enhance the capability for plutonium consumption (includinfissilw lo f e o contentu gP meeo wels t a ) s ta l other future objectives.

addition I plutoniuo nt m disposition bees ha nt i ,suggeste d that long-term disposa wastef o l s fro l typemal f reactor so improvee b n sca consuminy db g minor actinide addition si plutoniumo nt . This sessio devotes nwa examinatioo dt technologiee th f no s neede accompliso dt h these tasks using fast reactors. Most LMFR designs utilize sodium as coolant, and this is taken here as the reference concep r fasfo tt reactors e alternatTh . e coolant propose s leaddwa ; fro e poinm th f vie o t f wo plutonium disposition this concep s similai reference t th o t r e concepwilo s e considere b ld an t d equivalent.

Plutoniu efficientle b n m ca effectiveld an y y dispositione fasn di t reactors, firs removiny b t g "blankete th " uranium. This capabilit alreads yha y been demonstrated. Plutonium consumption nca furthee b r coreincreaseX exampler fo , MO a decreasinn y di b , fuee gth l volume fraction, increasing core leakage, or by adding neutron absorbing materials.

In addition to the above, it may be possible to remove the plutonium-producing isotope from fuel in the core region so that the net plutonium consumption increases significantly. Three papers in this session proposed this approach. Such a core can achieve a very high rate of plutonium consumption with vercoolanw ylo t void reactivity. Some design issues remai resolvede b o nt n i ; particular, the large reactivity loss with burnup and the small Doppler effect. Alternative to the use of inert matrix carriers authoe on , r proposed replacemen cor e witU th e ^ f o ht ^Th numbe.A f o r challenging problems remain to be solved for this option.

Options seeking more complete annihilatio plutoniuf no minod man r actinide fasn si t reactors require reprocessin eacf o hd spene operatinen th f e tgo th fue t a l g cycle desige Th . n characteristics of these reactors are changed by core reconfiguration, and careful examination is required to ensure performance th systet compromisedsafete d th no e an f e ymo ar . Scoping studies have given sufficient confidence that these goals can be achieved.

The search for suitable "carrier" materials for plutonium fuel was a central theme of the session considerablA . e amoun f wor o ts bee kha n don inern eo t matrix materials, especialln yi Russia. Preliminary irradiation test results were reported, mainly for mono-carbide and mono-nitride forms as well as for porous zirconium carbide saturated with fuel oxide. Plans are underway in France for irradiation tests of solid solutions, cermets, and cercers (ceramic-ceramic). Final selection will require several year f workso , give mane nth y variable investigatee b o st lone th gd timedan s require r irradiatiofo d d examinationan n t appearI . s that opportunities exis r internationafo t l collaboratio somn ni e aspect f thiso s work.

13 Session 6. Weapons Grade Plutonium, chaired by N. Zaribas

All three papers presented at this session addressed destruction of excess WPu. Two papers focusse hign o d h temperatur s coolega e d papee reactors on lighn o rd tan , water reactorsl Al . approaches aime t finada l disposa spenf o l t fuel wit furtheo hn r reprocessing. Issues with regaro dt radiotoxicit f spenyo t fuel wer t addressethesf eo no y e an papers n di .

In the LWR presentation, preliminary inert matrix fuel development activities were described, and a schedule for further development was presented. A number of technical issues regarding fuel behaviour under both norma accidend an l t conditions (both design basi beyond an s d design basis accidents) were identifie s needina d g further evaluation. These include fuel matrix/coolant interactions, fuel matrix/cladding interactions, general irradiation behaviou othersd an r .

Bot cooles hga d reactor presentations indicated that this reacto capabls ri f achievineo g high levels of excess weapons grade plutonium reduction. There are some data from irradiation of plutonium coated fuel particle HTGRn i s s that confir capabilite m th fuee achievo th t l f o y e high burnup.The presentatio combininn no cooles ga e dgth reactor wit acceleraton ha r acknowledged that this approach would require more development work tha muce nth h more well developed approach of usin cooles g ga onl e dyth reacto plutoniur fo r m destruction.

Acknowledgements:

The success of this meeting was in part due to the able leadership provided by H.Bairiot of Belgium.

ABBREVIATIONS

AGM advisory group meeting ATR advanced thermal reactor (Japan) CANDU Canada deuterium-uranium (reactor) OCR cooles ga d reactor HTGR high temperature gas cooled reactor IPPE Institut Physicf eo Powed san r Engineering (Obninsk) LMFR liquid metal (cooled) fast reactor LWR light water (cooled and moderated) reactor MA minor actinides (Am, Cm, Np, etc.) MOX mixed oxide (U, Pu)O2 MSR molten salt (cooled and fueled) reactor NPP nuclear power plant RIAR Research Institute of Atomic Reactors (Dimitrovgrad) RPu reactor-gradu eP THORIMS-NES thorium molten salt nuclear energy synergetics (based on molten salt converter, accelerator-driven molten salt breeder and dry processing plants) WWER water moderated, water cooled reactor (Russia) WPu weapon-gradu eP

14 Next page(s) Left blank SESSION 1: GENERAL OVERVIEW MANAGEMEN PLUTONIUF TO RUSSIMN I A

N.N. EGOROV Minatom, Moscow

V.M. MUROGOV, V.S. KAGRAMANIAN, M.F. TROYANOV, V.M. POPLAVSKI, V.I. MATVEEV State Scientific Center, Institute of Physics and Power Engineering, Obninsk

Russian Federation

Abstract

The Russian concept of plutonium management (both civil and weapons) is postulate baseth outee n dth o f reo fuel cycle closure, necessit o enhanct y e fuel efficiency, and decreasing radioactivity of disposed long-lived wastes. The key proble l countrieal r mfo s considering recyclin goptimizo t optionw no es si for the short-, medium- and long-term. Short-term plutonium management in Russia is based on safe and reliable storage of separated civil plutonium and ex-weapons plutoniu usee b m reactorsn n di untica t i l . Usin f accumulatego d plutoniu mfasn i t reactor questioa s i s f medium-terno m plutonium disposition options. Long-term options to be defined consider use of plutonium in VVER- type reactorburnine th d f plutoniugo san m exces minod san r actinide fasn si t reactors witcoresw hne .

t. INTRODUCTION

The increase in the accumulation of plutonium stocks in the process of today's reactor operations, as well as the plutonium released as a result of armament reduction evokes steadily raising concern of Russian and foreign societies abou futures it t . Expected quantitie f civiso l plutonium accumulaten di the spent fuel of now functioning Russian NPPs with a life time of 30 years are given in the Table I. Also in this table are the quantities of ex-weapons plutonium expected to be released in Russia.

The problem with plutonium is especially acute due to the public desire to mak disarmamene eth t process irreversible .o opposin Thertw e ear g pointf so vie plutoniumn wo viee regaro t ws On i . wastt da i onl s ya e that shoule db disposed in deep geological formations. The other view insists on the need of plutonium recycling. Plutonium recycling increases national energy resources d decreasean e radioactivitth s f long-liveo y d nuclear wastes. Fror mou standpoint this divergence in views extends from the difference in the status of fuel cycle technology and the comprehension of the role of nuclear power in particular countries.

country'a If se futurvieth f f nucleaweo o r powe s optimistii r d fuean cl cycle technology relate plutoniuo dt m recyclin sufficientls gi y develope: o dt

-follow non-proliferation policy;

17 -provid safeecologicalla d ean y acceptable plutonium based fuel cycle;

-be economically competitive,

then plutonium is considered a national asset. Otherwise, it is a permanent global threat. The key problem for all countries considering recycling options is now to optimize for the short-, medium- and long-term.

Ine considerin ar Russi e w a e recyclinth g g option along with these additional considerations:

-growing plutonium inventories because of:

- fast reactor development delay and

- arms reduction;

-good experience outsid f Russieo a using plutoniu mtherman i l reactors;

-increased public anxiet long-liven yo d radiotoxic nuclear waste;

-change Russiae th n si n economi politicad can l environment;

-great uncertainty in the future of nuclear power development;

TABLE I. AMOUNT OF PLUTONIUM IN IRRADIATED NUCLEAR FUEL OF MODERN RUSSIAN REACTORS AFTER 30-YEARS IN SERVICE

Reactor type Power, Pu1, t Pufis2, t GWt(e)

VVER-440 3 27 (3,7) 20 (16)

WER- 1000 7 55(9,1) 40 (33)

RBMK-1000 11 76 (8,3) 45 (37)

BN-600 0,6 12(0) 11 (11)

Weaponu sP - 10) 0(0 95 (95)

In total: 270 (21) 211 (193)

bracketn i 1 s mas f Pu-241+Am-24so 1 bracketn i 2 s countin Pu-24n go 1 decay after 30-years storage

18 -readines f o politicals , scientifi d industriaan c l structure f o westers n countrie Japad san helo nt p address plutonium utilizatio Russian i .

e supposW e long-ter th tha n i t m perspective, optimum plutonium recyclin havn gca e many identical feature Russin si othed aan r countriesr fo s A . the short- and medium-term, we think that Russian plutonium utilization plans could diffe o somt r e extent from thos f otheo e f thio r scountriesm ai e Th . presentation is to clarify this point.

2. EXPERIENCE GAINED WITH PLUTONIUM

Russia began using plutoniu nucleaa s ma secone r th fue n i l d hal f '50so f . In 1957 a core was fabricated of metallic plutonium alloy for the pulsing fast reactor IBR-30. In 1959-65 plutonium dioxide was made for the BR-5 and IBR-2 reactors.

Systematic studies of plutonium-bearing fuel began in 1970 in the BOR-60 reactor. This work were conducted in the framework of the fast breeder reactor development program and was aimed at the most effective use of plutonium to expand nuclear power fuel resources.

Until now coreo ,tw f weapono s s grade plutonium have been testen di the BR-10 experimental fast reactor. Large batches of MOX-fuel pins made by different technologies with plutonium of various isotopic composition have been tested in the BOR-60 research reactor. For many years this reactor has been operated recycling its own plutonium.

With continued successful MOX fuel tests, we expanded the scope of the e prototyptestth o t s e fast reactors BN-35 BN-600d 0an . Both reactors have been fueled from the very beginning with enriched uranium. The PO "Mayak" (Chelyabinsk) semi-industrial MOX fuel fabrication installation "Paket" provides these reactors with MOX fuel. Mayak has an annual capacity of 10 subassemblies. In the BN-350 reactor the tests have been performed with the subsequent investigation and chemical reprocessing of test fuel subassemblies with MOX fuel (350 kg of weapons grade plutonium). More than 2000 such fuel elements have been fabricated and tested in the BN-350 and BN-600 reactors. Post-reactor investigation showed that fuel element endurancs ewa t exhausteno d wit hburn-ua f 9-11po % h.a.

Serious studies were performe Russidn i vibrocompacten ao d fuelthio T s. granulated en d fuel, obtaine y differendb t methods useds wa , . Among thems i a metho f o electrochemicad l uranium oxidd plutoniuan e m oxide coprecipitation. All the fuel elements loaded into the BOR-60 reactor have been fabricated using this technology. Also, subassemblies have been tested BN-35e inth BN-60d 0an 0 reactors.

Successful MOX fuel testing for BN-type fast breeder reactors made it possibl o begiet n desig constructiod nan smala f no l series (3-4 units f fasto ) , plutonium breeder reactors, BN-800 on the South Urals and Beloyarskaya sites. A specia fueX l l planMO fabricatio r fo t n ("Complex-300" plannes i ) buile b to dt to provide fuel for the operation of these reactors. It will produce a maximum

19 subassemblie0 o90 f s annually. Constructio firse o BN-80th tw tf no 0 unitd san of Complex-30 bees 0ha n suspende financiao t e ddu l difficulties reactore Th . s initiae th n li stage ar f constructioeo n while constructio f Complex-30no % 50 s 0i complete.

Works is planned to back fit the "Granat" and "Paket" pilot installations at the Mayak site to meet the latest safety and environmental requirements of the Governmental Regulatory Organization (Gosatomnadzor).

Studies of plutonium utilization in thermal reactors are only now beginning in Russia. For a long time in Russia, much longer than in other countries, plutonium was meant to be used solely in fast breeder reactors as the most effectiv f thio se nucleaeus r material.

Now work is underway to study the fabrication technology of experimental subassemblies with MOX fuel for the WER-1000. Critical assembly "SUPR beins i " g constructe t IPPEda , Obninsk.

General information on MOX production in Russia is presented in the Tabl. e2

The nuclear fuel closed cycle realization started with the commissioning of firse th t Russian reprocessing plant, Mayak 1976n i , RT-e .Th 1 plan multia s i t - purpose enterprise which provides reprocessing of spent fuel from VVER-440 reactors, fast converter reactors (BN-350 and BN-600), icebreaker and submarine transport units, research reactors, and other power units. The plant capacity is 400 t/year in terms of the main fuel type, that is, that of the VVER- 440 reactor, making possible to reprocess the fuel from not only Russian NPPs t alsbu o from foreign reactor e samth f eo s type. About 3000 ton f speno s t fuel have been reprocesse thin di s plant.

TABLE II. MOX PRODUCTION CAPACITY IN RUSSIA

Plant/Facility Reactor Annual capacity Production in 1992

"Paket" at Mayak, BN-350 10-12 FAs 4 FAs Chelyabinsk BN-600 X MO 30 g 0k 100 kg MOX ) (abouPu % 20 t

"Paket" (modified) BN-600 s 40FA since 1993 1 tonne MOX

Facility at RIAR BOR-60 1 tonne MOX 600 kg MOX (Dimitrovgrad) BN-600 (vibropack)

Plant at Chelya- BN-600 60 tonneM sH binsk complex BN-800 (50-60% complete) (VVER-1000)

Plann i t WER- 1000 planner dfo Krasnoyarsk future Note: FA=fuel assembly; HM=heavy metal

20 Spent fuel reprocessing produce maie sth n product, enriched uraniud man civil plutonium. When the RT-1 plant is under full load, plutonium production is 2,6 tonr yearspe . Last year's plutonium production droppe tons6 0, .do t

The total quantity of extracted civil plutonium stored at the Mayak site is abou tons0 3 t . This plutoniu mstores fore i th f dioxidm o n use e di b n di o et fast reactors. The main problem of this forced, protracted storage of civil plutonium is decay of the shot-lived fissile isotope plutonium-241 into americium-241. This isotop vers ei y troublesome fro ecologican ma l poinf o t view. About 10% of the initial unloaded plutonium is plutonium-241.

powerfuw ne A l radiochemical plant RT- 2s undei r constructior fa t no n from Krasnoyarsk s principlIt . e objectiv o ereprocest wile b l s WER-1000 spen operation i firsts t it fuel t n pu I phase . e nb afteo t , r 2005 t wili , l reprocess 1500 tons of fuel per year. The plant cooling pond already contains 1100 tons of VVER-1 spent fuel subassemblies, while about 1000 ton f thio s s fuee ar l currently at other NPPs.

Accumulatio e abovth f o en mentioned plutoniu e Mayath t ma k site, considerable scales of planned plutonium extraction at the RT-2 plant of up to 10 ton f plutoniuo s e expecte th r yearmd pe an ,d releas f ex-weapono e s plutonium of up to 100 tons when intensive deployment of BN type reactors in the nearest future has a low probability, require corrections to be made in the earlier existing strategy of plutonium utilization mainly in fast breeder reactors. doubto n Ther e sar ee near-ter th tha n i t m management e measureb o t e ar s take r reliablfo n e plutonium storage n additioI e .storag th o f t ncivio e l plutonium, storag r releasefo e d ex-weapons plutonium should alse b o arranged. At the same time it is obvious that such storing should not be too long becaus f economiceo , ecological politicad an , l considerations.

3. MANAGEMEN SEPARATEF TO D CIVID LAN EX-WEAPONS PLUTONIUM

3.1. Medium-Term Option

The Ministr Russiae th f yo n Federatio r Atominfo c Energ developes ha y d a basic concepe managementh n o t f extracteo t d d civireleasean l - ex d weapons plutonium based on the following leading principles:

-Russian experience of plutonium management should be maximized;

-plutonium diversion resistance should be considered an important criteria;

-proposed medium-term option of plutonium management should be ecologically and economically acceptable and lead to the development of a long-term fuel cycle option.

Separated plutonium utilization based on a "nuclear center" at the Mayak, including the Shop-300 plant for MOX-fuel manufacture and an NPP with the 3 BN-800 units together with the operating RT-1 plant for chemical processing of

21 uranium spent fuel, fully meets these criteria. Construction of this center has already been foresee Russiae th y nb n nuclear program maie Th .n purposf eo these reactor o utilizt s eswa civil plutonium accumulate t Mayada resuls ka f o t chemical processin f uraniugo m reactor spen t BN-80e fuelTh . 0 reactor design is described by a breeding ratio (BR) of about 1. That is, it was expected that these reactors will use plutonium from thermal reactors as fuel for their first loading only, changin plutoniutheio gn t ow rfuture e thir th msn Fo i .purpos t ei is foresee organizo nt e BN-800 spent fuel processing.

Calculations made show that the problem of not only the accumulated civil plutoniu RT-e t th als t 1mbu a ol release thaal f o t d ex-weapons plutonium could be solved by creating a nuclear center at Mayak with the addition oft one BN- 800 reactor to be constructed on the Beloyarskaya site. It would only be necessar postpono yt e chemical reprocessin BN-80e th f go 0 reactor spent fuel until a considerable portion of already released civil and weapons-grade plutonium will have been involve fuee th l dcyclen i .

Fissionable material nonproliferation is addressed first, by limitation of released civil and weapons-grade plutonium travel within the militarized zone on the Mayak site and second, by minimization of the time required to

50

-50

-100

1990 1995 2000 2005 201(L,201Years 5 2020 2025 2030 2035 2040

Figure 1. Pu balance at the PO "Mayak" as a function of BN-800 reactors number (one through cycle, RT-1 capacity 2.6 t Pu/a)

22 transform already released plutonium into spent fuel. Radioactive fission products present in the spent fuel serves as a reliable barrier against diversion, and the complex isotope composition of civil plutonium makes warheads very difficult to manufacture.

In order to decrease plutonium content in the spent fuel and the costs of spent fuel subassembly storage improven a , d BN-800 reactor o design s nha blankets.

Weapons-grade plutonium consumption for one BN-800 reactor produces abou ton6 BN-801. e yeartsa th f I . 0 reactors operation i wer t epu n according e programth o t l civial , l plutonium accumulate l t releaseMayaal a d d an kd weapons-grade plutonium could be "disarmed" (that is, transformed into spent fuel) during the first two or three decades of the next century. (See Figure 1). Thus, weapons-grade plutonium utilization in the context of the basic concept is inseparably linked wit realizatioe hth prograe th f no fasn mo t reactors using traditional MOX fuel.

In general, it is possible to start "disarming" ex-weapons plutonium using a light water reacto suggestes a r Unitee th y ddb States r RussiaFo . , however, this solution is difficult to reach for several reasons. First, Russia has only just begun its experimental research using MOX fuel VVERs. We are considerably behin r foreigdou n partner thin si s field. Therthermao n s ei l reactor operating in Russia designed to use MOX fuels. Safety levels at the VVERs, even operating with uranium, do not meet long-term safety requirements for new generation reactors. Therefore, licensing a changeover from uranium fuel assemblies to MOX fuel assemblies at operational VVERs is highly doubtful.

Second, a changeover fuel to MOX-fuel loaded core VVERs "disarm" up weapons-grade plutonium would require twice as the number of BN-type reactors because annual consumption level f plutoniuo s r fabricatemfo d fuels for WER and BN reactors are different. In France, for example, they load one- e corthirth ef d o assembliesX wit hMO . This means that five-six times more VVERs wilneedee b l d than BNs.

Third questioe th , f radiotoxicitno f spenyo t fue importants i l s welli t I .- known that long-lived isotopes in spent fuel such as americium, neptunium, and curium complicate both MOX-fuel recycle technology and finding solutions to the long-term waste disposal. To a great extent, these problems are attribute e accumulatioth o t d f plutonium-24o n n speni 1 t fuelspecifie Th . c radiotoxicit f plutonium-24yo time0 4 1s si higher tha nbasie thath f co t isotope plutonium-239. During storage, the plutonium-241 isotope is transformed into the toxic americium-241 with a half-life of 433 years. This isotope is the main contributor to the radiotoxicity of transuranium elements in the spent fuel after the decay of short-lived fission products. During the operation of light water reactors with uranium, abou kg/GWt(e)*yea0 25 t f reactoo r r grade plutonium are built up. About 30 kg of this mass is plutonium-241. "Disarming" ore burning-up weapons-grade plutonium in thermal reactors increases three-fold the annual build-up of plutonium-241, compared with burning uranium-based fuel in VVERs. If this spent fuel is stored for the long-term, a considerable amoune plutonium-24th f o t 1 wile transformeb l d into americium-241, highly complicating further utilizatio f plutoniuno wastd man e disposal.

23 Fourth, the burning of weapons-grade plutonium at VVERs may produce several times more minor actinides as compared with the VVERs using uranium- based fuelburnin e basie Th th . f cg o isotope plutonium-239 during weapons- grade plutonium utilizatio t Mayana k facilities would resulaccumulation a n i t f no nuclides wit a totah l radiotoxicity three times higher thae nth thaf o t VVER/uranium variant.

The situation is quite different however, if we use BN-type reactors to utilize weapons-grade plutonium radiotoxicite Th . "disarmede th f yo " plutonium does not, for all practical purposes, exceed that of the initially loaded plutonium. Figur 2 showe s efficiency indice r weapons-gradfo s e plutonium utilizatio VVERn i BNsd VVER/uraniue san Th . m radiotoxicity inde gives xi r nfo comparison. We can see that BN gives the best radioecological effect; it is the most effective burne f plutonium-241o r . Therefore, wit l othehal r conditions being equal, it is more logical initially to utilize reactor-grade plutonium and only after, weapons-grade plutonium. The radiotoxicity index factor is highly important for the Mayak plant conditions with its difficult ecological situation.

The questio f plutoniuno m utilizatio therman i l reactor mann si y countries is connected with the relatively high cost of fast reactors. In Russia, the electricity generate firse th r cent y pe BN-60db t 0 4 mor s 0i e expensive than that of WER-1000. But the experience of BN-600 was taken into account in BN-80e th 0 metae projecth d l consumptioan t n facto e r centh pe f onls i o rt0 y8

16000

12000 -

8000 -

4000 -

-4000 -

-8000 WER WER BN WER WER BN U with WPu with WPu with with CPu with CPu 1/3 CPu

Figure 2. TRU radiotoxicity evolution after 1st recycle of weapon Pu (WPu) typN B e d reactorsanan d R civi (CPu.u P lWE n i )

24 BN-600 project improvee Th . d economic characteristic e BN-80th f o s0 fuel cycle have been reached throug e transitioth h n from uranium (whics i h ineffective for fast reactors) to MOX fuel and the further improvement of the burn-up factor.

The economic characteristic thermad e fasth an t f o s l reactors have also been compared , e increasebaseth n o d d safety standards requirer ou n i d country after Chernobyl. Inherent characteristics of fast reactors, as well as new technical solutions, enabled us to upgrade the BN-800 project to that of a world-level nuclear powe generationw rne plana f o t , with improved safety characteristics onle th ys i projec t I . t whic passes hha d all, including ecological, necessary certifications. Local authorities have already approve e projecdth t and have reache agreemenn da t with Minatom.

Bu t completewor no projectw s i kne n o ds with new, safer uranium reactors (VVER-500, VPBER-600). Preliminary estimates show that average capital investment in these projects is higher than in the VVER-1000, but corresponds approximately to the investment necessary for the construction of three BN-800 unit Yuzhno-Uralskayas a s . Another e factoconsidereb o t r s di that Russia has no industrial-scale MOX fuel fabrication for VVERs. Such a technological line is to be at the RT-2 facility, to be commissioned after 2005.

Regarding BN-type reactors mentiones a : d above Complex-30e th , 0 5 s 0i per cent completed. With sufficient investment, it can be commissioned by the end of the century, before the commissioning of the first BN-800 unit. After putting into operation three BN-800 units, Complex-300 will reach its maximum capacitfueX l MO fabrication n yi .

3.2. Long-Term Option

The second stag f o plutoniue m technology masterin n i Russig s i a connected with star p RT-u t 2 plant aftee yeath r r 2005 (near Krasnoyarsk) meant for VVER-1000 reactor spent fuel reprocessing. The projected quantities of plutonium to be extracted will be much larger than the possible needs of fast reactors. Plutoniu extractee b mRT-e o t th usee t 2db a plann di o t s i t VVER-1000 reactors speciaA . l worksho beins pi g planne e RT-th t 2d a plan t fueX fol MO rfabrication .

understoos i t I d that plutoniu therman i e mus l reactor forcea s si d decision limitee s numbei th d y f dan b recycleo r s becaus f radiotoxieo c minor actinide accumulation. Fast reactors will complete the fuel cycle for thermal reactors, solvin e problemth g f plutoniuo s m isotopic compositio e burninth d f o gan n minor actinides.

The aim of these system studies is to determine the optimal parameters of thermal and fast reactor cores and the optimal composition of the nuclear power grid. Criteria for fuel cycle efficiency should encompass all aspects of economics, ecology and fissile material nonproliferation.

At the present time in Russia, as it was mentioned earlier, the possibility of using civil plutonium in WER reactors with MOX fuel up to 1/3 of the

25 reactor core is being investigated. Also, calculational physical research is underway in the area of WER reactors with a full load of MOX fuel. Cermet fue alss i l o under developmen o providt t e more effective plutonium burning and increase the safety of the reactor core.

As for breeder reactors, the key direction of research is destruction of plutonium excesse radiotoxid an s c minor actinides store therman di l reactors. For these purposes new reactor cores with MOX fuel, and increased content of plutonium in fuel without blankets are under consideration. Also, new fuel composition on the basis of an inert matrix instead of source uranium are being developed.

Along with the traditional uranium-plutonium cycle, fuel cycles with thoriu d uranium-23man e undear 3 r consideration. These will o t allo s wu decreas minimua o et quantite mth f plutoniuyo nucleae th mn i r industrd yan minor actinide wastesn si .

In the Figure 3 different options of plutonium management are presented as a scheme. Some of them are already being realized in the countries with developed plutonium technolog d som e hypotheticaan yar e l ones, being considered, as a rule, in the countries having no experience of plutonium utilization e thinreasonabls i W .t i k e tha l sucal t h option e compareb s d using the same methodology and mutually agreed upon criteria.

It is obvious, that the aim of such a comparative analysis is not a refusal from Nuclear Power developmen plutoniud an t m utilization scenari n suci o h

Pu Disposition Options

_L Use for Energy Production Disposau P l Options

Conventional | Nonconventional MOX MOX MOX BN-800 WER | WER BN-Pu-Th HTGR Fusion Vitrification Single 1 Accelerator i 1 \ 1 WER Poisoning Pu-Th Open U-233 I Fuel Cycle BN Outer space V Cermet fuel _L Nuclear Explosion WER Breeder (Pb-Bi) ± MSBR

Figure 3. Plutonium Disposition Options. Comparison criteria: economics, safety, ecology, non-proliferation Boundary conditions: technical evolution, adaptation, reality framee th in f traditionaso l nuclear industry development

26 countries as France and Russia. It is an exchange of information, multiparametric scientific analysis ensuring choice of a scenario optimal in the frames of national strategy, multiobjective substantiation of this strategy.

CONCLUSIONS

1. The Russian concept of plutonium management (both civil and weapons) is based on the postulate of the outer fuel cycle closure, necessity to enhance fuel efficiency, and decreasing radioactivity of disposed long-lived wastes.

. Short-ter2 m plutonium managemen Russin i t s basei a safn d reliabldo an e e storag f separateo e d civil plutoniu "Mayak"O P t d ex-weaponm(a an ) s plutonium until it can be used in reactors.

3. Medium-term plutonium disposition option basee sar developmenn do a f o t Nuclear Power Center at PO "Mayak" (RT-1, Complex-300 and 3 BN-800s) accumulatee us o t d civiex-weapond an l s plutoniu mfasn i t reactors.

4. Long-term options to be defined consider use of plutonium in VVER-type reactore burninth d f plutoniugo an s m exces d minoan s r actinide fasn i s t reactors witcoresw hne .

5. Investigations are underway to estimate plutonium utilization options, taking into account non-proliferation, environment, safety, healt costsd han .

6. International cooperation with the task not only to develop current technical determinpolitio t t cbu e optimum long-term dispositio desiredns i .

Next page(s) Left blank 27 AN OVERVIE ACTIVITIEE TH OECF E WO TH DF SO NUCLEAR ENERGY AGENC PLUTONIUN YO M

. ZARIBAN S OECD/NEA, Issy-les-Moulineaux, France

Abstract

papee Th r present framewore splutoniuth A NE f ko m activitie summarized san s past and present relevant work.

. 1 INTRODUCTION recognises i t I d that civilian plutoniu monlt managemenno ye relatear e us o d t d tan technological, economic and safety considerations but also have wider political, non-proliferation, national and international policy and public acceptability implications. increasee Th d stock separatef so d plutoniu postponemene th d man ever to e nth abandonmen implementatioe th f to fasf no t reactor programmes have resulte growina dn i g interest, in a number of countries, in recycling plutonium in light water reactors. While uranium requirements, enrichment and fuel fabrication capacities and requirements, reprocessing capacit spend yan t fuel productio reliable b n nca y predicte e basie th th f sn o d o present operating and scheduled nuclear power capacity until about 2005, future separated plutonium inventory forecasts have large associated uncertaintie quite ar ed sensitivsan o et mixed oxide fuel fabrication and reprocessing assumptions. Quantitie separatef so d civilian plutonium have been increasin pase 0 th t3 r gfo expectee yearar d san continu o dt increaso et neae th r n efuturei . Unde firssecone d rth an t d Strategic Arms Reduction Treatie unilaterad s an (STAR) H d l pledgean T I s mady eb American and Russian Presidents, many thousands of U.S. and Russian nuclear weapons will retire within the next decade. As a result, a large inventory of weapon-grade plutonium on each side is expected to become surplus to military needs. Part or all of this surplus plutoniu eventually mma y reac commerciae hth l market. Quantitatively impacs e ,it th n to world natural uranium supply markets is not likely to be significant, but existing civilian fuel cycle facilities, services and programmes will be affected and should probably need to be adjusted accordingl copo yt e with this material. Policies adopted by governments will be affected by their knowledge of technologies available in the civil fuel cycle and of likely costs. These aspects, together with scientific studies that provide supporting information, are the subject of work at the NEA - other aspect treatee sar othedn i r international bodie bilateran i r so l arrangements between governments.

2. THE GENERAL FRAMEWORK OF NEA PLUTONIUM ACTIVITIES Technical NEA Committees have many years of accumulated experience in dealing with sensitive issues, including among others plutonium economics, safety, related technologica sciencd lan e aspect recyclind san g logistics. Relevant work, whic usualls hi y performe ad-hoy db c Expert Groups consistin nominatef go d high-level national expertss ,i primarily addressed to OECD Member Governments. The resulting reports present unbiased views whic alwaye har s approve representativey db technicaf so CommitteeA lNE s before publication. Historically, circulation of those consensus documents has been quite wide and attracte interese industryde th th f t o genera e th , lmedia e publith d .can

29 Past and present NEA plutonium work has been, essentially, technical in nature. Institutional aspects, non-proliferatio physicad nan l securit treate e Internationae yar th dn i l Atomic Energy Agency and elswhere. Staff members of the IAEA and the Commission of the European Union take active part in NEA activities and provide valuable input, ensuring at same th e time proper co-ordinatio optimad resourcesf nan o e lus . The objectives of the NEA plutonium studies are generally targeted on the need to identify suitable technical solutions, despite the existing large political uncertainties associated with their implementation. In this respect, it should be noted that national programme policied san quite sar e diverse. Public opinion concerning plutoniu, mis however, quite sensitiv almosn ei countriel tal s irrespectiv material'e th f eo s original nature, b civiliat ei militaryr no .

3. PAST ACTIVITIES The NEA has had in the past a number of plutonium related activities. In the mid-80s the NEA's Nuclear Development Committee (NDC) conducted a study on plutonium and published in 1989 the resulting report: "Plutonium Fuel - An Assessment". This provided the facts and then current views about plutonium and its civil use in the short and medium terms. It addressed questions influencing the choice of fuel options and illustrated how economic and logistic assessments of the alternatives could be undertaken. Plutonium handling and safety issues, based on available experience in OECD countries, were addresse "Safet e Nuclea199stude e A th th th n 3n df yNE i o yo r Fuel Cycle". The report underlined the extreme care which has been taken in both the design of plutonium handling facilities and the detailed operational procedures that have been adopted and noted their very good safety records. Regardin economice gth plutoniuf so m use recentle ,th y publishe studA n ydo NE the "Economics of the Nuclear Fuel Cycle" included updated analyses of the economics of MOX fabrication and of the entire reprocessing cycle.

4. CURRENT ACTIVITIES 4.1 Expert Group on the management of plutonium Ther clearls ei yneea decisionr dfo s abou managemene tth plutoniumf to . Althoug numbeha relatef ro d national activitie exis o internationasd d tan l meetingd san conferences are at present frequently organised, it was felt that the authoritative consensus repor internationan a f to l expert committee woul vere db y welcome decision-makery db d san industrialists alike. During 1993, an informal group of knowledgeable government officials and industry experts develope initia studn da A ylNE proposa plutoniun lo m management, taking into account that the resulting NEA report would be of use primarily to governments in evaluating their policie presentind san g publice theth mo t . That proposal has turnn ,i , been accepte NEA'e th y db s responsible Committees. An ad-hoc Expert Group was, therefore, assembled in early 1994, under the auspices of the NEA's Nuclear Development Committee, with the task of identifying, examinin evaluatind gan broae gth d technical questions relate plutoniuo dt m management. Recognising tha subjeca t this sintereswa f o t l countriesal o t , whethen i thet d no y ha r ro stock separated plutonium or any intention of using plutonium, the following nominated participant Expere th o st t Group: Australia, Belgium, Canada, France, Germany, Ireland, Italy, Japan, Korea Netherlandse ,Th , Norway, Switzerland, United Kingdo Unitee th d mdan

30 States. Exceptionally, experts from Russia were invited to participate in this activity because of the accumulated experience in that country with plutonium production, handling and use. The IAEA and the European Commission have also nominated members for the Expert Group. The Expert Group met in Paris in full session twice and its third meeting is foresee Sprin r . Designatenfo g1995 d sub-groups already produced valuable contributions alon linee gth s agreed durin meetingse gth . Accordin curreno gt t plans expectes i t ,i d thae th t group's final report wil publishee b l . earln di y1996 scope stud A Th thif e o detaile s ysNE i d below: "Poin departuref to : Plutonium arise operatinn si g reactors .hels i Mos t dn i i f to spent fuel. A growing amount has been separated but not yet used. There is public interest managemene inth thif to s plutonium includin becom y thay gan ma t e availabl civie th lo et market from ex-military uses. The NEA study is concerned with the technical options for management of this plutonium.

The study's main analyses will primarily address: A. Technologies that have already been implemented which providr efo medium-term storage of plutonium or for recycling the plutonium through reactors. A review will be provided of experience with them and a technical commentary on their potential deployment ove nexe . year0 rth so 2 t r so . B Future technologie joinee b sfurthe a thay y db tma r range somwhicn i , ehare cases, already under development technicaA . l revie additionae th wf o l options thay tma become available wil provided.e lb " As mentioned before, neither institutional matters whould be considered by the expert group nor military plutonium as such, but note will be taken of any specific problems or opportunities associated with plutonium of different isotopic composition. The study is not expected to make any policy recommendations, however, all plutonium technical management options, reflecting both the current experience and future developments will be covered. Accordingly, the Expert Group agreed during its second meeting tha structure tth repors it f eo t shoul followss a e db :

1. Introduction - Scope/objectives - Total plutonium arisings - Separated plutonium - Plutoniu mspenn i t fuel - Reprocessing - Report organisation . 2 Experience gaine currend dan t trends - Storage - Purification - Transportation - MOX fabrication - Experience with constructin fabricatioX gMO n plants - MOX in reactors: . thermal . fast .CANDU - MOX reprocessing

31 . 3 Future developments - Advanced reactors, cores, fuels and fuel cycles - Disposal of plutonium . 4 Country programmes

2 4. Working Part physicn yo f plutoniuso m recycling The Nuclear Science Committee of the NEA set up in early 1993 a Working Party deao t l wit statue trendhth d san physicf so s issues relate plutoniuo dt m recycling with respec tback-ene botth fueoptimae e o ht th th l f cyclo dt o ld utilizatioean plutoniumf no . objectivee Th thif so s projec provido t e ar t e Member Countries with up-to-date information on: cor e fued th ean l acycl) e issue: sof . multipl1 e recyclin plutoniuf go partly-loademn i d LWRs fuel X wit,hMO 2. full MOX fuel recycling LWRs cores; flexibilite th b) fasf yo t reactor produco st buro t r neo plutonium within standard fuel cycles (eg.: using MOX fuel), or advanced fuel cycles (eg.: metal, nitride fuels); c) the core and fuel cycle physics issues related to the use of plutonium fuel without uranium (eg.: plutoniu inermn o t matrices enhanco )t e plutonium burning capability; d) the physics issues related to the use of uranium from recycling; core fued th ean l ecycl) e issue plutoniuf so m recycling through advanced converter reactors (eg.: CANDU PHWR, ATR). Workine Th g Party will also provide advic nucleae th o et r communite th n yo developments neede meeo d t requirement e tth s (dat methodsd aan , validation experiments, strategic studies) for implementing the different plutonium recycling approaches. Belgium, Canada, France, Germany, Italy, Japan, the Netherlands, Switzerland, United Kingdom, Unite sIAEe Stateth Ad san hav e nominated member Workine th o st g Party which has already met twice. A number of benchmark exercises have been completed and the Group is expected to finalize its work and produce its state-of-the-art report on scientific aspect plutoniuf so m recyclin early gb y 1995.

. 5 SUMMARY The growing plutonium inventories may well exacerbate the discontent with nuclear energy expresse somn di e quarters extene th o tT .thasees i desirablt s ti na e thae tth optio usinf no g plutoniu mediume th mn i long-tero -t m shoul maintainede db shoult ,i e db demonstrated that civilian stocks of plutonium will continue to be managed well in the short-term. Possible national approaches should give full consideration to the following: rationalee th . decisionr sfo s adopte thosy db e countries intereste pursuinn di g plutonium use; availabilite th experienc d . an f yo usinn ei g technolog segmentl al r e yfo th f so plutonium recycle route;

32 . the potential for further improvements in technology aiming, for example, at even higher standard environmentaf so l protection, safety, worker healtd han reduced costs; . priorities for further research and technical developments.

The past and present NEA programme on plutonium aimed at addressing the above issues and provided assistance to Member Governments, decision-makers and the industr produciny yb g authoritativ comprehensivd ean e evaluation relatef so d technologies, scientific aspects and economics. The NEA will continue in the future with similar activities and may also expand its programme, as necessary.

Next page(s) Left blank 33 WASTE MANAGEMENT ASPECTS OF (Th, Pu)O2 FUELS P. TAYLOR, W.H. HOCKING, L.H. JOHNSON, RJ. MCEACHERN, S. SUNDER AECL Research, Whiteshell Laboratories, Pinawa, Canada

(Presented by D. Menetey)

Abstract

__ ^ This repor briea s ti f overvie issuef wo s relevan performance th o t irradiatef eo d (Th,Pu)O2 as a waste form for geological disposal. Fuels of this type, among others, are being considere burninr dfo g plutonium from dismantled nuclear weapon powen si r reactors, including CANDU systems. The high chemical stability and low aqueous solubility of thoria make this type of fuel attractive as a waste form. In contrast with fuel2 inertnese th UO , thorif so oxidatioo at n dominate schemicae mosth f o t l issuef so fuel disposal. The overall performance of a thoria-based fuel waste form is likely to be determine "instante th y db inventoriep " ga releas e th f mobilef o s o e fission products such V2s a *I. This turnn i , , wil controllee b l d largelin-reactoe th y yb r power histord yan probably also by details of fuel fabrication. Limited experience with thoria-based fuels (chiefly (Th,U)O^) indicates that, for given power and burmip levels, gas releases can be substantially lowe grain-boundard r an tha p nga fuelswit2 e hTh UO . y inventorief so fission products are expected to be correspondingly low. A fabrication route involving molecular-level mixing (e.g., sol-gel process) would be preferable to powder blending, because microscopic heterogeneities in the fuel could adversely affect the retention of fission products. Pilot-scale irradiation, post-irradiation examinatio leachind nan g studies are required to support this preliminary assessment. Other issues that need to be addressed include impurity specification minimi?o s(t ^ formatio long-livef no d activation products) criticalitd an , safeguardd yan s issues that might influenc desige eth f no fuel-handling facilities.

1. INTRODUCTION

There is an urgent need to identify appropriate methods to deal with the plutonium recovered from dismantled nuclear weapon militard san y stockpiles makd an , t ei inaccessible for future weapon use. At present, debate centres on the options of either direct disposa plutoniue th f lo nucleaa mn i r waste glass burninr o ,nucleaa n i t gi r reacto disposind an rirradiatee th f go d plutonium-bearing fuel (Holdren 1994)e Th . latter approach has the advantage that useful energy would be obtained from the plutonium, and hence some return could be obtained on the large investment A number of different reactor and fuel types are being examined for then* capability to burn weapons plutonium. The CANDUfljreactor system is particularly attractive, flexibilite becausth s ha operat o t yet i e wit hvarieta differenf yo t fues li t typesi d an , abl accommodato et efula l core loadin plutonium-bearinf go g fuel (Dastu 1994)L a t re .

Three generic fuel types could, in principle, be used to bum plutonium in a power reactor Uranium-base) :(a d mixed oxides, (U,Pu)O Thorium-base) 2;(b d mixed oxides,

[1]: Canada Deuterium Uranium, registered trade mark of Atomic Energy of Canada Limited

35 (Th,Pu)O2; (c) Inert-matrix fuels, in which PuO2 is dispersed in an inert diluent such as zirconia (Aide et al. 1994). The (Th,Pu)O fuel option is attractive because breeding of

^U would allow conversion of over 90% o2 f the ^u at high burnup in a once-through cycle (Veeder and Didsbury 1985). 2

Thorium-based mixed oxide alse sar o appealing fro mwasta e management perspective, because ThO chemicalls i y stabl almosd ean t insolubl groundwatern ei thin I .s report,

we explore waste-managemen2 t aspects of thoria-based plutonium fuels. We compare soine of the key chemical and physical properties of UO2 and ThO^ and draw from experience in the assessment of geological disposal of irradiated UO2 fuel (Johnson et al. 1994), to identify issues that would need to be addressed in a thorough assessment of thoria-based fuel disposal assume W . e tha fuee th tl would contain abou wt.t3 % PuOn 2i ThO. We also assume a similar disposal scenario to that envisaged in the Canadian

Nuclea2 r Fuel Waste Management Program, i.e., direct disposal of used fuel bundles in corrosion-resistant containers, surrounded by a clay-based buffer material, within a vault excavated deep in granite (Hancox and Nuttall 1991).

. 2 CHEMISTR THORIF YO A

1 2. Redox Chemistry

By far the most important chemical difference between ThO2 and UO2 is the fact that thoriu mpresens i maximus it n i t m oxidation state, Th(IV), whereas uraniut no ms i Under oxidizing convertee conditionsb n comparativele ca th 2 o dt UO , y soluble uranyl cation, UO^, and its derivatives. This reaction and the corresponding reduction dominate the geochemistry of uranium (see below), and an understanding of the kinetics of oxidative dissolution of UO2 is central to the performance assessment of irradiated UO2 fuel as a waste form. Oxidative dissolution of the matrix is not an issue with thoria fuel. Redox conditions could affect the leachability of 233U from irradiated thoria, but this would be limited to surface dissolution and is unlikely to be a major concern.

The inertness of thoria to oxidation is also relevant to interim dry storage of irradiated fuel before geological disposa maximue Th l m acceptable temperatur storagy dr r f efo eo CANDU UO2 fuel in air is typically 150-175°C, because at higher temperatures oxidation

defecteUo n Ot i oUO f d element causn sca e powderin fuee th l f gmatrio d xan 3 8

2 splittin fuee th l f gcladdino g (Boas Vandergraad ean f 1977, Hasting 1986L a t s,e Novak . 1983al t e , Wasywic . 1993)al t he . Somewhat higher temperature acceptable b y sma r efo higher-burnup UO^ such as LWR fuel, based on data reported by Thomas et aL (1993). Matrix oxidation is not an issue with thoria-based fuels. Furthermore, the thoria structur easiln eca y accommodate oxidatio minof no r solid-solution components sucs ha (seu P ed Sectioan U n 2.4) r exampleFo . , detailed investigatio U-Th-e th f no O system has shown that thoria-rich solid solutions, (Th^UjC^+y), are stable in air if x I 0.5, and U3O8 is only formed from more uranium-rich compositions (Cohen and Berman 1966). Thus, fuel oxidatio unlikel s nconceri a e b o yt n durin storagy gdr (ThJftOOf eo d ^an henc maximue eth m storage temperature woul limitee db somy db e other factor, probably cladding degradation (EPRI1989).

2 2. Aqueous Chemistry

Most metal(IV) dioxides have very low solubilities hi aqueous solutions except at extrem conditionsH ep under o , r redox conditions favouring either reductiv oxidativr eo e dissolution. Because ThO iners 2i oxidatioo t extremeld nan y difficul reduceo t t s i t i , virtually insoluble over a wide range of solution conditions.

36 solubilite Th crystallinf yo e thori t 25° aa pH>5d absence Can th n i ,complexinf eo g agents bees ha , n estimate t 10"da mol/kg part2 quadrillior r o ,spe n (Langmuid an r Herman 1980). Eve14n lower solubility estimates for crystalline thoria have been suggested bi (1987)y RyaRa d ,nan base solubilitn do y measurement hydroun so s thorium(IV) oxide estimated an difference th f so stabilitn i e y betwee hydroue nth crystallind san e forms. solubilite Th y increase fallH sp rapidls e beloadditionn th I s ya . w 5 , both inorganid can organic ligands can enhance the solubility of ThO2 in neutral to acidic groundwaters (Langmuir and Herman 1980, Felmy et al. 1991).

Lemire and Garisto (1989) calculated solubility limits of thorium in groundwaters contactin granitia 2 fue n i gl UO c waste typ e vaulth e f envisageo t Canadiae th n di n Nuclear Fuel Waste Management Program calculatee Th . d values represene th t solubility of hydrous thorium oxide, as opposed to crystalline thoria, and are therefore very conservative (i.e., overestimates purpose presenth e r th )fo f eo t repore Th t calculations were performed for a distribution of water chemistry parameters representing expected conditions in a granitic waste vault, and a distribution of thermodynamic values representing experimental uncertainty limits mose Th .t frequent values for the calculated thorium solubility (for 40,000 randomly selected cases) are 9 5 11 about 10' - mol(Th)/kg(H2O) at 25°C and 10" mol/kg at 100°C. These values are slightly higher and slightly lower, respectively, than corresponding values for uranium under reducing conditions. Smal finitt bu l edistributione tailth n si s extende muco dt h higher solubilities; in the case of thorium, these are mainly an artifact of high uncertainty limit somn so e complexatio t thoughn realisticonstantsno e b e o ar t d c (RJan , . Lemire, personal communication, September 1994).

We conclude that ThO2 solubilit extremels yi (comparablw ylo unde2 UO r o et reducin g conditions) over a wide range of aqueous solution conditions. Nonetheless, it would be desirabl perforo et m experiment calculationd san defino st e more accuratele yth conditions (e.g., pH and groundwater compositions) under which solubility may be elevated. The release of actinides and those fission products that are retained by the thoria matrix is expected to be limited by the solubility of ThO2. Such release would be exceedingly engineeren sloa wn i d disposa l type vaulth ef o tenvisage CANDr dfo fue2 n I UlUO nature, thorium is often closely associated with clays or oxide minerals such as bauxite. Thus mighe on , t expect some acceleratio thorif no a dissolutio couples i t i nf a i dvi diffusio sorptioo t n clan no y buffer materia r othelo r mineral surfacese Th . enhancement of the dissolution rate of UO2 as a result of mass-transport-precipitation couplin bees gha n studie Garisty db Garistd oan o (1986, 1988). This phenomenon becomes importan redoa f i t x front exists nea dissolvine rth resultge UOTh . s imply

thadissolutioe th t n rat thorif eo a coul similarle db y enhance coupliny db strono gt 2 g adsorptio buffere th n i , althoug impace hth likels smallee i t b o yt r because th f eo extremely low solubility of thorium under the expected groundwater conditions.

Similar arguments would appl possiblo yt e alteration processes conversio,e sucth s ha f no

ThO ThSiOo t reactioy b n with dissolved silica othen I . r words, this conversion wile b l 4

exceedingl2 y slow in typical groundwaters, even though it is thermodynamically favoured (Kaminen Lemird an i e 1991).

23 Radiation Effects

Radiation damag nucleao et r fuel bees ha sn studied extensively (Matzke 1982)n I . general, radiation damag refractoro et y oxides increases their susceptibilit dissolutioo yt n

37 and leaching. However, dissolution experiments, along with isotopic studie mineralf so s suggest that radiation damag thorio et fairls ai y readily annealed (Eya Fleisched lan r 1985a, 1985b, 1985c, Vanc Gascoynd ean e 1987). Neithe susceptibls i 2 r ThOUO r 2eno metamictizatioo t n (radiation-induced transformatio amorphoun a o nt s state). Thin sca be understood on the basis of a low crystallization temperature relative to the melting point; recrystallizatio highls ni y favoure thermae th n di l spike associatedn witio n ha track. The low degree of radiation-induced damage, along with a tendency for defects to anneal, suggests that radiation-enhanced dissolutio ThOf no 2 fuel wil slighte lb .

Radiation-induce (U,Pu)Odd enhancemenan diffusio u 2 P d t UO 2a an n ni U f o t temperatures below 1100° welCs i l known (Matzke 1992) impacn ;a thi s wastn sha o t e management issues because in-reactor grain growt limites hi metay d b diffusion n lio . significanThero n e ear t difference diffusioe th n i s n rate tetravalenf so t metal ions (Th , U4*, Pu4"1") in ThO^ and the rate of metal ion diffusio4+n is somewhat slower than in stoichiometric UO2. We thus expect in-reactor grain growth and related fission-product segregatio same occuo nt th eo t rextent UOlessn r i o , s 2.a , effece Th radiatiof o t rate fission-gaf th e o n no s releas qualitativels ei same y th botr efo h UO thoria-based 2an d fuels (Nai. 1981al t k,e Kaima . 1990))al t eace ln I . h case eth

diffusion coefficient at a given temperature is constant at low doses, then drops

4 3 dramaticall dose th increases es i y a 1Ôo d t fissions/m 2 because fissio atoms nga e sar trappe radiation-inducey db d defects (Matzke 1992, Naik 1992). Radiation-induced damag thus ei t expecte sno havo dt e adverse rate effect fission-producf th eo n o s t release from a thoria-based fuel.

separatA e radiation effecissue fue2 th use r radiolysif es o fo i tl d UO e watef sth o n ro oxidative dissolutio (Sunde2 1990L UO a f nt ,o re Shoesmit Sunded han r 1992). This si another process that involve oxidatioe sth concera tJ(IVf nt o no U(VI)o s )t ni witd an ,h thoria therefors i t I . e unlikely that radiolysis will hav significany ean t influenc thorin eo a dissolutio disposaa n i l vault.

2.4 Natural Analogues

Useful information abou behavioue disposath a 2 fuet n i lUO f o rl vaul bees ha t n inferred fro mineraloge mth groundwated yan r chemistr uraninitf yo e mineral depositsn A . example that has several close parallels to a fuel disposal vault is the Cigar Lake ore body in Saskatchewan, Canada (Cramer and Smellie 1994). Natural analogues for thoria fuel disposa more ar l e tenuous because thoriu t highlno ms yi concentratey an y db geochemical processes; thi anothes si r consequenc inertness it f eo redoo st x processes (Langmui Hermad ran n 1980, Baye. 1974)al t additionrn e I . , mineral whici sh h thorium is the major component, such as thorianite (ThO^ and thorite (ThSiO4), are rare. Large thorianite crystal occuo sd r rarel somn yi e pegmatite rocks (Baye . 1974)al t re . Even though there is no natural formation rich in thorianite, the general immobility of thorium in instructivee naturb y e ma example r Fo . , concentration dissolvef so d thoriumn i surface waters are exceedingly low (Langmuir and Herman 1980). Natural deposits of thorium usually occur either as resistate minerals (i.e., insoluble residues of chemically resistant accessory minerals, suc monazites ha substitutea , d rare-earth phosphate, that remain after rock weathering) or through adsorption and retention on clay or oxide alteration minerals (if the thorium originates as a minor component of the major rock-forming mineral weatherea n si d formation).

38 2.5 Compatibility of Actinides with Thoria

Thoria crystallizes with the fluorite structure, as do all other actiniae dioxides (Wells 1975). Extensive solid-solution formation occurs between these oxides, and the fluorite structure can also accommodate substantial levels of actinides in other oxidation states, suc Am(nis ha U(VI)d )an wels a ,mans a l y fission product 1966x s (BretFo ,d an t Kleykamp 1985, Mayer et al. 1993). Thus, no phase segregation of actinides is expected to occur within the fuel, either during operation or after disposal, and it is reasonable to assume that release of actinides will be controlled by the slow dissolution rate of the thoria matrix, provided that the fuel is initially homogeneous. Partial segregation of uraniu plutoniud man s beemha n show occuo nt incongrueny rb t vaporization undee rth steep temperature gradients experienced under severe reactor operation conditions (Green et al. 1983, Baptiste 1984). This is unlikely to be sufficiently extensive to affect the waste-form performance of mixed actinide fuels, but it should be addressed in post-irradiation examination of (Th,Pu)O2 fuels.

3 FISSION-PRODUCT SEGREGATION

Calculated environmental releases and subsequent radiation doses arising from a CANDU UO2 fuel disposal vault are dominated by the "instant" release of soluble and mobile fission product particularn s(i , 129I) fro fuel/sheate fuee mth th l f o p hga (Stroes-Gascoyn . 1987al t ee , Johnso . 1994)al t ne . Grain-boundary inventorie alsy osma releasee b d rapidly compares a , d with matrix dissolution likels i t I .y that similar findings would emerge from a detailed assessment of thoria/plutonia fuel disposal, especially given our expectation of extremely slow matrix dissolution. Therefore, it is important to consider the irradiation history and microstractural behaviour of the fuel, and to have reliable information on the segregation of mobile fission products to the gap and grain boundaries in thoria-based fuels. Such analysis is also important because several fission products have been implicated in fuel failures caused by stress-corrosion cracking of the cladding (Cox 1990) additionn ;i , high level fission-gaf so s releas lean elemeneo ca dt t swelling.

At present, international experience with irradiatio thorif no a fuel generaln i s d an , informatio detailen no d post-irradiation examinatio particularn ni limiteds ,i . Promising results have been reported on gas releases from thoria-based fuels (see Section 3.1). In reviea post-irradiatiof wo n examinatio variouf no s plutonium-bearing fuels, Cadelld ian Uppens (1984) refer only to isotopic analysis of (Th,Pu)O2 fuels. Pinheiro et al. (1988) concluded that (Th,Pu)O2 fuels show promise for extended-burnup, once-through use of recycle plutonium t mos bu ,theif o t r materials characterization wor e bases th kwa n do use of cerium as a surrogate element for plutonium in fuel formulations. Katz (1986) describe metallographe dth thoria-basef yo d fuel materials t witbu ,h emphasi methon so d development and the importance of impurities in fuel fabrication. He reported nothing unusual in the porosity and fission-gas bubbles observed in irradiated (Th,U)O2 fuel from the Shippingport reactor.

3.1 Grain Growth and Fission-Product Segregation

Gram growt centrae th n hi l regio fuef no l pellet majoa s si r caus fission-gaf eo s release gape toth , becaus gasee otheeth d san r incompatible element swepe sar t from their original resting places in the fuel matrix and become concentrated at the grain boundaries. There, they form features suc fission-gas ha s bubble noble-metad san l particles (Matzke 1980). Interlinkage of fission-gas bubbles on grain-boundary

39 intersections eventually creates tunnels that permit ventin othef go r fission producto st the fuel-cladding gap. Thorium oxid somewhaa s ei t better thermal conductor than UO2; i thighe a alss oha r melting poin slowed an t r cation diffusion (Section 33). Therefore, for a given power rating and fuel geometry, it would be expected to run cooler and undergo less gram growth (Meiera Westermand nan n 1977). Bot grain-growte hth h kineticthermae th d san l conductivity, however sensitive ar , impuritieo et s suc fissios ha n products. Therefore, direct measurements of grain growth and gas release under power-reactor condition essentiae sar identifo t l y appropriate operating conditiono t d san assess disposal performanc fuele th .f eo

Fission-gas release dat (Th,Pu)Or afo availablet no e t indirec2ar bu , t evidence suggests that release rates will be somewhat smaller for thoria-based fuels than for UO2. This conclusio lowe e bases ni th n rdo diffusion rat xenor efo ThOn i 2 tha2 (MatzknUO e 1992, Shiba smalle e 1992th d )ran burst releas ThOn ei 2 (Kaima . 1990)al e t le Th . expecte releass ga w edlo rates from thoria-based fuel supportee ar s in-pily db e experiments on ThO2 and (Th,U)O2 fuel assemblies. Goldberg et al. (1977, 1978) measured fission-ga thoria-base1 5 f so t releasse a d n efuei l rods ove ranga r lineaf eo r powers, burnups and compositions. They gave an expression for the rate of fission-gas release, which suggests that rates are considerably lower than for UO2 under comparable operating conditions. Accelerate releass dga bee s eha n reporte t higda h buraur pfo (Th,U)O onsee 2 fuelsth thif t o t sbu , phenomenon appear delayede b o st compares a , d fue2 witlh UO (Gioveng . 1981)al t oe . Also, higher-than-expecte releases dga s have been observed in some Canadian tests with (Th,U)O2 fuels. This has been attributed to a granular microstructure, which both impeded heat transfe provided an r d networkf so porosity that facilitated gas release (Smith et al. 1985).

In many cases, the segregation and hence the leachability of volatile, non-gaseous fission products, suc cesius ha iodined man correlates i , d with fission-gas release (Stroes-Gascoyn . 1987)al thut d expecee e an ,sw release th t thesf eo e fission producto st be lowe thoria-basea r rfo d fuel than UO2. Jone. (1977al t se ) reporte releases ga w dslo for (Th,U)O2 fuels alsd oan , noted that fission-product release from defected thoria elements was one to two orders of magnitude lower than for UO2. Experimental data obtained by Matzke (1967) suppports this notion; he found that the release of Br, Cs and Rb from thori generalls awa y slower than from UO2. For the purposes of this report, we assume that fuel failures would be rare, and that the primary waste-form would be intact (Th,Pu)O2 fuel bundles. The above discussion suggests that thireasonabla s si e expectation. Obviously large-scaly an , f o e eus (Th,Pu)O2 fuels would be contingent on the demonstration of acceptable failure rates. Failure mechanism experimentar sfo l (Th,U)O2 fuel assemblies have been discussed briefly by Bain et al. (1977).

2 3. Oxygen Potentia Fission-Producd an l t Chemistry

Actinide fission changes the O/M ratio and hence the oxygen potential of oxide nuclear fuels. Fissiof no U causes only slight change oxygen si n potential, whereas Pu fission release significansa 235 t quantit f surpluyo s oxygen mainl e highee du ,th o yt r yiel noblf do e 239 metals (Kleykamp 1985). To assess the effect of fission on oxygen potential in 233 239 (Th,Pu)O2 fuel, some knowledg influence th f eo f eo welUs a s a l Pu fissios ni required. The fission-product spectrum for 233U is depleted in noble metals, as compared with Pu, therefor mitigaty ma oxyget e ei e th n surplus generatey db Pt ua extended burnups 239. 239

40 Changes in oxygen potential of the fuel can markedly affect the chemistry of some fission products, for example, molybdenum may be oxidized from the metallic state to M (Kleykamp 1988, 1990) higA . h oxygen potentia alsy o ma lincreas mobilite eth f yo one of the radionuclides of particular concern in nuclear fuel waste management (Johnson et al. 1994). Uranium dioxide has a high capacity to accommodate surplus oxygen as interstitial defects or defect clusters in the UO2 lattice (Willis 1987); the surplus oxyge compensates ni oxidatioy db uraniuf no mforo t m r U(VI U(Vo ) ) species. Pure thoria doe t havsno e this capacity, henc effece e th surpluf o t s oxygen no fission-product chemistr possiblr claddingye (o morth e b n yy o e pronounce)ma d thas ni case th e with UO2. Clearly evolutioe th , oxygef no n potential during operatiof no (Th,Pu)O2 fuels will be quite different from that in UO2 fuel, and this process and its possible consequences addressede neeb o dt .

33 Diffusion Properties of Thoria

In general, diffusion kinetic oxide th n esi matrix control fission-product segregation, except in the core of fuels operated at high power, where grain-growth occurs. Three distinct diffusion processes addressede neeb o dt . Oxygen self-diffusio mucs i n h faster than cation diffusioinvolves i t bu botn i nd^ honl ThOUO yd indirectly2an t alla f n ,i ,i fission-product release mechanisms. Diffusion of the principal cation (i.e., Th or U) governs grain growth, which in turn affects fission-product release at high power ratings. Finally, diffusion of the fission products themselves governs their release under most operating conditions.

Thoriu self-diffusion mio stoichiometrin i n c ThO slowes 2i r than oxygen diffusio fivy nb e or six orders of magnitude. The slow rate of this process makes experimental measurements difficult earld an ,y data show generally poor reproducibility (Shiba 1992).

More recent data indicate that the cation self-diffusion rate in ThO approaches that in 2 UO2 above 2000°C, but drops off somewhat more quickly at lower temperatures, because of a larger activation energy, 627 vs. 540 kJ/mol (Matzke 1987, 1992). Furthermore, a small degree of hyperstoichiometry increases thermal lattice diffusion of cations hi UO2 by several orders of magnitude because of its impact on the uranium vacancy population, whereas ThO2 practically always remains stoichiometric (Matzke 1987). Grain growth should thus be slower for ThO2 than UO2 fuels, and may be negligible, under the operating conditions normally encountere water-coolen di d reactors.

Diffusion of fission products in UO2 and ThO2 remains poorly understood, but generally appearvacanciesn io h involvo T st r o . eHigh-temperatureU , out-of-pile annealing experiments on lightly irradiated or ion-implanted samples appear to be consistent with modestly lower fission-product diffusion rates in ThO2 than UO2 - roughly paralleling the differenc cation i e n lattice diffusion (Matzke 1967, 1980, Akabor Fukudd an i a 1991, Prussin et al. 1988, Naik 1992). Fission-product migration in-reactor involves further complexity; indeed, Matzke (1980) has suggested that five different diffusion coefficients are required to model fission-gas transport! Nonetheless, the overall trend is evidently maintained: under equivalent operating conditions, fission-product segregation and release tend to be lower for ThO2 than UO2 fuels (Section 3.1).

4. FUEL FABRICATION

Much of the preceding discussion hinges on initial homogeneity of the fuel, which raises issue th fuef eo l fabrication fuee initialls th i l f I . y homogeneou moleculaa n so r scale, then its chemical properties will be governed by the chemically stable thoria. If the fuel

41 fabricates i blendiny db g thori plutonid aan a powders, however likels i t i ,contai o yt n microscopic plutonia-rich domains (grains or subgrains) within a nearly pure thoria matrix behavioue Th .plutoniue th fissios f r o it activatio d d mnan an n product they nsma longeo n controllee rb thoriae th y db , since they woul concentratee db microscopin di c domain verf so y high local burnup.

Microscopic inhomogeneity in (U,Pu)O2 and (Th,U)O2 fuels prepared using powder-mixing technolog bees yha n show causo nt e significantly enhanced fission-gas release (Cadelli and Lippens 1984, Ishida and Korei 1994, Smith et al. 1985), and similar results coul expectee db d with (Th,Pu)O2. Preferential leachin othef go r fission products fro high-burnupe mth , Pu-rich grain sucn si h fuels would als expectede ob . Preferential dissolutio alsplutoniuf s ni o opossibilita u P ms a y under reducing conditions. Therefore likels i t i , y that inappropriate fabrication processes would negat benefite eth s that arise fro desirable mth e chemica physicad an l l propertie thoria f so a fuel matrix. Small-scale tests with (Th,U)O fuel indicate that sol-gel fabrication procedures yielda

more homogeneous product tha2 n powder-mixing techniques, and its irradiation performance is promising (Hastings et al. 1982, Onofrei 1986). Further work is clearly neede verifo dt y that suitably homogeneou fabricatee sb fuen fula ca l n l do industria l scale.

Reactor operation also affects fission-product releaselineae th t rA .powe r ratings typical

of CANDU reactors grain growt naturan hi fue slighs UO li l t similaA t r power, 2 (Th,Pu)O2 fuels should exhibit graio littln r neo growth; however, these fuely ma s operat t higheea r power densities than natural UO2. CANDU reactors have flexibilitn yi

fuel managemen fued tan l design thaensurn ca t e that (Th,Pu)O fuels would operatt ea 2 simila lower ro r linear power ratings, compare curreno d t r example fue2 Fo lUO t e th , 43-eiement CANFLEX bundle reduces peak ratings by about 20% compared to the 37-element bundle (Lane et al. 1993). Furthermore, changes can be made to the internal element desig loweo nt r fuel-operating temperatures directly. Example sucf so h change graphite sar e disks between pellets annular o , r fuel bothr o , . Hence thera s ei real possibility that (Th,Pu)O fuel coul operatee db CANDn di U reactors with minimal

gas release. 2

5. OTHER ISSUES

A number of additional issues that would need attention in developing a plutonium-burnin fuel-disposad gan l schem addressee ear d briefly here.

In addition to fission products, some activation products are significant in the assessment

14 17 fueoUO f l disposal. Specific example 14 (formee C sar d from impuritieN n i d O san

l (fro l impurities) ^C 2 fuelme ^C d th an ) . Surprisingly bees ha nt i , shown nca thal ^C t make a significant contribution to the overall dose arising from UO fuel disposal

(Johnson et al., in preparation). Activation products will also need 2 to be considered carefully when assessing thoria-based fuelsr exampleFo . som,e b ther y e eincentivma e t stringense o t t specification chlorinn so nitroged ean n impurity level (Th,Pu)On si 2 fuel. Criticality considerations will influence the design of facilities for storing and handling irradiated (Th,Pu)O2 fuels. Clearly, thi alss si omajoa r consideratio fuer nfo l fabrication. The radiation characteristics of (Th,Pu)O2 fuels may differ significantly from typical CANDU fuels, and thus affect the design of handling facilities. Finally, although

the destruction of plutonium in (Th,Pu)O fuel is a net benefit from a proliferation 2

42 safeguards viewpoint, the fabrication and storage of this fuel will introduce distinct safeguards issues, as compared with conventional CANDU fuel.

. 6 CONCLUSIONS hige Th h degre chemicaf eo solubilitw llo stabilite th thorif yo d yan a make irradiated thoria-based fuels such as (Th,Pu)O2 attractive as waste forms for direct geological disposal. Furthermore, ther goos ei d reaso expeco nt t lower fission-gas releases (and correspondingly lower gap and grain-boundary inventories of other fission products) in thoria fuel wit2 s thahUO comparabln i e power history orden I . realizo rt e these beneficial qualitie thoria-basef so d fuels appropriatn a , e fuel-fabrication process muse b t utilized to achieve an acceptable degree of microscopic homogeneity. Detailed post-irradiation examinatio leachind nan g studie thoria-basef so d fuels, coupled witha thorough understanding of their physical and chemical properties, are needed to support these preliminary conclusions.

We thank P.O. Boczar . CarterTJ , . LemireRJ , . Matzke . VeedeHj ,J d helpfur an , rfo l comment drafa thif n so o t s report.

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Next page(s) Lef7 t4 blank RATIONAL Pu-TRANSMUTATION FOR

A. LECOCQ EURIWA, Orsay, France

K. FURUKAWA Institute of Research and Development, Tokai University, Kanagawa, Japan

Abstract

achievo T e 233y production from Thorium, with simultaneous destructiof o n Plutonium, nuclear specialists have proposed numerous ways of which the consequences may be very differen othee th r r fo tproblem peacefue th f so l nuclear utilization: Safety, Waste management, Industrial Deployment.... Rational choices made havb r attemptino efo t solvo gt e these difficultiess a o s , nuclear energ acceptablee yb . The Molten-Salt technology, applied for Fission reactors -supplying energy-, coupled

with Accelerators -waste destroyers & producers of fissile material (233(j) from Thorium-, allows to define a worldwide strategy, called THORIMS-NES, for the global supply of a clean and abundant energy. The non-aqueous reprocessing processes, used as a junction between the previous application Fissioe th thes d f devicesw so nan ene , could resor e presenbth t stockf so unreprocessed spent fuels recycliny b , g the mTHORIMS-NESn i .

1. INTRODUCTION

The world is changing: the end of the cold war, the economical crisis which concerns as welEast-Wese th l t relation North-Soute th s sa h ones exponentiae ,th l increas f populatioeo n which breeds economical and cultural poverty, the renew of conflicts for national and or religious incentives, are sufficient to convinc us that our human society suffers a deep mutation. The scientific and technical community cannot stay out of this crucial debate for the future of our civilization. Time is come for us -specialists of nuclear energy- to draw the lesson paste th f ,s o making benefit fro year0 m5 f nucleao s r energy researched an s industrial deployment.

positive Th e aspect f presenso t nuclear Fission Energy canno denied e shale b t e w s se l a , neee th laterf broa do t bu ,d changes urgemako t s seu rational choicen solvo sow t s eit problems. The GLOBAL SUPPLY OF ENERGY and PLUTONIUM DISPOSITION (or NUCLEAR PROLIFERATION REDUCTION firse th te understooprioritiess ar ) wa t organizere i th s a ,y db s of this meeting under the aegis of IAEA. Besides these first challenges, the nuclear communit alss fino yt ha d rational solution SAFETe th r sf fo nucleaY o r applicationse th r fo , WASTE MANAGEMENT INDUSTRIAe th r fo , L DEVELOPMEN worla t Ta d scale, without omitting eithe ECONOMIe rth C ASPECT prospecte th r FUTURe o , th r sfo E DEPLOYMEN varioue th f To s technologies.

2. THE TWO FIRST PRIORITIES

2.1. Global Energy Supply Afte year0 3 r f industrializatioso observey ma e nw : -very few small civilian power stations are deployed due to their high cost and by fear of proliferation. -only about ZTWe.year were produced sinc starte eth . -only small amount f spenso t fuereprocessede ar l , inducin deega p threa f dpollutioo r nfo the next centuries. -Thorium resources are neglected.

49 Thorium utilizatio producno t e energ consideres ywa initiae th n dli stag f nucleaeo r energy studies, because of the good neutronic properties of 233|j under thermal neutrons Thoriue th flux t Bu m. elemen fissilo n s e ha isotopet fissild an , e seed necessare sar staro yt t energy production. Beside soos nwa st discoverei d tha transmutatioe tth f Thoriuno m into 233u, produces

simultaneously 232(j -j-njs jsotope, parent of strong gamma emitters (^^Bi anc| ZOSji^ makes indispensabl remote-handline eth recyclgo t Thoriuew intne oa m matrix fissile th , e 233u obtaine reprocessiny db spene th f gto Thorium fuel.. Several projects using Thorium cycle (MSBR, HWSR, CANDU-Th, LWBR), were compare r theidfo r efficienc solvo yt energ e eth y supply proble mindustriao n [1] t bu , l application occurs, partly on account of the difficulties for 233y recycling.

But now new factors must be considered. The Greenhouse effect occurence, as explained in an other paper of this meeting [2], and various problems of the U-Pu fuel cycle (sect. 5.1), raise anew the advantages of Thorium which is 4 times as abundant as uranium, and can supply 100 times more energy. Besides the amount of Thorium waste (TRU) is several order f magnitudso e smaller than tha f Uraniuo t ] m[3

Since 10 years, several seminars and publications testify to the new interest for

Thorium use,[4-6] pollutioe Th . f recycleno d 233 232yy yb handican j wna s jcr hwa pfo Thorium use consideres i , usefus a prevenw o t l dno diversioe th t terroriso nt nuclead man r weapons production. [7]

2.2. Plutonium Dispositio r Proliferatio(o n n Reduction)

Two grades of Plutonium are generally defined according their isotopic composition:

Weapon Grade Plutonium (WGPu) or military Plutonium made of 239pu (>93% purity) and Reactor Plutonium (RPu) which is a mixture of the various isotopes produced in the usual power reactors distinctioe .Th n between bot rathehs i r specious sinc possibls i et i o et mak dirtea y weapon wit t h wili futureeasy e e d RPuth b l n an i ,produc,o t , e WGPu frou mRP by Laser enrichment.... Also the big amounts of WGPu (200 to 270 t) and of RPu (1000 t by the year 2000) have to be destroy. Moreover, this long half lived element has a high toxicity.

Many interesting projects have been published for fissioning, in power reactors, either WGP t wili see e s b la thi un i u [8-9]s RP IAE r o ,A meeting.

t firsA t sight t i seem, s rationaclassicae us o t l l power reactors with thermalized neutrons (or fast neutrons) to destroy Plutonium; but these old technologies perfectly mature for electricity production, might be adapted to a new fertile material: Thorium. Safet operationn yi , reprocessin spene th f go t solid fuel, recyclin f 233go y producee th dn i Thorium matrix, managemen wastee th f o t, hav econsideree alsb o t d fro starte mth . These new difficulties have to be overcome by a nuclear industry tightly connected, till now, to the U-Pu fuel cycle. Ooppositee nth , since more tha year0 n2 s ago, Molten-Salt reactors specialists have proposed to start the Thorium fuel cycle, by using Plutonium (WGPu or RPu) as a fissile.[10-11], and with a quite new technology grounded on a Fluoride liquid fuel, experimented successfully at ORNL[12].

Since the last 10 years, many improvements have been brought to the original Oak Ridge project (Molten-Salt Breeder Reactor :MSBR) which was focused on the breeding by the 232jh_233(j cycle. In an other paper presented [13], it is possible to appreciate the progresses done, since the first fission of Plutonium in the Thorium salt of the Experimental Reactor, MSRE, more than 25 years ago.

Beside Burner n i fissio e u sth P f nso Reactors mentiones a , d above developmene th , f o t e Molten-Salth t Plutoniue technologth r fo y m ywa destructio openw ne sa n with

50 Table 7: Global energy problems and achievable solutions by NEW FISSION TECHNOLOGY [THORIMS-NES]

TECHNOLOGICAL PROBLEMS -> ACHIEVABLE SOLUTIONS

RESOURCE [NEW FISSION TECHNOLOGY) U : localized, monopolized Th :non-local. . popular.

SAFETY 232 n — > 233U ' ecosystem (fissile) fossil ("thermal pollution high therm lencyc i f ,.ef '\ J ac id rai n no NOX , SOX 1 "Greenhouse" effects no COj nuclear /Trans-U [Pu. Ara| ,Cm o produn n o 1 c1 Rad-waste-1 : short : 5~ I 0 years power generation POWER STATION (•siting : restr icted nea o utilitt r y 1ishmenb ta s e f o t ^power size ' large sma ~ larg l1 e breeding cvc1e Iprocess-heat supply i ndust.. district

productio Spallationy nb . Than application a o kt f thino s process invente year4 [14]do 1 sag , e Molten-Salth t target release littla s e energy f neutrono t overal t bu ,lo a sl which transmute Thorium into

The technical connection between the burners reactors releasing thermal energy by Fission (F-plan),[13 e productioth d an ] f 233(no Acceleratory b j s (A-plan)[2], opens seducing prospects for the future uses of nuclear energy.

3. PRESENT INDUSTRIAL WORLD DEPLOYMENT

The undeniable penalties supported by the Nuclear Industry grounded on the Uranium Fission, result fro usee mth d fue appliee l cyclth d dean technolog shows ya Tabln no : e1 Fro U-Pe mth u fuel cycle: -by historical and technical necessity, the U-Pu cycle was the starting point of nuclear energy, but also of Proliferation. -less than 1% of natural Uranium is fissionned and produces energy, due to enrichment necessit losd f fissilspense yo an th en i t fuels. t breeds-i mais a , n waste, Fission Products (FP) heavd ,an y elements (TRU) with long half-lives and high toxicity. Secondary technological waste appear also in the reprocessing Purex plants, (sect. 5.1.) Fro appliee mth d technolog Fissioe th yn i n Industry: -low thermal efficiency, uncertain safety. -need f enricheso d Uraniu nuclea e r mosmth fo f o t r reactor types; -weak flexibility in load operation, very limited in size so that the always-large NPP are unexportable in small and poor nations. -reprocessin spene th f gto fuel f any)s(i , inefficien wastr fo t e managementr welfo s s a , a l the recovery of unused fissile materials, (sect. 5.2.)

51 . PROPOSAL4 RATIONAF SO L IMPROVEMENTS

4.1. Safet f Nucleao y r Reactors Although ther generaa s ei l consensu thay safete sa th t f o st existino y s i P gNP satisfactory, we know that it is based on stochastic risk analysis, and is relying on engineering safety facilities (ECCS, RHR, ACCS, etc) These devices aim at correcting the weak points of the fission technology which are the pressure of the coolant, the excess of reactivity of the fuel, the Zirconium cladding chemically reactive with steam, the internal storag f PP...eo . r searcInou f rationalityho have w ,compar o et e these solid fuels reactors whice har improved day by day, -with the future hope to become "Inherently Safe"-, to some other proposals with liquid fuels which rel naturan yo l scientific laws

Table 2 Natural elements and practical separable isotope havmy tiny thermal cross-section datw ane s (Igarashbasee th n do Furukawa& i , 1990)

vnatura! c-oss-section abundance; m barn (natural eletrent)

0 '9

ID] (0 0148K) 0 5i9 t -i 332 S) 3 53

{80 08s) 7 COO75 )3 (- rie

Be 5 7

7 9 6 8 33 8

9 |0 39

10 {92 5*1 45 4 (3Li 70 500)

1 1 ]2Mg 63 18 37Rb 350 12 14Sl 17 IS 20Ca 430 13 82Pb 171 20 I6S 530 14 172 21 ,,Na 530 5 15 ' Zr 185 22 50Sn 626 IS 231 1 17 '^H 332 6 57 7C1 33500

Molten-Salt reactors described in this meeting [13], used liquid Fluondes mixtures of high thermal capacity, chemically inert to air and water, without pressure The operating conditions allow to establish a walk-away safety due to a weak excess of reactivity, a negative temperature coefficient and a small residual heat. The engineered safety facilities consist onl statin yo c equipment (dram tank). Furthermore, such reactor designeo s e sar d that therreleaso int environmenn e P oes F th i f eo evenaccident.(ne n a th f n i to o storagf eo Xenon, Krypton and Tritium in the fuel, no escaping ability for Iodine and Cesium ions which are trappe ionie th cn d i medium)[ ] 8 1 15- Molten-Salt Accelerator ] usin[2 s g similar fluondes mixtures witsame th h e macroscopic features, hav same eth e chemica physicad an l l behaviour Besides the safee yar r by their subcntical operation

4.2. General Waste Management. Nuclear wast producee ear treatmene d or fro e mth t eventuae tilth l l reprocessinf go spene th t fuels, passin enrichmenty gb , assemblies fabricatio electricitd nan y production. The more yar r les eo s dangerous according their concentration, radioactivity, half-life period kine f emitteth ,d o d theirayd san r abilit enteo yt livine rth g matter Till nowspene fate th f th ,eto fuels mos,e whicth e t hdangerousar eithes e b wa ,o t r reprocessed to recover Plutonium and isolate the high-level radiowaste for storage after vitrification coolee b o t d r enougo , h before burying them underground

52 233 239 UF4 and/or PuF4 [Breedin Chemicag& l Processing Center] i Solid-fuel Compounds \[ | AMSB | _ |AMSB| _ JAMSB] _ '• 1 _ j_ !____ Production Plant ( 4-30 t breeders) i i Solid-fuel Assembly Xv target Fablic. Plant salt* ! X\ fertile salt** / <— — 7LiF 2 • «UFX N 4 < < Chem Proces. s Plant). — -jchem Proces. sPlant. ) t plants2 ( t ) vU-——^ (P.P.) . 1 1 ^ 1 } Radio-waste 1 plant( Plant )j ] 1 i 1 i / * ' i t t ^ / < 1 heavily safeguarded

supplyingtf dirty 'r ^ fuel salt fuel salt Solid Fuel Cycle System i i very —— 1 — i i ^ _ *-„*-., i small small small large i ' i [0.1-1 TWe)

7 233 MSR MSR MSR MSR ! ! ! J (*) LiF-BeF2-ThF4- UF4 [mini [FUDJ [FUJI] FUJI ] ' 7 : Î ~ (**) LiF-BeF2-ThF4 i i 233 1 Î (if) target salt* + additive UF4 ^--i, V 1 . _ ' . . . .. i

Simple Molten-Salt Fuel Cycle

Fig 1. Molten-salt breeding fuel-cycle in THORIMS-NES co'mposed of Lft fissile-producers (MSB powed -an r stations (MSR) But public who disagree partly with both solutions, welcome the new proposals to transmute FP and to destroy TRU into less dangerous materials, by high neutrons flux irradiation or by spallation.

MSR proponents have preceded this historical change since ORNL peopl studied eha d dan tested various method proceso st linFluoridn e so eth e fueo react ls salta breedinh a o s , g rati f 1.06oo .

4.3. Reprocessin n i THORIMS-NEg S At the time of ORNL works, the target was to reach a breeding ratio as high as possible, by using materials with very small cross-sections (Table 2), and pulling out of the neutrons thermal flux, by physical or chemical processes, the high cross-section products 14 147 1 5 2 ( 9Sm, TSSxe, 143^, PmSm;..) alsd 1 , oan , 33p parene a,th 233uf o t o avoit f d the formation of 234pa The whole process, described 25 years ago [19], was achieved in the Fluoride medium.

Nowaday proves i t si d [20] tha faisible th t e doubling yeartim0 2 t r MSBa es i sfo d Ran least 30 years for FBR. These values are too high for bringing a right answer to the fissile neednexe th f t so century whicn i , doubline hth g time shoul lese db s tha years0 n1 . Alss i t oi more rationa economicad an l havo t l e good converters, eas operateo yt , withou chemicay an t l plan havo t td e[13] alsan , o Accelerators abl produco et amountg ebi f fissilso e 233^ witha satisfying doubling time. [2]

But reprocessing has to be done from time to time applying mainly the non-aqueous (or Dry) chemical process experimented by ORNL. At the exception of gaseous FP (Xe, Kr, T), the treatment is done by batches of Fluorides salt, either leaving the burners for

extraction of various FP and addition of new fissile (233j j Plutonium), or coming from r o

Spallators to take out the Spallation Products (SP) and the production of 233yj w\^ simultaneous addition of Plutonium. That is the 1st part of our "D-plan" which is the link, through the chemical reprocessing plants, between Burner Acceierators.(Fig.ld san )

4.4. Significant features of waste with THORIMS-NES Amount of Radio-waste, for a supply of 900 TWe.year: [21 ] -Ores treatment: the global demand of Thorium, is surprisingly small and ~2 Mt only for the world in which ~35 % will be transmuted to fission and spallation products. -Spent fuel reprocessing coul totaP e decreaseF e :th d f b o l d abou fro% m30 t tha f LWo t Ry b the improvement of the thermal efficiency, the produced transuranium elements (TRU) will be two or three ordre of magnitude less than the amounts in U-Pu fuel cycle , and essentially they are staying in fuel-salt as a fissile or fertile species. -Today low-level radio-wast producee e ar varioue th t da s fuee stageth lf processeso s : chemica mechanicad an l l fabricatio fuelaftee d th an ,f r n o irradiation during decladding, chemical reprocessing refabricatiod fuew an , lne assemblies e th f no huge .Th e amounf o t these wastes shoul minimizee db ordee th f 1/10dn o ri 0 processe y dr wit e hth s appliedo t THORIMS-NES. -Structure material mad f Hastelloeo wilyN l hav inducen ea d radioactivit 7 mrem/hr0. f yo , after 1 year coming mostly from 94Nb; It might be fully reused after vacuum remelting. -Graphite moderator kept in place for the whole reactor life could be partly reused after peelin deptn i g m takabouho t m e5 awat0. radie yth o deposits.

Radio-waste Elimination: Fissions-Products of half-lives longer than 30 years, which are coming from own MS fission reactors and from other reactors, might be incinerated in THORIMS-NES. [22] -eithe forn i r f moltemo n Fluorides: "" 29 I5 Csan3 , 1 MSy d51! b SAMSBr Rm90 o d Sran ,,

93Zr, 126Snan 7 CAMSB3 dy 1 s b ; -or in metallic form casing in graphite, for example: 59^ 63^ 94^b andTc99 by MSR AMSBy b d P . 7 o(thes0 r AMS1 d e Bspeciean s migh recyclede b t )

54 TR valuable Uar e fissil r fertileo e isotopes should an ,effectivel e db y utilize fissios da n reactor fuels not reproducing them. For this purpose MSR will be the best facility, in

practise, we are examining the FUJI IV, modifying the composition of FUJI II, in which about f 100-20 o U fertilreplaced s i U 233 an f e; TR o y ThF0Kg 3 db 1/ decreases 4i d fro2 m1 mole% to 8 mole% resulting in the fissile, which is supported by the further addition of TRU in burning rate of about 50 kg/year.

5. FUTURFROE PASE TH MTH O T ET

An other rationa fino best e ds th i l tchoic, meando r passino et s fo g fro presene mth t nuclear industry, which should burn solid fuels with mor r leseo s Thorium matrix, towards MS application THORIMS-NEn si ; 13]S[2 . The junction between both fields betweey sa , thao t ns i t "PAST "FUTURE"d an " e th s i , reprocessing of the spent solid fuels.

5.1. Purex process The present debate between OPENED FUEL CYCLE (OFC) and CLOSED FUEL CYCLE (CFC) result from the incapability of the Purex process to treat the whole spent fuel production. Practically, this technic is convenient only for ores treatment (U or Th), and for military Pu extraction from slightly irradiated fuels.

The weak points of Purex process are as follows: -Light water, use solvents da , induces criticality hazards which preven havo t t e vessels usuas i bige biget d i ls an r(fo a r r economy chemican i ) l industry. -Nitric acid and TBP molecules , sensitive to the various rays of the spent fuels, are partly radiolysed givd ean , by-products which distur basie bth c proces destroye neee d b san do t d or eliminate radio-wastes da . -With this process, all radio-waste are diluted in enormous liquid volumes flowing in numberless pipes and sub-critical apparatus.... for finally being re-concentrated in volumes as small as possible. It results a large production of contaminated materials and many radioactive solutions to manage. fuee -Whateveth l e burn-upb y same ma r th , e proces applieds i s opposition i , e th o nt general chemical rule whic chooso t hs i suitablea e proces eacr sfo h composition range.

5.2. DRY processes This proposal to use a non-aqueous process, and chemical mineral products stable under atomic rays ,e formeavoith l al dr difficulties: criticality hazards, radiolyse, dilution...etc These DRY processe lyine welsn ar g o l known law f thermso electro-chemistryo& d an , usnumeroue eth s data about oxides, chlorides, fluorides compound elemente th l al r f sfo o the Mendeleieff Table. (ANL-5750, JUL-565-CT, IAEA-STI/PUB/424, ..etc)

Their main features are as follows: -Medium whicn si h these processe performee sar d -solid, salt r gaseouyo s phases- havo en moderators. This allows operation with fissil d radioactivan e e material f higo s h concentration. -The stabilit f reagentyo s makes these methods highly productive, compact alsd oan , forms small quantities of waste. -Dry fuel reprocessing methods (Fluorides volatilization, pyro-electrochemical or irradiated metal refining) allo wo treat burn-uy t fuelan coolin d f so p an g time, witha small amount of technologic stages.

The Fluorides volatilization process studie experimented dan r U02-Pu0dfo 2 fuely sb french-russian teams [23] allows the recovering of Uranium with a decontamination factor of 10E6. An experiment showed that this method would be sufficient for Uranium extraction from U-Th fuel. [24] The Fluorides salt of MSRs and AMSBs are tightly connected with the volatilization process; 1 5 years ago, neutronic calculations have shown that the LWR spent fuel can be re-

55 used (after a coarse treatment by Fluorine) as fuel for MSRs; in that application, MSRs can be the bins of LWRs. [25]

r D-plaThusou f THORIMS-NESn ni o , Fluoridee th , s volatilizatioy n ke proces e th s si to obtain the higher efficiency from the present nuclear fission industry. It allows: -the eliminatiotoxicitw lo e yth wastf no e -the recycling of the other PF, Actinides, TRU, to the Burners (F-plan) or to the Accelerators (A-pIan),according their neutronics properties e Globath d l an ,energ y demand.

. RATIONA6 L PROSPECTS

considers u t Le , lastly particulao tw , r prospect r Nucleasfo r Energy: Nuclear Fission Industry shoul usee db d everywher worlde th chiefld n ei an , n i y small and poor nations where there is a deep lack of clean and cheap energy. To reach that goal there is a huge need of small and economical power plants. Till now, almos nucleae th t r fission energ produces ywa r electricadfo l purposest Bu . electricity cannot be stored, which is a high penalty for solid fuels reactors of weak flexibility in power production. Chemical applications of nuclear heat should fill that lack. by increasing the enthalpic content of the products involved in the chemical reactions.

6.1. Small economical Reactors Nuclear island is only a part of electricity plant, so that the price of energy supplied, for all facilities using solid fuels, is rather indépendant of the nuclear type. But the cost of small power stations is biger because it is necessary, for each size, to make specific applied technologic researches about: -definitio speciad nan l fabricatio r eacnfo e hnuclea th par f o t r cell: reactor vessel, exchangers, pumps... -calculations, fabrication and tests for fuels, cladding, assemblies.... -studies and fabrication of internal core structur, control rods, loading machine, safety devices... -rules for normal operation, loading and unloading of fuel, emergency conditions... -spent fuel storage management.... Thus, small nuclear reactors, issue from the old technologies, are as complex as the big ones, and they are not more flexible; they are freezed for fuel composition, nature of production (which is generally the unstorable electricity) and the range of power supplied.,

Small MSR hav same eth e economical advantages already describe dbigee herth r refo ones, suc operations ha , maintenance work, fuel management without fuel handing machine, .etc.[2;13] t witBu .h MSR nucleae th , r island shoul rathee db r independene size th f th e o f to plant: -no specific fabrication and tests of fuel according the size -no specific and complex safety devices In addition to these features, comes the specific quality of large MSR which is the flexibility in power from 10 to 100 %., according the hours, seasons, etc.. So there is no nee conceivo dt e several nuclear reactor variouf so s size answeo st varioue th r s demands. With a convenient overlapping, two sizes should be theoretically sufficient. For example: -The Small size, like FUJI I (400 MWth nominal) running from 40 to 400 MWth. -The Big size, like FUJI II (2200 MWth nominal) operating from 220 to 2200 MWth. So, it appears that Small MSRs [26-28] could be the energetic tools, escorting the spontaneous progres Developine th f so g Nations.

6.2. Chemical heat Applications Nuclear hea vers i t y little utilize r chemicadfo l applications maie th ; n reaso thas ni t the highest temperature in all classical LWR is not sufficent to perform chemical reactions.

With MSRs which operat t higheea r temperature (53 o 700°C0t e thermath ) l conditions are much better. By the high temperature of the fuel salt (700°C and more)

56 easily reached with the alloy Hastelloy N, MSRs can supply district and industrial heat (oil deep refining plants, synthetic fuel production...), in better conditions than HTGR, because of hige th h thermal capacit Fluoride th f yo e salt, which avoid largo to esa dro f temperaturpo e during the heat exchange process. More interesting for the environment, is the possibility to produce hydrogen by reaction between natura (CH4 s steamd ga l an ) .reactions o [29Tw ] , called "the methane reforming", and the "Shift reaction" used straight the heat of the salt without passing by electricity productio wated nan r electrolysis. For higher temperatures, easily accepted by the salt, it is proposed, in the future, to typeusw ene f graphits o e (like glass-fiber carbon) realizo t d an e, heat transfer forn i t f mo radiant heat exchange overcomo T . hige eth h temperature pump problem suggestes i t i , do t us naturaea l convection regime,and/or gas-lifting.[30]

7 CONCLUSIONS

Afte year0 3 r f theoreticaso applied an l d researc connectiohn i n wit hstrona g effort for the industrial development, which have shown the advantages and the limits of the nuclear fission energy, the time is come to give a new impulse by rational choices, to increas potentialitiess it f o e acceptanc s botus it d e hpublicth ,an e th y eb . Some progresses shoul obtainee db thadn i t purpos shiftine th y eb g fro U-Pe mth u cycle to the Th cycle, with simultaneous burning of the big Plutonium stocks. However it is possible to go further with deep improvements in the other fields concerned by the acceptability of Fission Energy: Safety, Waste management, Proliferation, Chemical applications..., by rational issues about the technologies to use for its future industrial deployment. The applications of Molten-Salt technology, well known by the international nuclear community, can be this essential mutation, by offering a Global strategy for the supply of a clean Energy amountsg bi n i , , everywher lona r Worldge fo timeth d en i an , . The main rational choice for International deciders is to find the right balance between the weight of the Past and the necessity to prepare the Future through an International cooperation..

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[23] PEK . A"FLuoridI e Method r Irradiatesfo d Fuel Reprocessing". Pape 83/1L M r . Fifth COMOCON Symp. Invest, in the Field of Fuel Reprocessing and Radio Waste Disposal. Mariànské Làzne, Czechoslovakia April 1981.

[24] SKIBA O.V, A.A.MAERSHIN, P.T.PORODNOV, A.V.BYCHKOV. "Nuclear fuel cycle basen do dry method f fueso l processin automated gan d processe fuef so l element production." Proc. Int. Conf. Global-93, Seatle Sept. 1993

58 [25] ENGEL J.R, RHOADES W.A. GRIMES W.R. DEARING J.F. "Molten Salt Reactor for efficient nuclear fuel utilization without Plutonium separation" Nucl. Techn. 46 (1979) p30-43 (ORNL-TM 6413)

[26] FURUKAW , LECOCAK. . PreleminarQA y "Nexe Examth n to , Generation Nucl. Reactors" in comparison wit Smale hth l Thorium Molten-Salt Reactor. (Toka 988i1 Univc )De .

[27] FURUKAW , MINAMAK I K;,OOSAWA T.,OHT , MITACHAM. KAT& . K IO "Design studf yo small Molten-Salt Fission Power Station suitabl r couplinfo e g with Accelerator Molten-Salt Breeders" Emerging Nucl . SystemsEn . , p235-239 (1987) Worl, dSc Pub.

[28] FURUKAW , MITACHAK. KAT& . . K "SmalIO Y l Molten-Salt Reactors wit hRationaa l Thorium Fuel-Cycle" Nucl. Engin, and Desgn, 139, pi 57-165 (1992)

[29] LECOCQ A., FURUKAWA K. "Hydrogen Production for the Next Century" Energy and Environment into 1990s 1st World Renew. Ener. Cong., Vol.5 p2949-2954. Perg.Press

[30] ABALI t al."Phys.peculiae S NS. f higo r h temp.MS free th e Rn i convect . regime wite hth us f gaslifteo " Atomkern. Kerntech. Vol.38 (1981) p21-23

Next page(s) Left blank 59 SESSION 2: GAS COOLED REACTORS AND THORIUM ASPECTS PLUTONIUM DESTRUCTION WITH PEBBLE BED TYPE HTGRs USIN BURNEu GP R BALL BREEDED SAN R BALLS

K. YAMASHITA, K. TOKUHARA, R. SHINDO Departmen HTTf o t R Project, Japan Atomic Energy Research Institute, Japan

Abstract

It was made clear that; pebble bed type HTGRs using Pu burner ball u balls(P sd breedean ) r ball h balls(T s ) possessea s potentia o burt l n weapons-grad 0 Gwd/T74 e o totat puTh . u lP e amounts of Pu and 239Pu of can reduced to about U20 and 1 %, respectively.

1. Introduction Ther some ear e problem nucleae solvee b th o n st i d r design when plutonium is burned in reactors. Power coefficients may become positive due to giant fissile resonance peak of Pu in thermal energy range. Abrupt power excursiono t occuy e sma rdu small delayed neutron rate ß of Pu when a positive reactivity is adde o t reactord s n unanticipatea e.g y .b d controd ro l withdrawal mos e pebble th .tTh typd f promisino e be ee HTGon s Ri g reactors that can solve the problems. The first problem will be avoide y mixinb d u witP g h fertile materia. U lr o lik h T e Secondary, the reactivity addition by unanticipated control rod withdrawal will hardly occuroperatioe th n si n condition because the excess reactivite t necessarth r burnu no fo yo t s i pe du y continuous fuel loading, and the reactor is operated in the condition wher econtroe mosth f to l rod fulle sar y withdrawn[l]. Furthermore, the fuel temperature does not rise abruptly due to the large heat capacit graphitf yo corn ei e positiva eve f ni e reactivit addes yi somy b d e accidents. However, a problem remains that fissile materials transmuted from fertile material se spen existh tn i tfue l eve mosf ni f o t initial Pu is burned up. The residual fissile material made difficul o dispost t e spenth e t fuel directly. Otherwisee th , reprocessin spene th tf g o fue necessars l i effectiv e th r yfo e eus of the fissile materials. Thus e havw , e foun a reactod r e b syste n n whicca i m u P h burned out so that the reprocessing of spent fuels is no more necessary. This syste pebbla s typmd i ebe e HTGR usin burneu gP r balls (Pu balls) and breeder balls (Th balls). The Pu ball contains plutoniu mixet mno d with fertilh T ee materiaTh . Th f lo ball contains only fertil e cor Th s loademateria i e. Th d f o l with thes o type tw ef fue o s l balls randomly. Thee removear y d continuously frocor e reloaded mth e an e th d r int core fa oth o e s burnup achieves si e expecteth o dt d values. This e systeb n ca m applied for burning of Pu obtained from nuclear weapons destructed or spent fuels of LWR etc. This paper presents the principle of the reactor system and the burnup characteristics when weapons-grade Pu (94 % 239Pu and 6 % 24°Pu) is burned in the reactor system.

63 üa New burner balls

Breeder bails Burner balls

Burnup measurement

Burner balls

Breeder balis

Fig Flo1 . f breedewo r ballburned san r bali corn si e

2. Movements of balls in core h ballT e showu balle movement P ar sd Th n e i nan s th f o s Fig.l multi-pase Th . s fuel loadin e reactos appliei gth r rfo d system so that most of Pu in the Pu ball will be burned up ultimately. The burnups of the Pu balls are evaluated by measurin e strengtth g f y-layho s after thee exhaustear y d from the core. The Pu balls will be reloaded again if the burnup is less thaexpectee nth d burnup ballu P e s. Th wil dischargee lb d from the reactor system as spent fuels if the burnup reaches to the expected value e excesTh . s reactivity necessare th r fo y reactor operation is maintained by loading fresh Pu balls from ballh coree T reloade e th e tosar .f Th po d repeatedls a r fa o ys fuee th l integrit maintaineds yi multi-pastechnie e .th Th r cfo s fuel loading has been accumulated through the operation of the pebble bed type experimental HTGR (AVR) in Germany for about 20 years. This technic can be applied for the reactor system[2]. Destructio. 2 u P f no e achievablTh e maximu bal u mP evaluates lwa e burnu th f po d witfollowine th h g relation[3]:

64 Tabl SpecificationeI cor f sfued o ean l Thermal power* 200 MW Core height* m 4 9. Core diameter* m 0 3. Moderator Graphite Coolant Helium Pu ball Pu amount 5 g/bal0. l Number in core about 180000 balh T l Th amount g/bal4 1 l Number in core about 180000 Specifications referred to Russian HTR-Modul developmenn i e (5) t

where burnu: B f spenpo t fue f reactorlo whicn si whole hth e fuels are replaced with n-time reloadinf so g (GWd/T) B1: burnu spenf po t fue reactorf lo whicn si whol e hth e fuels replacee ar oncr dfo e (GWd/T). e pebble burnuth Th d typf be eo pe HTGR (B s ntwici ) e tha1 B n because the n can be regarded as the infinite due to the continuous loading. B was calculated with HTGR fuel lattice

burnup calculation cod1 e DELIGHT[5] specificatione .Th Tabln si e werI evaluatioe eth user fo d burnupf no .simplifieA d modes li used for the evaluation of the burnup. Pu, Th and U are distribute e fueth l n e coatei kernedth f do l fuel particles (CFPs) in the ball fuel in the evaluation model. The effective cros sevaluates sectio wa resonance h T th y f ndb o e calculation. The effective changth n i e e multiplication factor kgives efi f n

Fig.2 n showi s i Fign t n.i I tha.2 ^ becometB GWd/t0 e s37 Th . Pu

1.5-

Max. achievable burnup 740GWd/tPu

' s ' "- 1

0.5

0 500 1000 Burnup(GWd/tPu) Fig Chang2 . effectiven i e multiplication factor with burnup of Pu ball

65 1000

sTotal Pu Pu-239' TO 500 Max. achievable D. burnup

s740GWdApu

Pu-24Q_ ——— ~v^

. Pu-242_ - >V ---> "'-•- * 0 300 600 900

Burnup (GWd/TPu) FigChang3 . amounten i isotpeu P f so s with burnu balu P lf po (Initial amount of Pu is normalized to 1000kg)

achievable maximum burnup is 740 GWd/tpu. This is obtained by substituting 370 Gwd/tpu into B1 in the equation ( 1 ). The change u iisotopeP n s with burnu s show i pe amoun n FigTh i n f . o t3 . total Pu for 740 GWd/tp is 20 % of the initial value. The amount of 239 Pu becomes lesu s than 1 % of the initial values. The Departmen f Energo t n USi y A (USDOE s reporteha ) d study u dispositioP n o e th advance r nfo d LMR (ALMR d advancean ) d BWR (ABWR)[6], Tabl I I showe s residual amount isotopeu P f o s s when e weapons-gradth f 100s o burnei u g 0eP k d wit e ALMRhth , ABWd an R e reactoth r system conceived here u fueP .l without fertile materials was burned in case of the ALMR. Pu was mixed with natural uranium in the case of the ABWR. The residual amount of

Table II Residual plutonium and generated energy by loading of 100 plutoniug 0k n mdiffereni t reactor types

(Before After burnup [burnup This HTGR ALMR* ABWR* i Total Pu(kg) 1000 199 618 648 239Pu(kg) 540 6 '476 279 240Pu(kg) 60 72 1 130 233 241Pu(kg> 0 55 12 97 242Pu(Xg) 0 66 0 39 Generated i 1A- * n ** l v 1 n 1i v i o ' n i AY

* .'They were referred to "plutonium disposition study" in USDOE^'. ** rEnergy generated by burnup of fissile materials converted from Th and 238U was included. ***:Enrichroent was regarded as 3.5w/o. This was evaluated from burnup of 37.1GWd/t given in reference1".

66 e reactoth r totarfo u systelP abous i m t one-thir ALMe th Rf o d d ABWR an residuae .Th mucs i l hu amoun P les f sto than ALMd Ran ABWR generatee .Th d energ reactoloadine y yb th r 100f grfo o 0g k system is lager than those of the ALMR an ABWR. the others. It results from burnu 233f po U converted fro Fro. resultse Th mth m , it can be concluded that the quality of the weapons-grade Pu can e stronglb e b y n degrade e reactoca th u P y b d r e systeth d an m converted to energy effectively. The amount of 233U in Th ball will be saturated to 0.2 g after a certain duration in the core. The burnup of the Th ball is about 7 GWd /tTh for a year. The Th ball can be used for 16 years if it can be supposed that the integrity is maintained to e averagth e burnu f Thoriumo p , Hochtemperatur-Reaktor(THTR- 300)(109 GWd/t)[7]. . Maximu3 balu P mln i amoun u P f to Powe e fresu s largP th i r u ball e P hn e powe i eth Th .f o r ball decrease with progress in burnup. The maximum amount of Pu baliu nP limite s maximule i th y db m permissible powe balla n ri . The power of 4.5 kw was used in the design of HTR module as the maximum permissible. This was determined by the maximum permissible thermal stress in the ball. Figure 4 shows the change powebale u iP nth lf r o wit h burnup dayh T sbal u P d whe le an nth ball are loaded with 0.5 g Pu and 14 g Th, respectively . Axial power peaking is not considered in power change in Fig. 4. The maximum power appear beginnint a s burnupf o g reae .Th l maximum core poweThiabou . s obtaineth s i ekW n s wa ri t4 usiny db e gth axial power peaking of the HTR module (1.7)[8]. The real maximum powe lowes ri r tha maximue nth m permissiblt I . kW e 5 powe4. f ro can be concluded from the results that the optimum amount of Pu in the Pu ball is 0.5 g when the core is loaded with the same h bal T h ballsnumbes T u i ballle P e powe d th f Th .an so rf o r almost constant afte saturatioe rth concentrationU f no . About reactoburnee e b th 6600 n y db ca ballsr u 0P , Pu ,f o namel g k 3 y3 system specified in Table I for a year.

—Max, permis 4?Qw_ejjof_a fuel .ball

CO =94

CO .Q "05

co 2 ^Burner ball

CD o Q_ Breeder ball 500 1000 Burnu (daysy pda ) Fig Chang4 . bal u poween P i l wita f o rh burnup

67 4. Conclusion

s madIi t e cleae rb tha0 GWd/te n burnu74 th ca t f o p u

achieve usiny db pebble gth typ d ebe e HTGR ballu usinP d p e san gth Th balls. e th e reduce b f Th o n e o abou% ca t d amoun0 u 2 P t f o t initial value finae .Th l total amoun fissilf to e materiale th n si

spent Pu ball will be less than 1 % becausPue will decay to Am. The substantive amount of fissile materia241 l is a small amount of 239Pu in the Pu ball. It could be concluded from very low o morn qualites i u thaworth P e t i tth f o yf reusinr o y fo t i g weapo reprocessind nan fuew ne l r productiongfo . Thereforee ,th spene disposee b rocu ballth P tn k n ca si dsal t mining directly[9]. Energy generate reactoe th y db r systee sam th mor s emr i efo Pu amount than those of the ALMR and ABWR. The irradiation tes coatef to d fuel particle (Pu-CFPu P f so ) had been done by in Belgonucleaire in 1974. It was shown the integrit Pu-CFe th s stil f Pwa yo l maintaine 0 burnur 60 fo d f o p GWd/T[10] densite e fueTh .th l f kerneo y s loweree lwa th n i d

CFP tpu o avoid the pressure rise by fission products. Further reduction in the fuel kernel density may be necessary to realize the burnup of 740 Gwd/tpu. amoune destructeu P Th f e reactoo t th n i dr syste onls i m y 33 kg for a year. It is small against total amount of Pu produced destructioe byth nucleaf no r weapons smale reasoe th .Th lr nfo amounu destructiblP f to e reactoburnes i thas th i e u n P ti d r system effectively. However, the problem can be solved by the spiking of Pu balls. The Pu balls are made radioactive by loading in the reactor core so that the diversion of Pu will be made difficult. Large made numbeb ballu P en f sca radioactivo r n i e the reactor system becaus ballu eP continuousls i s y irradiated in operatio e pebblth reactord n ni ebe e irradiate.Th balu P d l will be stored temporary and be burned again in the reactor system.

ACKNOWLEDGMENTS The authors express their sincere thank o Messrst s . I.Murata, A.Saikusa, S.Shiozawa, T.Tanak d T.Tobiokan a r fo a valuable comments and support.

REFERENCES [1] HUEBEL,H.J.,LORNERT,G.H.,:Kerntechnik,47, (1985) 159pp. ] IAEA[2 : "Operation Experience with Nuclear Power Station Memben si r States",(1974-89) [3] GRAVES,H.W., "Nuclear Fuel Management",(1979). GREBENNIK,V.N.,MOSEVITSKII,I.S.:JAERI-M92-215] [4 , (1993) ] SHINDO,R.,YAMASHITA,K.,an[5 d MURATA,I.:JAERI-M 90-048,(1990),(in Japanese). [6] USDOE:"Plutonium Disposition Study"(1993). Assoziatio] [7 n EURATOM/BBC-Krupp /KFA:"Sicherheits bericht des THTR Prototyp 300 MWe",(1992). [8] TEUCHERT,E.,RUETTEN,H.J., and HAAS,K.A.:Juel- 2618,(1992). [9] KIRCH,K.,et al.:Nucl.Eng.Des.,121, (1990), 241pp. [10] BAIROIT,H.,et al.:Nucl.Technol.,23,240,(1974).

68 THE PLUTONIUM CONSUMPTION MODULAR HELIUM REACTOR (PC-MHR)

D. ALBERSTEIN, A.M. BAXTER, W.A. SIMON General Atomics, San Diego, California, USA

Abstract

The Plutonium Consumption Modular Helium Reactor (PC-MHR) can effectively destroy excess weapons grade plutonium, consuming more than 90% of the initially charged plutonium-239 and 65% of the initial total plutonium in a once through cycle - high plutonium destruction without recycle. Because the PC-MHR uses no fertile fuel, no new plutonium is created. The spent fuel discharged from the PC-MHR has characteristics that giv et unusualli y high resistanc diversioeo t proliferationd nan e Th . amount of plutonium per spent fuel element is very low; the spent fuel element mass and volume are high; there is no developed process for separating the residual plutonium from the spent fuel; and the discharged plutonium isotopic mixture is severely degraded - well beyond typical "reactor grade" plutonium. The discharged fuel element suitablee ar s , wit o furthehn r processing r direcfo , t disposaa n i l permanent geologic repository PC-MHe unique Th .th s eRha safety characteristicf so safetR othedesignsR MH y e MH rdesig Th . n objective t througme e ar sh inherent safety feature desigd san n selections that take maximum advantag f theseo e features. Some of these features include: (1) helium coolant, which is single phase and inert with no reactivity effects; (2) graphite core, which provides high heat capacity and structural stability at very high temperatures; (3) refractory coated plutonium oxide particle fuel, which allows extremely high burnup and retains fission products at temperatures much higherthan normal operation; (4) negative temperature coefficient f reactivityo , which inherently shuts dow e corth ne above normal operating temperatures annularn a } powe(5 w lo d , an ;r density uninsulaten cora n ei d steel reactor vessel surrounded by a reactor cavity cooling system, which enables passive heat transfer while maintaining fuel temperatures below damage limits PC-MHe Th . R ca deployee nb neae th r n di ter m becaus bases i t ei technolog n do y demonstratey db more than 50 gas-cooled reactors built and operated worldwide since 1956.

. 1 INTRODUCTION

A resulsa f receno t t arms reduction agreements betwee Unitee nth d Stated an s Russia, large quantitie f weaponso s grade plutoniu uraniud man m wil declaree b l d excess by the governments of the two countries. While the excess highly enriched uranium can be relatively easily dispositioned by diluting (or denaturing) it with natural uranium and using it as fuel in commercial nuclear reactor systems, disposition of excess weapons grade plutonium present mucsa h more difficult challenge numbeA . r of options for disposition of excess weapons grade plutonium are being evaluated in Unitee th d StateRussian i d san . Although discussions ofte f ligho ne tfocuus n so water reactors (LWRs) fueled with mixed oxide (MOX) fuel assemblies containing weapons grade plutonium or on use of liquid metal reactors fueled with large quantitie f plutoniumso , other innovative option available sar consuminr efo g excess weapons grade plutonium inventorie producind san dischargega d product thas i t unattractive for use in weapons applications and, as a result, presents no incentive or nee r reprocessindfo recycled gan .

69 Amon optione gth s being evaluate Unitee th n di d State r dispositiosfo f excesno s weapons grade plutoniu modulae th ms i r helium reactor (MHR) graphita , e moderated, helium cooled reactor system Plutoniue Th . m Consumptio (PC-MHRR nMH ) offers high levels of plutonium destruction (more than 90% of the initially charged plutonium- 239 and more than 65% of the initial total plutonium) in a single pass through the reactor - high plutonium destruction without recycle. This high level of plutonium destruction is achieved because the PC-MHR uses no fertile fuel (no new plutonium is created becausd )an refractors eit y coated particle fue capabls i l f reachineo g burnup levels in excess of 750,000 MWd/t heavy metal while fully retaining its structural integrity.

The MHR offers a unique combination of advantages for the disposition of weapons grade plutonium. In addition to its ability to achieve high levels of plutonium destruction without recycle, the PC-MHR fuel cycle has unusually high resistance to diversio proliferationd nan plane Th .t uniqufeaturee th f o e l safetsal y characteristics uranium-fuelee oth f d version safetMHRe e th Th f .s o y designR objectiveMH e th f so are met through a combination of inherent safety features and design selections that take maximum advantag f theseo e features plane Th .t relie passivn so e featureo st control heat generatio removad nan l under accident conditions without dependence on operator intervention or active, A-C powered safety systems.

. 2 PC-MHR PLANT DESIGN

The MHR has been under development by the U.S. Department of Energy (DOE) since 198s par a f 3DOE'o t s advanced reactor program desige Th . n draws upoe nth considerable gas-cooled reactor technology base. More than 50 gas-cooled reactors have been built and operated worldwide since 1956. As a result of this experience, PC-MHe deployee th b relativele n Rth ca n di y near term without nee r extensivdfo e fundamental researc developmentd han . Primary engineering development needs relate to fabrication processes for plutonium fuel in a large scale application and to integratio f poweno r conversion system component reference th r sfo e design recently adopte commerciae th r dfo programR MH l . havR versioneo MH beee Tw th n f sundeo r developmen commerciae U.Sth e n th i . n i t l MHR program reference Th . e plant desig o 199t p 3nu consistet MW f fou0 do 45 r reactor modules, eac f whicho couplehs i steaa o dt m generator that produces high temperatur pressurd ean e stea conventionaa r mfo l Rankine cycle steam-electric power generating system. A four module facility operates at a net plant efficiency of 38%, resulting in net electrical generating capacity of 692 MWe.

In 1993 extensive studies were conducte e comerciath n prograR o d MH l mo t determin suitabilite eth f replacinyo steae gth m cycle power conversion system with a closed Brayton cycle gas turbine power conversion system. Use of the Brayton cycle is desirable because it increases the plant efficiency from 38% to about 48%, This dramatic improvemen therman i t l efficiency result substantian si l improvemenn i t plant life cycle costs, less thermal discharge to the environment, less radioactive waste produce unir f electricadpe to l energy generated maximud ,an m utilizatioe th f no energy conten fuele additionn th I .f o t evaluatione th , s examine feasibilite dth f yo increasing the thermal power rating of the reactor to the maximum that can be accommodated withi same nth e reactor vessel size.

70 Based upon these studie determines wa t reactot si MW dO r thacor60 a te coupleo dt a heliu turbins mga e direct cycle power conversion system coul adoptee db e th s da •commercial MHR reference design. A four module reference MHR plant of this design can produce 1144 MWe from 2400 MWt of reactor power.

As part of the DOE sponsored Plutonium Disposition Study, both the 450 MWt steam turbins havR ga t eeMH beeMW n0 assessed60 e th plutoniue d Th an . cyclR meMH consumption, diversion/proliferation resistance, and safety characteristics of the two plants are essentially the same. The plants differ primarily in the number of reactor modules that need to be deployed to disposition a given inventory of weapons grade plutonium within a specified time period (as a result of their different thermal power ratings) plane th n ti , life cycle costresula s f theis(a o t r different plant efficiencies, their different power ratings, and their different capital costs), and in the amount and complexit powee th f yo r conversion equipment.

DOE has not, as of the time of preparation of this paper, yet published the result of plutoniur evaluations turbinit s fo R ga e eMH m th f dispositionso . Hence PC-MHe ,th R design described in this paper is the steam cycle version of the design that uses a 450 t reactoMW r module.

The steam cycle PC-MHR reference plant consists of four 450 MWt reactor modules capabl f providineo t electricane ga l generating capacit MWe2 reacto69 e f Th yo . r core is contained in an uninsulated steel reactor pressure vessel that is connected by a cross vessel to a steam generator vessel. All three vessels are designed to ASME Section I Boile II Pressur d an r e Vessel Code requirements. The locatee yar d below grade in a high pressure, low leakage containment with characteristics typical of containments of U.S. commercial light water reactors. The reactor cavity within the containmen cooles i tpassive a y db , natural convection, water cooled reactor cavity cooling system. This system operates at all times. Under accident conditions it provide r passivsfo e remova f reactoo l r decay hea conductioa vi t radiatiod nan n from uninsulatee th d reactor vessecooline th o t l g system panels. Hea transportes i t y db reactoe th r cavity cooling syste naturaa mvi l convectio heaa o nt t exchanger, where rejectes i atmospheret e i th o dt , whicultimate th hs i e heat sinkreactore Th . , steam generator, cross vessel primard an , y heat transport system arrangemen showe ar t n in Figure 1.

1 2. PC-MHR FueCord an le Design

Standard refractory coated particle fuel, shown in Figure 2, is used in the PC-MHR. Coated particle fuebees ha ln use higdn i h temperature helium cooled reactor morr sfo e than 25 years. The fuel is in the form of tiny plutonium oxide fuel kernels coated first with a porous graphite buffer layer followed by layers of silicon carbide and pyrolytic carbon. This syste f mparticlo e coating TRISreferrea s si s a O o dt coating e Th . diameter of the coated fuel particle is approximately 615 /vm. The particles are mixed with graphitic material and formed into cylindrical fuel rod compacts approximately 12.45 mm in diameter and 49.3 mm in length. The fuel rod compacts are inserted into hexagonal prismatic graphite fuel element blocks, as shown in Figure 2. The completed fuel elemen approximatels i t y 0.7 heighn i 90.3d m widtan 5 m e acrose sth flats, and it weighs about 115 kg. A standard fuel element contains 202 fuel holes, coolan8 10 t channels, about 3000 fuecompactsd ro l aboud ,millioan 2 2 t n coated fuel particles. Extensive operatio tesd ntan dat refractorn ao y coated fuel particles confirm their performanc beyond an maximue o t d th p eu m temperatures tha expectee ar t o dt occu design i r n basis accidents.

71 CONTRO DRIVED LRO / REFUELING PENETRATIONS

STEEL REACTOR VESSEL

MAIN CIRCULATOR ANNULAR MOTOR/MACHINE ASSY REACTOR CORE

MAIN LOOP SHUTOFF VALVE

STEAM OUTLET SHUTDOWN COOLING HEAT EXCHANGER

STEAM GENERATOR VESSEL SHUTDOWN LOOP STEAM SHUTOFF VALVE GENERATOR SHUTDOWN CIRC MOTOR/MACHINE ASSY

FEEDWATER INLET

FIGURE 1 : PC-MHR HEAT TRANSPORT SYSTEM

72 Plutonium utilization is not a new application for high temperature helium -cooled reactors. Six irradiation tests have been conducted using high-enriched (up to 88% plutonium-239) plutonium coated particle fuel. Five of these tests were performed as Dragoe parth f o t n Reactor Project {United Kingdom late th e n sixti 1960s)e th hd an , was performed by General Atomics in the Peach Bottom gas cooled reactor in the early 1970s eacr Fo f thesh.o e tests fuee th , l consiste f TRISdo O coated plutonium fuel kernels. The Dragon tests consisted of loose particles in graphite holders. The Dragoe th fue r fo ln test fabricates swa Belgonucleairy db EuropePeacn e i th r hFo . Bottom test, standard fuecompactd ro l s were fabricate k RidgOa et a dNationa l Laboratory. Test conditions included irradiation at temperature up to 1450°C, peak fuel burnu 747,00o t p pu 0 MWd/t fasd an t, neutro 10x n22 5 fluenc2. n/m o E t ( 2p eu > 0.18 Mev). These temperature and burnup conditions are expected to envelope those experience coatee th y db d fuel particle PC-MHRf thesa o l n si al e r testsFo . e ,th plutonium fuel performed well, establishin feasibilite gth f higyo h burnup, TRISO- coated, plutonium fuel for the MHR. reactoe Th r core consist 0 hexagona84 f so l prismatic fuel elements stacken a n di annular array of 84 columns, 10 elements high, as shown in Figures 3 and 4. One half of the fuel elements is replaced annually. Because the PC-MHR uses no fertile fuel material, excess reactivity control and negative temperature feedback is provided through use of other materials. For the PC-MHR these functions are provided by use of erbium oxide poison rods locate n selectei d d fue l e graphitholeth n i s e fuel elements. Use of graphite as moderator and structural material in the PC-MHR results ithermana l neutron spectrum tha relativels i t y hard compare typicao dt l light water reactors, i.e. pea e averagd ,th kan e neutron fluxet highea e sar r energy levelse Th . harder spectru plutonium-23V msuce s i 3 h 0. tha e tth 9 thermal absorption resonance lies within the thermal flux spectrum, which enhances the plutonium-239 consumption rate relativ otheo et r nuclides. Furthermore e corth ef i , heat abovp u s e normal operating temperatures, which would occur during certain design basis eventse th , thermal flux would shift into the strong erbium absorption resonance and then into the large, 1.0 eV plutonium-240 capture resonance, providing negative reactivity feedback to terminate the transient.

2 2. PC-MHR Safety Design

The PC-MHR features the same unique safety characteristics of previous uranium- fueled, steam designscyclR safete MH Th . y desigPC-MHe th f no s baseRi n do retaining virtuall fissiol yal n products withi fuee nth l particle coatings unde accidenl ral t conditions integrite e particlTh th . f o y e coating barriea s sa o radionuclide t r s i s maintaine y controllindb g core heat generation, assuring adequate mean f heaso t removal, and limiting air/water chemical reactions with the core graphite or the fuel particles. The safety design objectives of the MHR are met through a combination of inherent safety feature desigd san n selections that take maximum advantag f theseo e features. Som f theseo e features include heliu) (1 : m coolant, whic singlhs i e phase and inert with no reactivity effects; (2) graphite core, which provides high heat capacity, slow thermal response, and structural stability at very high temperatures; (3) refractory coated plutonium oxide particle fuel, which allows extremely high burnup and retains fission products at temperatures much higher than normal operation; (4) negative temperature coefficien f reactivityo t , which inherently shuts dow core nth e above normal operating temperatures annularn a ) powe(5 w d lo , an ;r density corn ei an uninsulated steel reactor vessel surrounde reactoa y db r cavity cooling system, which enable passive heat transfe ultimate th o t r e heat sink while maintaining fuel temperatures below damage limits.

73 FUEL PARTICLE FUEL COMPACT PRISMATIC FUEL ELEMENT

OUTER ISOTROPIC PYROLYTIC CARBON 1

SILICON CARBIDE BARRIER COATING TRISO COATING INNER ISOTROPIC PYROLYTIC CARBON 1.9. 4IN

POROUS CARBON BUFFER

Pu KERNEL 0.49 IN.

TRISO COATED PARTICLE

FIGUR : E2 PC-MH R FUEL ELEMENT COMPONENTS CENTRAL PENETRATION

IN-VESSEL FLUX MONITORING UNIT (IFMU)

OUTER REFLECTOR IN-VESSEL FLUX MONITORING UNIT (IFMU) )(6

WEAPON'S GRADE NEUTRON PLUTONIUM FUEL CONTROL & ELEMENT REFUELING COLUMNS4 (8 ) PENETRATION (18)

• 36 CONTROL RODS (IN CORE & OUTER REFLECTOR) RESERV2 1 O E SHUTDOWN CHANNELS (IN CORE)

FIGURE 3: PC-MHR REACTOR CORE ARRANGEMENT INNER REPLACEABLE REFLECTOR

FUEL ELEMENTS

(ACTIVE CORE BOUNDARY)

OUTER REPLACEABLE REFLECTORS

LOWER REPLACEABLE REFLECTOR

FIGUR : E4 PC-MH R REACTOR CORE ELEVATION

76 These safety design features resul reactoa n i t r thawithstann ca t d los f coolanso t circulation or even loss of coolant inventory and maintain fuel temperatures below damage limits crediblo N . e severe accident scenario involvin glarga e releasf eo radionuclides fro core beems th eha n identifie PC-MHRe th r dfo . Extensive analyses of a spectrum of design basis events and beyond design basis events confirm that the safety characteristics of the PC-MHR are the same as those of earlier uranium-fueled MHR designs. Accordingly, it is expected that regulatory requirements will be met with margin and that offsite doses resulting from design basis accidents will be less tha threshole nth t whicda h protective action requiree offsite sar th r edfo population.

3. PLUTONIUM DESTRUCTION CAPABILITY

Figure 5 presents a comparison of the plutonium destruction capability of various plutonium disposition options recyclo N . f speneo t fue assumes i l e reactoth r dfo r options. It is clear from this figure that the vitrification option, in addition to recovering energ nonthe eof y conten weaponthe of t s grade plutonium, does nothing to diminish the usefulness of the material in weapons applications. Furthermore, options that reduce the weapons grade material to a discharge equivalent to that containe commerciadn i i reactor spent fuel (the "spent fuel standard" referree th n i o dt U.S. National Academy of Sciences report on disposition of excess weapons grade plutonium) recove energe rth y content from only 25-30 plutoniue th %f o leavd man e a discharged isotopic mixtur (ane ebeenb s thadn ha ca )t use nucleaa n di r weapons application.

Figure 5 indicates that the PC-MHR is the most effective option for destroying weapons grade plutonium without recycl maximizind ean energe gth y content recovery from the material. The PC-MHR is capable of consuming more than 90% of the

NET CONSUMPTION % 0 9 PU23 TOTAL Pu 0%

WEAPONS GRADE VITRIFICATION COMMERCIAL PC-MHR PLUTONIUM SPENT FUEL SPENT FUEL STANDARD

FIGUR : E5 PLUTONIU M DESTRUCTION CAPABILIT VARIOUF YO S DISPOSITION OPTIONS

77 initially charged plutonium-23 initialle mord th f 9an o e y thacharge% n65 d total plutonium in a single pass through the reactor. The discharged fuel is less than 30% plutonium-23 containd an 9 s large quantitie f plutonium-24o s -242d 0an , which significantly reduce usefulness sit weaponn si s applications. Buildu f actinidepo s other than the higher isotopes of plutonium in the spent fuel is relatively small - equivalent to about 2% of the total actinide discharge, predominately americium-241 and -243.

Dependin specifice th fuee n gth o lf cyclso e chosen fou,a r module PC-MHR plann tca process 0.9 to 1.1 metric tonnes of weapons grade plutonium annually to the level of destruction shown in Figure 5. Over a nominal 40 year design lifetime, a four module plant would process approximately 36 to 44 metric tonnes of material. The plutonium processing rate is directly proportional to the number of reactor modules deployed and thermae th o t l power ratin f eacgo h module.

The plutonium destruction capability of the PC-MHR is not limited to the consumption of weapons grade plutonium. Evaluations by General Atomics show that the high level weaponsof s grade plutonium destruction made possibl higthe h eburnuby p capabilit coatee th f yo d particle alsfue n achievee oca b l usiny db g commercial reactor grade plutonium as the feed material. Therefore, the PC-MHR offers a relatively near term option for solving a problem that has been recently characterized by the U.S. National Academ f Scienceyo requirins sa g long term step resolvo st e- reducin e gth proliferation risks pose t onl excesdy no y b s weapons grade materiale t th als y sbu ob large and growing inventory of plutonium in commercial reactor spent fuel.

4. RESISTANCE TO DIVERSION AND PROLIFERATION

The PC-MHR fuel cycle offers a very high level of resistance to diversion and proliferation. The plutonium content of each fresh fuel element is very low (less than one kilogram), and the plutonium is much more difficult to recover from fresh PC-MHR fuel than from fresh MOX fuel assemblies. It is very difficult to divert a sufficient quantity of PC-MHR spent fuel to obtain a weapons-useful quantity of plutonium; the technolog separatinr yfo g plutonium spen froR t beemMH t no fuens developedha l d ;an the isotopic discharge of the spent fuel makes it unattractive for weapons applications.

Diversion of PC-MHR spent fuel is difficult for all the reasons that apply to any reactor spent fuel (high radioactivity, large mass, high internal heat generation rate, extensive safeguard securityd san ,reasoe etc.) on t als r nbu ,ofo tha uniqu s i tPC-MHRe th o et : plutoniuw lo m conten fuer pe tl element. Whil plutoniue eth m conten singla R n i t eLW MOX spent fuel assembly is of sufficient quantity to use in a weapons application, one would have to divert up to seventeen truckloads of PC-MHR spent fuel elements (each element weighing abou kilograms5 11 t obtaio t ) similana r quantit f dischargeyo d plutonium. The diversion of up to seventeen truckloads of spent fuel shipping casks is a considerably more daunting and readily detectable task than diversion of one truckload of LWR spent fuel.

Even if such a large quantity (both mass and volume) were diverted, the technology for separating plutonium from MHR spent fuel has not been developed, in contrast to LWRe th r whicfo , h plutonium separation technolog wels yi l establishe wideld dan y known procesa f I .r separatiofo s f plutoniuno m from PC-MHR spent fuel were developed amoune ,th f equipmento t required woul muce db h more extensive thar nfo other fuels, and the number and complexity of basic processing steps would be much higher.

78 Finally, althoug mixtury han f plutoniueo m isotopes can principlen ,i usee b , mako dt e a nuclear device, the plutonium discharged from the PC-MHR, being particularly high n plutonium-24i -24d 0an 2 content s muci , h mor en weaponi difficule us o st t applications than the discharge from reactors that do not achieve as high a burnup level. The increased difficulty in weapons design, construction, and delivery results from the higher critical mass required, the higher spontaneous neutron emission rate (which increases yield uncertainty), the higher radiation hazard (which makes it more difficul proteco t t t assembly personnel from excessive radiation exposure)e th d an , higher decay heat generation rate fro dischargee mth d plutonium (which make st mori e difficult to design and assemble a device and to maintain the tight mechanical tolerances required) n additionI . e relativelresula th s f a o , t y short half liff o e plutonium-241 (14.4 years), which constitute- PC e s th approximatel f o % 38 o t 3 y3 MHR plutonium discharge requiree th , d critical mass will increase with time after separation, further complicatin nucleae gth r desig devicea f no .

The diversio proliferatiod nan n resistanc e PC-MHth f eo R make t i swel l suiter dfo international deployment for plutonium disposition. The processes involved in fabricatio f PC-MHo n R fuel, reactor operation d spenan , t fuel disposition lend themselves to full implementation of IAEA safeguards controls. Discussions were begun in June between General Atomics and the Russian Ministry for Atomic Energy (MIIMATOM) regarding replacement of the plutonium production reactors at Tomsk with PC-MHRs.

5. SPENT FUEL DISPOSITION

The low plutonium content per spent fuel element and the undesirable mixture of plutonium isotope spene th n si t fuel leave little nee r incentivdo reprocessinr efo r go recycle of the PC-MHR spent fuel. Evaluations by both General Atomics and Oak Ridge National Laborator U.Se th .n y i indicat e tha refractore tth y spent fuel elements are an excellent waste package that is suitable for direct disposal, in whole element form inserted int omultipurposa e canister geologia ,in c repository TRISThe . O coated particl benefie eth fues f lono tha l g term containmen f radionuclideo t s without relying on the performance of the waste package. Quantitative assessments show that the TRISO coating should maintai integrits nit millioa r yfo n year morr srepositoro a n ei y environment. In contrast, the expected lifetime of zircalloy cladding in a repository, even under dry conditions, is less than 1000 years.

6. CONCLUSIONS

The PC-MH effectivn a s Ri e syste mmaximizo t levee eth f destructioo l f excesno s weapons grade plutonium without recycle. The plant produces nuclear electricity at high level f efficiencso y while providin uniquga e leve f nuclealo r safety PC-MHe Th . R fuel cycle has high resistance to diversion and proliferation, making the system particularly well suite r internationadfo l deploymen r plutoniufo t m disposition.

Next page(s) Left blank 79 ADVANTAGES AND LIMITATIONS OF THORIUM FUELLED ENERGY AMPLIFIERS

. MAGILLJ . O'CARROLLC , . GERONTOPOULOSP , , K. RICHTER, J. VAN GEEL European Commission, Joint Research Centre, Institut Transuraniur efo m Elements, Karlsruhe, Germany

Abstract

The recently proposed energy amplifier reactor concept, based on the use of an accelerator driven sub-critical assembly with thoriu e breedinth s ma g fuel, offers potentially significant advantages over conventional reactors in the nuclear fuel cycle. The main advantages stated are addition ,i providino nt reactoga r whic essentialls hi y sub-critical, that much less transuranic actinide wast s generatei e e ris f nucleath o k d an dr proliferatios i n substantially reduced. In this paper, we analyse the proliferation and radiotoxicity problems associated wit varioue hth s energy amplifiers using 232 238d T hbreedinUs an a g fuelss i t I . shown that there are major advantages to be obtained in radiotoxicity using 232Th especially if the bred 233U and additional uranium isotopes are recycled. In addition, the fuel can be rendered proliferation resistant through mixing with the isotopic dénaturant 238U. This is in contrast to uranium fuelled reactors where there is no natural dénaturant for plutonium isotopes. It is also shown, however, that through the use of this isotopic dénaturant for proliferation resistance of thorium fuel, a new source of radiotoxicity arises. The proliferation and radiotoxicity problems are now coupled together and this leads to limitation in the overall advantages to be gained. This is the proliferation / toxicity dilemma of a thorium cycle. Nevertheless, dependin n whico g h aspects have priority i.e. proliferation resistance, radiotoxicity of the spent fuel, or combined proliferation - radiotoxicity, a thorium fuelled energy amplifier can have clear advantages over conventional reactors in the nuclear fuel cycle.

1. INTRODUCTION

Energye th n I Amplifier (EA) concept [1,2 inpue ]th t energ amplifiee th o yt provides i r d by high energy charged particles produce a particl y b d e beam accelerator e chargeTh . d particles, typically protons of energy IGeV, collide with a heavy target (Pb, U, or Th) and each proton produces approximately 50 spallation neutrons of average energy 20MeV. The amplifier consist f sub-criticaso l assembl 232f yo Th/233U with exampler fo , , kgg 0.9= - 5 such that a further neutron multiplication of l/(l-keg) = 20 can be expected. Depending on the moderator/coolant used, these neutron completele b n sca y thermalised (graphite moderator), partially thermalised (pressurised water moderator) or remain fast (lead coolant) and result in thermal, pressurised water or fast energy amplifiers denoted by T-EA, PW-EA, and F-EA respectively. In contrast to conventional critical reactors (where for thermal reactors typically 3% fissile material is required in the fresh fuel to give kgff = 1), steady state operation of the T-EA and PW-EA results in breeding equilibrium (with typically 1% fissile material) with each fissioned nucleu fuea y ls b nucleus bred fro fertile mth e Th. Refuellin requires gi d onlo yt 232 overcome fission product avoio poisonint r o d / materiad gan l damage. Thi interestins si g with respect to the F-EA where, because of the much smaller fission product poisoning, very high

81 burnups can, in principle, be achieved. The fuel fabrication step may also be avoided by placing the natural 232 blankea T n hi t aroun core dth e unti sufficiena l t concentratio f fissilno e 233U has been bred. Another interesting aspect of the EA is that the driving neutrons are independent of the sub-critical assembly. Fro e neutromth n kinetics poin f viewo t , these neutrone b n ca s considered as delayed in that they do not appear immediately from the fission process (hence t appea exponentiae no the th o yn d ri l ter neutror mfo n multiplication [3]) criticaa n .I l reactor 0.7e th %s i t (for 235U) delayed neutrons which make controe reactoth f o lr possibly eb reducing k^ for prompt neutron multiplication to 0.993. Obviously reactivity insertions which give a okgff > 0.007 will lead to rapid neutron multiplication. In an EA with ^3-= 0.95, a okgg- requires i 5 >0. o producdt same eth e effect. Conversel toleratn Aca E e eyth much greater reactivity excursions than critical reactors thir s.Fo reaso possibilite nth reactivita f yo y induced accident is much reduced. As a consequence, this may lead to less demanding safety requirements, fewer studies of hypothetical accident scenarios, and/or faster licencing procedures. There are, however additionao tw , l problems associated with conventional nuclear fuel cycles i.e. proliferatio wastd nan e disposal followine th n I . g section considee sw r these aspects in detail.

2. RADIOTOXIC1TY OF THE SPENT FUEL

radiotoxicite Th nuclida f yo determine s ei Annuae th s it l y Limidb 1 f Intako t AL r eo value. These ALI values incorporate dose limits, radiation weighting factors, tissue weighting factors, metaboli biokinetid can c informatio I valuen AL ingestior [4]e sfo Th .inhalatio d nan n for radionuclides have been evaluated [4j. The ratiotoxicity of a nuclide is given by f^ Radiotoxicity = (0.02Sv)—— (1) ALI

where A is the activity of the nuclide expressed in Bq. The factor 0.02Sv is the yearly limit of occupational exposure with a maximum permissible exposure of 0. iSv over a five year period. In the following, results are given relative to the ALI values for ingestion (results for inhalation are greater numerically but follow the same trends). This study focuses on the potential radiotoxicity of spent fuel without considering any specific accident scenarios or treatment of the mobility and migration of nuclides in the geo- and biospheres. For example, N produces pi smaln di l amounts, however solubls i t i , waten ei r makin t morgi e mobild ean potentially more dangerous than other actinides. Not ethoriue thatth n i , m fuel s cyclei p N , produce amountn di ordee s on magnitudf ro e les suraniue thath n i m cycle. The actinide inventory in the thorium spent fuel from the T-EA, PW-EA and F-EA's have been taken frof [2Jinventore mRe Th uraniue . th r yfo m spent bees fueha ln calculated using RADONN [5]totae th fig n ., I l 1 .actinid e radiotoxicity resulting fro uraniue mth m fuen i l a T-EA with a burnup of 40GWd/t at discharge is given. The decay evaluations and radiotoxicity calculations were made using RADONN [5] developed at ITU. The fission products pose a problem in the short term as they dominate the radiotoxicity for the first tew hundred years. There are also potentially dangerous long lived fission products such as "Tc and I. There is a difference in the inventory of fission products produced by 23 3U and 235U fissio 129n due to the slight mass difference. However, the total radiotoxicities hardly differ and are not displayed.

82 In the figures 1-9, reference is made specifically to T-EA, PW-EA and F-EA reactors. The radiotoxicity results, however, also apply to conventional critical reactors with similar neutron spectra i.e. thermal (graphite moderator), partially thermalised (pressurised water), and last reactors radiotoxicitiee Th . actinidee th f theid o s an s r daughter product groupee ar s d together according to their chemical origin. Hence in fig.l the label "Pu" implies the four plutonium isotopes 239Po, 240Pu, 242Pu, 243Pu (which are present at discharge) and all their daughter products. This representatio advantage th clearle s se n ha effecte n yth e ca f thae so on t recycling particular isotopes from the waste. At a cooling time of lOOy, the "Pu" contribution to the radiotoxicity is dominated by 241Am (433y) which is the daughter of 241Pu (14.4y). Final decay products of 24lPu are 237Np (2x10^) and 229Th (7.7xl03y). At 10^ the "Pu" is dominate 240y db P u (6.6xl0 daughte3d y)an r 236U (2.3xl07 y)t 10maie A .5 yth n contribution comes from parent 239Pu (2.4x10^) with daughter 235U (7xlO*y) and 242Pu (3.7xl05y). Initially the "Cm" curve is dominated by 244Cm (18. ly) with daughters 240Pu (6.6xl03y) and 23<>U (2.3xl07y). 245Cm (8.5xl03y) decays through 24lPu (14.4y 241d A)an m (433yo )t 237Np (2.1x10^). At discharge the "Am" consistes primarily of 243Am (7.4x103y, decay products 239Np (2.4d), Pu (2.4x10^)d Uan , (7x10^) a smal d lan ) amounf o tA m (lOh, decay 239 235 244 products 244Cm (18.ly), 240Pu (6.6xl0 2363d y)U,an (2.3xl07y)). Although the "U" consists initially mainly of 238U (4.5xl09y), there is a small quantity of 239U present. This 239U decays almost immediatel 239o t y Np (2.4d thed 239o an t n) Pu (2.4x10^ gived )an radiotoxicitsa y contribution whic approximatels hi y twice tha 238f o t U. Beyond 105y the toxicity arises essentially from 210Pb a daughter of 238U. Hence the radiotoxicit naturaf yo l 238 Uapproximatels i valu e thire th yf on edove " o give"U r y 10*ynb . In fig. 2, 3, and 4, the ingestion radiotoxicities of spent thorium fuel with a burnup of 40GWd/t from T-EA, PW-EA, and F-EA's are shown. In the PW-EA the light water moderator results in a thermal neutron spectrum with considerable epithermal and fast neutron components. In the F-EA, use of a lead moderator/coolant results in a primarily fast spectrum. Hence in the PW-EA and F-EA's the higher energy neutron (n,2n) reactions with the 232Th are possible and give rise to 232U and 23 4*a. These isotopes play a key role in a thorium fuel cycle. In these figures., the label "U" represents the five uranium isotopes 232U, 233U, 234U, 235 236d UU/?/ttan , theil sal r daughter products fig.2n I .maie th , n contribution comes initially from 233U (1.6xl05y, with daughter 229Th (7.3xl03y)), 238Pu (88y, daughter 234U (2.5xl05y)), and 232U (70y, daughters 228Th (19. ly), through to 208Pb). In figs. 3 and 4, both 232 233d UUan dominate t 10 toxicite A .5 yth dominates yi 229y db T hdaughtea 233f ro U. botn I T-E e PW-EA'd hth Aan maie sth n contributio "Puo nt fros i " m latet 238a d Pr uan times 239Pu. In the F-EA there is no plutonium contribution. In the T-EA (fig.2) the "Pa" radiotoxicity is determined by 233Pa (27d, daughter 233U (1.6xl05y)) and the "Pa" and "U" toxicities show similar behaviour bot n PW-EI .e hth F-EAd isotopAe an th , e 231Pa (3.3x10^), produce highey db r energy neutron (n,2nn si ) reactions with 232Th, contribute "Pa"o st . The "Th" radiotoxicit s determineyi 232y db T h (7.3xl0 daughted 3yan ) r 228l al R n ai cases. The "Th" curve may be considered a reference level for one tonne of natural thorium. In both the T-EA and F-EA's, 237Np is produced in small quantities. In a comparisofig 5 . e ingestios madni th f eo n radiotoxicitie f spenso t uraniud man thorium fuels fro a mT-EA-(burnu p 40GWd/t). Clearlye firsth t n i ,30,00 0 yeare th s radiotoxicit thoriue th f yo m wast approximatels ei ylesfacto0 a 3 uraniue sf th o thar r nfo m waste. From 30,000 to 106 years the radiotoxicity of the thorium waste is higher than that of uraniu approximately mb yfactoa fivef toxicite o r spene Th . th f yto thorium PW-Ea fue n i l As i higher than fro ma T-E e higheAth initiallo t r e concentratioydu f 232o n ) U 2n fro , m(n reactions radiotoxicite Th . thoriuf o y m fro mF-Ea factoa s Ahighei n te r r than that fro- T ma EA due the fact that in breeding equilibrium the concentration of 233U is a factor ten higher.

83 The radiotoxicity of spent thorium fuel from an EA can be reduced further by recycling the uranium isotopes making use of the fact that the 233U can be used to seed thorium for fresh fuel. In fig. 6, the effect of recycling 99% of the uranium is shown. It has also been assumed

that after discharge enough time (some months elapses )ha d before recyclin alloo gt w mosf o t

233 P thUdecao at e. 233 see e Frob o y t n figure n mthath ca t ei t ove firse rth t 30,000 yeare sth radiotoxicity of the thorium waste is now a factor of 200-700 lower than that of the spent uranium fuel. In figcomparisoa 7 . e radiotoxicities madni th f eo f speno s t thorium wastd fuean l e resulting from recycling. If only the uranium is recycled (99% including the Pa which has 233 decayed) radiotoxicite th , reduces yi d factoonla y yb r 2-10 ove firse th r t 30,000 yearse Th . remaining radiotoxicity is primarily due to the isotope Pa which is not separated. Only when 231 this 231Pa is recycled in addition can major reductions in radiotoxicity (100 - 1000) result. I neffece fig th recyclinf , o t.8 g spent thorium fuel fro mF-Ea shownAs i . Recyclin% g99 uraniue ofth m lead reductioa o st radiotoxicitn ni f 5-50yo 0 ove firse rth t 30,000 yearsn i f I . addition the 23 lPa is recycled the reduction can be increased to 60-1000. In fig. 9 the effect of burnup on the radiotoxicity is given for spent thorium fuel in a F- EA. It can be seen that after 103years the radiotoxicity (of actinides and daughters) per unit mas independents si burnupf o differencee Th . year2 accounte10 e isotope st ar th a s y b e r dfo concentratioUe th , f whicno h increases with burnup radiotoxicite Th . unir ype t masn sca 232 also be divided by the burnup to give the radiotoxicity per unit of thermal energy obtained by fission and this is also shown in fig. 9. Hence, taking the fuel to extended burnup does not lead to a dramatic decrease in the radiotoxicity per unit energy produced.

3. PROLIFERATION ASPECTS OF A THORIUM FUEL CYCLE

From the previous section it has been shown that considerable advantages regarding radiotoxicit spene th obtaine e f b tyo fuen ca lthoriuf d o througe us m e (hth Th) rather than 232 natural uraniu breedea s ma energe r th fue n i yl amplifier. Even without recycling waste th , e from the thorium fuel in a T-EA is approximately a factor 30 less toxic than the waste from uranium ove firse rth t 30,000 years cooling time recycliny B . uraniue gth m - isotopeT e th n si EA, and in addition the Pa in the PW-EA and F-EA, this factor can be increased by another orde f magnitudero . Ther , howevereis 231 problea , m here sinc e recycleeth d uranium froe mth

spent thorium fuel coul divertee db used dan construco dt nucleaa t r explosive highle Th . y

232 fissilU eproduce233 d by irradiationT ohf is present in the fuel together with other uranium isotopes. The reactor grade uranium does not have the neutron reproduction characteristics of pure 233U, nevertheless, criticality can be achieved. In contrast to reactor grade plutonium associated with the uranium cycle, where sophisticated engineering is required for the implosive assembly [6,7], the reactor grade uranium fro thoriue mth m cycle coul assemblee db relativela y db y simpl typn egu e device. Because of the high neutron background from the plutonium isotopes, a gun type assembly of such material will result in predetonation. An estimate of the neutron background threshold beyond which an implosive assembly is required is 10 neutrons/kg/s as shown in fig. 10 (in the 4 Trinity device [8] 240e ,th P u conten smalles wa t percenx r si tha e nth t value which currently defines weapons grade Pu. Here we take a reference value of one percent to fix a value beyond which an implosive assembly is required - this results in 1.2X104 neutrons/kg/s). Also shown in fig. 10 are the neutron background rates tor the recycled uranium isotopes in spent thorium fuel fro T-EAe mth , PW-EA F-EAd an , t differensa t burnups. Clearly reactoe th , r grade uranium has a low neutron background allowing critical assembly with a gun type device. (An alternative proliferation pathway [9] is to chemically separate the Pa from the waste. Since 233

84 this decays to 233U with a half life of 27 days potentially pure, highly fissile, material can be obtained. It would, however, require an entire charge of a T-EA i.e. 5 tonnes to obtain enough criticae 233on r Ulfo mass (approximatel [10]))g k y5 . A potential deterrent to diversion is, however, the gamma activity of the material. The major source of this gamma activity is 2U8T1, a daughter product of 232U, which emits highly penetrating 2.6MeV gamma photons. The quantity of U in the uranium recycled from the spent thorium fuel fro mT-EAa , PW-EA F-Ed functioa an , As a burnuf no shows pi tabln i e232 1. Eve t burnupna 40GWd/tf so PW-Ee th ,hig a s hA ha concentratio 232f no U. Wit F-Eha A reaco t s hha burnup e on excesn si f 160GWd/so attaio t t n simila expectee rb levelso t s dA . fro T-EAe mth concentratione th , 232f s o muc e Uar h les PW-EAe s thath r nfo .

Tabl Fractiona. e1 l content (ppm 232f )o Urecyclen i d uranium from spent thorium fuel from different Energy Amplifiers as a function of burnup.

Bumup T-EA PW-EA F-EA 40GWd/t 200ppm 3100ppm 500ppm 80GWD/t 200ppm SOOOppm 900ppm 160GWd/t - - 2200ppm

The dose rate from 5kg (approximate critical mass) 233U containing 100, 1000, and 10,000ppm 232 Ut varioua s times after chemical separatio shows ni fig.n i . Als1 l o shows ni e LD-5th 0 whic hchanc % give50 f deata seo h followin e houon g r exposure. Clearle yth material is strongly gamma active, but only in the case of high concentrations of 232U of 10,000ppm doe activite sth y lea significano dt t barrie diversiono rt t shoulI . notee db d thae th t concentration of 232U can be increased in the PW-EA and the F-EA by recycling the isotope 231Pa [11]. The presence of 232U in the material will, of course, complicate the uranium recycling procedure, in table 2 we give the lead thicknesses necessary to reduce the gamma dos 2.5uSv/ho t em rat1 t ea r fro233g mUvariout 1k a s times after chemical purification.

Table 2. Lead shield thickness (cm) necessary to reduce the gamma dose rate at 1m distanc 2.5uSv/ho et r 233frog Um1k with various concentration 232f so U following chemical purification. Ageing Time(days) 232TJ/233TJ 10 30 100 400 L 4000 lOOppm 6.75 10.4 13.3 16.4 18.4 lOOOppm 12.1 15.5 18.4 21.5 23.5 lOOOOppm 17.2 20.8 23.5 26.0 28.1

It should be clear that if recycling is to be part of the thorium fuel cycle then there is a potential proliferation problem associated wit recyclee hth d reactor grade uranium contrasn I . t to reactor grade plutonium ifrom the uranium cycle, the reactor grade uranium can be isotopically denatured wit isotope hth e criticae Uth tabln I . , l3 e masse f mixtureso f o s 233 238d givenUe Uan valuee ar Th . s her relevane F-Eee ar th Ao t t sinc reactoe eth r uraniu 238 m consists almost entirel calculationse f yo th Un I . , nuclear datr afo P u have been useo dt simulate 233e estimatth Uo T . criticae eth l mixtur groupe mase 233th on f se o eth relation s given i 239 refs. [6,7,10] have been used. Clearl ymixina g rati approximatelf oo (party9 s 238 (par1 U o )t 233U) will require a total mass of the explosive device of approximately one tonne (between 940kg for the bare sphere and 1200kg for the core plus reflector) - a value which is completely impractical. On the other hand, from table 3, a 1:1 mixing ratio (which implies 10% 238U in the

85 fuel) has a critical mass of only 14kg with a reflector. A fifteen tonne F-EA fuel inventory [2], which contains 1500kg reactor grade uranium (total powe 2GW« r ths sufficien)i , 3 8 r fo t (without reflector) to 300 (with reflector) critical masses. Clearly, to safeguard this uranium, a mixing requiredratis i 1 o 9: betwee d . an Wit 1 n 4: hlowe a 238e % rther1 th Ulimin 40 4: i s f ei t o fuel and this will be the dominant factor in the radiotoxicity.

Table 3. Approximate critical masses of isotopically denatured 233U for various mixing ratios. Data for 239Pu have been used to simulate 233U.

without reflector* with reflector* with reflector o/0233rj(238U:233rj) 233U+238rj 233ij-|-238ij Total mass (kg) (kg) (kg) 10%(9:1) 940 230 1200 20% (4:1) 240 61 560 25% (3:1) 180 46 480 33% (2:1) 110 28 390 50% (1:1) 54 14 300 100% 18 5 210 * critical mass with/without a 10cm uranium reflector

contrasn I F-EAe th , o t t reactor grade uranium froPW-Ea mr T-Ea o A A contains

significant quantities of the isotope 234U. Since the fission properties of 234U are similiar to U, 238 this acts also as an isotopic dénaturant. From table 3 it can be seen that a 1:1 mixture of 233 critica238a d s UUan ha l mass simila thao r235t f o t U. Henc nucleae eth r dat 235r n aUfo ca be used to simulate the reactor grade uranium from a T-EA or a PW-EA. tablcritican e I th e4 l masse mixturef so reactof so r grade uraniu 238d givenme Uan ar n I . these calculations nuclear dat 235r afo U have been use simulateo dt reactoe th r grade uranium. Her emixina g rati approximatelf oo (party4 s 238 (par1 U o )t reactor uranium) will resula n i t total mass of the explosive device of approximately one tonne (990kg without reflector, 1200kg with reflector), again a value which is impractical. With a mixing ratio of 1:1, which implies approximatel 238 amoun% n yU2 (a t equa thao 233t lf o t U+ 234U), 48kg reactor grade uraniu mrequires i criticaa r dfo l mass (with reflector) smalA . l five tonne T-Er PW-Eo A A fuel inventory (total powe 250MW« r th) contains approximately 1.5-2 %r 75-100ko f reactogo r uranium. Hence, the entire core is sufficient for about five critical masses and diversion of such material should thus be easily detected. This 1:1 mixing ratio, however, must be regarded as a lower limit for proliferation resistance. From the radiotoxicity, the 2% 238U used to denature reactoe th r uranium will resul radiotoxicita n i t yfactoa les0 5 rs than that resulting froe mth 238f o Ubreedes e a us r fuel t shoulI . notee db d that even after accountin differene th r gfo t power ratings of the fuels (for F-EA 160MW/1, for T-EA SOMW/t) approximately five times as many critical masses can be obtained from non-denatured F-EA fuel than from T-EA fuel/or the same power produced. In summary, the proliferation problem associated with the reactor grade uranium can be overcom denaturiny eb g with 238U. However introduciny b , dénaturane gth t 238U int fuele oth , we limit the radiotoxicity advantages associated with the recycled thorium fuel. This is (he proliferation / toxicity dilemma of the thorium based fuel cycle. Denaturing the reactor grade uranium wit equan ha l quantit 238f yo U thi( s lead 238a Uo st concentratio shoul) 2% e db « nof regarded as a lower limit for non-proliferation. The resulting radiotoxicity is a factor 50 lower than that obtained by using 238U as the breeder fuel in a T-EA or a PW-EA. (Notice here that 232 238 232 238 since Th "burns fastea t "a r rate U (ac( Th 7.4b)= , ac( U 2.7b))= burnua , f po 40GWd/ 232r fo tT h fuel correspond bumua o st f ISGWd/po 238 e toxicite Uth e r .Th th tfo f yo

86 Tabl . Approximate4 e critical masse f isotopicallso y denatured reactorr variour jfo s mixing ratios. Datr afo U have been use simulato dt e a<*orrj. 235 re without reflector* with reflector* with reflector o/0reacu(238U;reacrj) reactorrj-)-238Tj reactorrj-j-238{j Total mass (ks) (kg) (kg) 20% (4:1) 990 250 1200 25% (3:1) 650 160 930 33% (2:1) 400 100 710 50% (1:1) 190 48 500 100% 62 16 320 * critical mass with/without a 10cm uranium reflector

238U waste is dominated by 240Pu and 239Pu and the concentrations of these isotopes at 15GWd/ approximatels i t valuee th f 40GWd/t o sa % y75 t -referree Uth o dt burnup) . Wit F-Ee th hsituatioe Ath s morni e extreme 238o isotopicallT . y denatur e requireon e s between 4 and 9 times as much U to denature and this will result in a radiotoxicity 238 approximately 40-90% of that obtained by using 23*U as the breeder fuel. Clearly, isotopically denaturin thoriue gth m breede F-Ee r th fue An i l significan doey t leaan sno o d t t advantagen si radiotoxicity. In addition, isotopic denaturing leads to an increase in the absolute amount of 233U require gammhence d th dan o et a activitmateriae th f yo l require construco dt explosivn a t e seee b device nn froca t mI . tabl thate4 , wit mixinha g rati 4:1f masoe o th ,fissilf so e material (reactor uranium plus dénaturant U) for a weapon is approximately 250kg. Approximately 238 one fifth of this is U i.e. 50kg. Hence the dose rate from the U will be ten times higher than thos casPW-Ee ea th f give eo n I fig. An i .11 recycle 233 d uranium houe on , r exposur thio et s 232 material will lead to a lethal dose 100 days after chemical separation. The T-EA recycled uranium will be a factor ten less gamma active.

4. CONCLUSIONS

comparisoA fuelf thermana o n i s l energy amplifie burnua t a r f 40GWd/po t shows that spent thorium fue considerabls i l y less radiotoxic than spent uraniu mfirse th fuel t n I lOOOy. , the toxicity reductio approximatels ni frod yfactoa an m , lOMof50 r reductioe yth factoa s ni r thoriuf o e 10Us m. breeder fue botn PW-Ei le hth F-Ed Aan Aless i s favourable with regaro dt radiotoxicity. In the case of the F-EA the radiotoxicity of the spent fuel after 40GWd/t is a facto highen te r r than thae mucth t fro ho e t T-E mhigheth e Adu r level f fissilo s e 233U required partle .b Thin y sca compensate goiny db higheo gt r burnups (since fission product poisoning is less severe). However even with a burnup of loOGWd/t one can only recover a factor four in radiotoxicity (fig. 9). We conclude that the use of thorium breeder fuel in combination wit hT-Ea A will potentiallleaa o dt y significant decreas radiotoxicitn ei y even i a once through, non-recycling scenario. radiotoxicite Th e furtheb n yca r decrease recycliny db uraniue gth m isotopes. These uranium isotopes can then be use to seed thorium and produce fresh fuel. From the T-EA, the recycled wastradiotoxicita s eha t leasya factoa t les0 s10 r tha (non-recyclede nth ) uranium waste over the first Î0,000years (thereafter it is about one order of magnitude less toxic). With the PW-EA and F-EA, recycling the uranium isotopes does not lead to as significant reduction

87 radiotoxicite in th T-EA e casmuste e th th f eadditiono On .n n i i , s ya , recycl isotope eth e Pa before major reductions in radiotoxicit 23i y can be achieved. maie Th n disadvantage associated with recyclin uraniue gth m isotope thas si t this reactor grade uranium is fissile and constitutes a proliferation problem. In contrast to reactor grade plutonium associated wit uraniue hth m cycle, where sophisticated engineerin requires gi r dfo the itnplosive assembly reactoe th , r grade uranium fro thoriue mth m cycle coul assemblee db d in a relatively simple gun type device. Because of the high neutron background from the plutonium isotopes typn gu e a ,assembl sucf yo h material will resul predetonationn i t . A possible solution to safeguard the reactor grade uranium is to denature with a small amoun 238f o t UmixinA . g rati 238f oo U:reactor Uapproximatelf - o PW T-Ea d r Afo an 1 y4: EA, and 9:1 for a F-EA, would result in a total mass of the order of one tonne (fissile material plus reflector ensurd an ) impracticabilite eth f sucyo h materia r constructiofo l nucleaa f no r explosive additionn I . materiae th f i , l resulted fro mPW-EAa houe on , r exposure would result ilethana l radiation dose. Clearly ,thesf howevero e e us leve e f dénaturanth ,o l t will contribute majoa r sourc radiotoxicityf eo mixinA . thoriue gresultth % 1 Urati5 n 4: i m » f o o n s fuei l will giv radiotoxicitea y reductio 238 onlf no yfactoa ove0 r2 valure uraniu a thath f f eo to m fuen i l a T-EA. In the F-EA, there is radiotoxicity reduction is less. Hence, although the proliferation resistanc recyclee th f eo d uraniu messentialls i y guaranteed sourcw ne radiotoxicitf a e,o s yha been introduced. This we call the proliferation / toxicity dilemma of the thorium based fuel cycle. It mus emphasisede b t , however, tha contrasn i tisotope u P o t twhicr sfo h thero n s ei natural isotopic dénaturant usee , b 238 denaturo dt n Uca reactoe eth r uranium produced with thorium breeder fuel. If proliferation is the major issue, then an essentially proliferation resistan designede b n tca wastf I . e disposa maie th ns li issue , order magnitudf so e reductionn si radiotoxicity can be obtained. But through the proliferation / toxicity dilemma, it has been shown that we cannot have both proliferation resistance and negligible toxicity simultaneously. For the combined proliferation and toxicity, the best that can be done is to denature the reactor mixinuraniu1 1: a gmn i ratio n with 238U (thereby ensuring relative proliferation resistance) and reducin radiotoxicite gth factoa (whic0 y yb r«5 h arise dénaturan% s «2 fro e mth t used).

Also, dependin whicn go h energy amplifie useds i r , different isotope levelth f so e 232U will be produced. ThroughT it1s 208 daughter , which is strongly gamma active, one will require shielding in any recycling / fabrication stages. However this gamma activity will also allow one monitoo t r movement materiae th f possiblsd o lan e diversion. We conclude that a thorium based fuel cycle in combination with the family of energy amplifiers does offer a variety of options for improvements in the nuclear fuel cycle with regard o proliferatiot n and/or radiQtoxicity. Further studie e requirear s o clarift d e potentiath y l advantage acceleratof so r driven sub-critical systems over critical systems fro pointe mth f so view of safety, optimisation of power production, and transmutation of particularly hazardous short and long lived fission products. frrrrn ^fff

\ .^— ~- ... % if' X. ^ s \ \ / V s' \ .*•' \ \ V \ \ \ \ \ \ *\v X \ " i i \\ Total —— Total H,jx. j "Plutonium" ...... "Uraniu ING . RADIOTOXICIT Y (Sv/tonne ) m" "Curium" _.._.. "Pluton um" \ \ "Americium" \ \ _._._ "Protac inium" \ \ "Neptunium" r— —— "Thoriu Tl" "Uranium" ...... "Nepturlium" \ O-^tOW-fe-OlO^-JC » 10' _ X i _J > L — J . A DOOOOOOO O 102 103 104 105 106 10' 101 102 103 104 105 106 107 COOLING TIME (years) COOLING TIME (years)

Figure 1. Ingestion radiotoxicity of spent uranium fuel (burnup 40GWd/t) Figure 2. Ingestion radiotoxicity of spent thorium fuel (bumup 40GWd/t) from Thermaa l Energy Amplifier (T-EA). Actinide theid san r daughtere sar fro mThermaa l Energy Amplifier (T-EA). Actinide theid san r daughtere sar 00 grouped (" ") according to the chemical nature at discharge. groupe accordin) " d" ( theigo t r chemical natur discharget ea . 108E ;

10' ; . » • * • \ ^^^% *\ 106 s : f * """s. - '\ 105 \ \ \ ^ \ \ \ \ \ \ ü 10^ \ ———— \- \ \ : x \ \_ o \ 1 o -. 1 V —i - —'- . — *J' Q s I ---.—._ -.—.._..-. \ \ \ ———— Total i 10» "Uraniurri" \ "V Total "Protacti lium" "Uranium" \ ' \ "Plutonium" \ N "Protactinium" 1 10— _ _ — "Thorium" "Thorium" i "Neptuniurn"

•

10' 7 6 10 5 10 4 10 3 10 2 10 1 10 10 101 102 103 104 105 106 107 COOLING TIME (years) COOLING TIME (years)

Figur . eIngestio3 n radiotoxicit spenf yo t thorium fuel (bumup 40GWd/t) Figure 4. Ingestion radiotoxicity of spent thorium fuel (burnup 40GWd/t) from a Pressurised Water Energy Amplifier (PW-EA). Actinides and their fro mFasa t Energy Amplifier (F-EA). Actinide theid san r daughtere sar daughters are grouped (" ") according to the chemical nature at discharge. grouped (" ") according to their chemical nature at discharge. £ 10 "]————I——— Uranium (T-EA)

107

c0) c Thorium (T-EA) 106

o

5 O 10 Thorium (T-EA) Q (recycled)

102 101 102 103 104 105 106 107 10 COOLING TIME (years) COOLING TIME (years) Figur . eCompariso6 ingestioe th f no n radiotoxicitie spenf so t uranium Figur . eCompariso5 ingestioe th f no n radiotoxicitie spenf so t uranium thoriud an m fuel thoriud san m waste from T-Ea A (bumup 40GWd/t). and thorium fuels (burnup 40GWd/t) from a T-EA. The radiotoxicities of The thorium waste results from the removal of 99% uranium and spent thorium fuel from a PW-EA and F-EA are also shown. protactinium from the spent fuel. 108c

10 7 1 2 3 4 5 6 5 1Q 10 4 1Q 1Q3 10 10 10 10 10 10 10 10' COOLING TIME (years) COOLING TIME (years)

Figur . eCompariso7 ingestioe th f no n radiotoxicitie spenf so t thorium Figure 8. Comparison of the ingestion radiotoxicities of spent thorium fuel and waste from a PW-EA (burnup 40GWd/t). The thorium waste fuel and waste from a F-EA (bumup 40GWd/t). The thorium waste results fro remova uraniue m% th protactiniu99 d f mo l an m froe mth results fro remova e uraniumth % protactiniu99 d f mo l an m froe mth spent fuelradiotoxicite Th . spenf yo t T-EA uranium fue shows i l n spent fuelradiotoxicite Th . spenf yo t T-EA uranium fue shows i l n for reference. for reference. O) l (O E g l o Q

CD Z

t.tti i i i «t.MI——t . . t...j—' • • ""*—t i i mm—i t tun 10° 102 103 104 105 106 10r COOLING TIME (years) Figure 9. Ingestion radiotoxicity of spent thorium fuel from a F-EA at burnups of 40, 80 and 160GWd/t.

implosive assembly required

gu1 n type assembly possible

T-EA PW-EA F-EA 239pu+1o/0240pu Figur . Neutroe10 n emission rate spenr sfo t thorium fuel fro T-EAe mth , PW-EAt a F-Ed A an , different burnups reference Th . e implosivvalun a r efo e assembl mixtur% 1 giveys i a f y eno b 24<¥u in 239Pu.

93 10

LD-50 after one 4 10 hour exposure u0) C ra -*-to» TD I 103 0 -*-» CO 104ppm

CO E, til 232U 103ppm 111 CO § 10'

...I 10° t i_ .!_l .Il l i I l l t t l 10° 101 102 103 104

AGEING TIME (days) Figure 11. Dose rate as a fonction of ageing time from 5kg of recycleUd contaminate233 d with 232Tj

REFERENCES

] CARMINATI[1 , KLAPISCH,F. REVOL, R. , , ROCHEP. . J , , CH., RUBIO RUBBIA, A. . ,J , C.^in Energy Amplifier Cleanerfor Inexhaustibleand Nuclear Energy Production Driven Particlea by Beam Accelerator, CERN/AT/93-47, (1993) [2] RUBBIA, C., The Energy Amplifier Concept: A Solid-Phase, Accelerator Driven, Sub- Critical Th/233U Breeder Nuclearfor Energy Production with Minimal Actinide Waste, presente Inte .th t Confda Accelerator-Driven o . n Transmutation Technologied san Applications Vegass La , , July 25-29, (1994). ] RIEF [3 TAKAHASHI, H. , Some, H. , Physics Considerations Actiniden i Fissiond an Product Transmutation, presented at the Int. Conf. on Reactor Physics and Reactor Computations, Tel-Aviv, Jan. 23-26, (1994). [4] INTERNATIONAL COMMISSION FOR RADIOLOGICAL PROTECTION, Annual Limits on Intake of Radionuclides by Workers Based on the 1990 Recommendations, Annals of the ICRP publication 61, Pergamon Press, (1991).

94 ] PEL MAGDLL, [5 D. , Radonn, ,J. SoftwareA - Package r Radionuclidefo Analysis., Institutr efo Transuranium Elements, Karlsruhe, (1994). [6] SERBER, R., The Los Alamos Primer, University of California Press, Oxford, England, 1992 [7] HODDISON, L., HENDRIKSEN, P. W., MEADE, R. A., WESTFALL, C., Critical Assembly: A Technical history of Los Alamos during the Oppenheimer years 1943-45, Cambridge University Press, (1993). [8] CARSON MARK, J., Explosive Properties of Reactor Grade Plutonium, Science & Global Security (19934 , ) 111-128. VOLPIE D Proliferation, , ] ,A. [9 Plutonium Policyd an Institutional/ Technologicald an Impediments to Nuclear Weapons Propagation, Pergamon Policy Studies, Pergamon Press, (1979). SEYFRITZ] 10 [ Nukleare, W. , Sprengkörper Bedrohung- oder Energieversorgunge di r fü Menschheit?, Karl Thiemig Verlag, München (1984). [11] RAMPOLLA MethodofIncreasing, S. . ,D Deterrente th Proliferationo t of Nuclear Fuels, U.S. Paten . 4,344,912tNo , Aug(1982), 17 . .

Next page(s) Left blank 95 MIXED PLUTONIUM-THORIUM FUEL USE IN FAST BREEDING REACTOR SECOLOGICA- E TH F O Y LWA ACCUMULATED PLUTONIUM BURNING

E.Ya. SMETANIN, V.B. PAVLOVICH, G.N. KAZANTSEV, I.Ya. OVCHINNIKOV Institute of Physics and Power Engineering, Obninsk

B.Ya. ZILBERMAN, L.V. SYTNIK Radium Institute, St. Petersburg

Russian Federation

Abstract

The variant of Pu-Th fuel is considered for burning plutonium accumulated in the fast reactor. The possibility of its extraction reprocessin f go fue followee l us component y b d n i reactors s i s addressed.

A wide range of fuel compositions is used at different nuclear reactors. t presenA compositioe th t f uraniuno m dioxide wit requiree hth d enrichmens i t used mostly. Compositions with metal uranium, uranium alloy e knownar s . Tests with uranium carbide and uranium nitride fuel have been performed.

naturaa s i t I l question about Thorium applicatio nuclean i r powerd an , certainly, it may be used in the mixture with plutonium oxide. Such application solve problee sth mf militarcivid o an l y plutonium disposition with simultaneous production of the Uranium-233 isotope. Due to long irradiation, for example in a fast reactor core, the isotope being formed will contain near 100 ppm of LJ- 232 isotope, that exclude militara s a t si y evee material ideus n na o at .

Plutonium-thorium oxide fuel can be manufactured by the technology in many aspects MOX-fuee similath o t r l manufacturing onecase th f thieo .n I s composition homogeneous precipitation of the oxides' mixture from aqueous solutions with following calcination, pellets' pressin d annealingan g alloo wt obtain fuel elements of sufficient strength, radiation resistance and heat conductivity.

97 One of the main questions of such a composition application in nuclear power is a possibility to reprocess it after burning in the reactor.

necessars Ii t o analyzt y e carefully possibilitie f sucso fueha l chemical dissolving with possible variants of the preparatory operations. Does the effect of formation of plutonium oxides solid solutions in thorium oxide take place (as in the case with MOX-fuel)? How does direct dissolving e nitrith occu cn i r acido hydrofluoriD ? c acid additions give considerable acceleration? Do they provide complete fuel breakdown? Is there an catalyse effecth f o t casy t usean e n ?I problem f sucso fueha l breakdown can be solved.

In our opinion it is necessary to pay attention to the development of fuel composition kindf h PuOo sT - ,4 ^wher e PuO« particles with shielding facing, for example Nb, are distributed uniformly in the metal thorium. It allows loweo t workine th r g temperatures considerabl simplifo t d spene yan yth t fuel elements reprocessing. Besides it is proposed to use PuO^ + ZrC fuel compositions with achievement of the required content of PuO^ in the products by means of a porous workpiece from zirconium carbide impregnatio y plutoniunb m nitrate, repeated many timese . th Conten ; % f zirconiu0 o t7 - 0 6 m o carbidt p u s ei workpiece should have giveth e n geometry. Processe f o dryins d an g calcination should follow the impregnation. Under such fuel composition application U-233 won't be produced becaus f thoriueo m absence increasee Th . d heat conductivit d relativelan y y simple technological procedur e productth f o e s manufacturin e alsar go advantage f thiso s composition. The present fuel composition is more simple during the spent fuel elements reprocessing.

Fuel components' separatio d theian nr purification from main fission product complicatee sar d taskssolvee b t the n modery bu ,db yca n methods. The firs to t extractas s i k t U-23 y tributyb 3 l phosphate solutionf o s comparatively low concentration (5 - 10 %), under this process plutonium mus stabilizee b t aqueoudn i s solutio f 3-valenno t form.

98 The following operations, depending on the chosen decision, may provide joint thorium-plutonium extractio y extractionb n methods with their purification froe fissiomth n products e obtaineTh . d solution aftee th r components' ratio correction (additio compensato t n amounn e a th r f efo o t burnt plutonium fuee senth e l b manufacturin o t ty ma ) g operations.

Taking into account thorium nuclear-physical properties (Th-228 isotope appearance) it is also possible to separate these components at one of the stages of the extraction process, in this case plutonium can be directly sent to recycling with a fresh lot of thorium, and the extracted thorium can be stored during several years.

The most reasonable is to recycle the two products with remote manufacturing of fuel elements with biological shield, which provides permissible control radiation levels at working places. While civil plutonium handling it is necessary condition, and thorium addition can only impact on the biological shield thickness. The obtained fuel elements, assemblies must be store factora t plantsd a an ytransported an , shielden di d containers. Such an end of the technological process allows to decrease requirements on the fuel components absolute purification from the fission products at least up to the value of plutonium and thorium own activity.

Technical-economical parameters that follow from plutonium-thorium cycle realization:

1. Civil and military plutonium burning is provided, depending on a reactor capacity casnecessite a th f eo n I . y military plutonium conversio civie th l o nt one may be promoted.

2. Thorium that is widely distributed in nature is involved in the fuel cycle; the technolog extractions it f o y , purificatio manufacturind nan f productgo s from this materia practicalls i l y developed.

3. Uranium-233 isotope is produced, the most safe thermal neutrons reactor may be created on its basis. In the fast reactor core considerable amounts of

99 this isotope with the increased content of U-232 are produced, and in the blanket area more pure uranium-233 is produced. technologe Inth f fueyo l element assemblied san s developmene th r fo t uranium-23 thoriu— 3 m thermal reacto t i shoulr e "fresh"-extracted'b r o d "fresh"-purified uranium-233 usage foreseen thin I .s cas biologicae eth l shield thickness of the technological equipment may be decreased. Such a fuel storage and transportation should be provided with biological shield of warehouse containersd san , remote device r loadinsfo g into reactors.

4. Plutonium burnin fasn gi t reactor uranium-23d an s 3 production will cause reduction of the transuranic elements production — neptunium, americium, curium, and will simplify handling with high radioactive wastes of these reactors spent fuel reprocessing.

0 10 Next page(s) Left blank SESSION 3: CANDU AND LWR ANNIHILATIO PLUTONIUF NO CANDMN I U REACTORS

A.R. DASTUR, D.A. MENELEY AECL CANDU, Mississauga,

R.A. VERRALL AECL Research, Chalk River

Ontario, Canada

Abstract In this pape e show r w that high neutron econom well-thermalizea d yan d neutron spectrum provide major advantages in plutonium annihilation with thermal reactors. High neutron economy reduces core fissile requirement consequentld an s e plutoniuth y m leve dischargen i l d fuel, eliminating reprocessin recyclind gan g cose well-thermalizeTh t d spectrum reduces Pu240 formation by increasing the fission to capture cross section ratio of Pu239. The CANDU reactor which achieves high neutron economy by using heavy water as moderator/coolant and by eliminating reactivity suppression through on-power refuelling, can, in the absence of U238, annihilate 90 % of the initial fissile inventory in a once-through (non-reprocessing) fuel cycle. Recycling the spent fuel (without refabrication) is available as an option to consider eventual storage without reprocessing absence Th . f U23limiteea o s 8ha d effec reacton o t r dynamicd san can be addressed with the use of current reactor technology. A program has commenced at AECL seleco t neutronicalla t y inert materia contaio t l plutoniume nth . Candidates include MgAl2O4, ZrO2, SiC , BeO, Si others initiad n a ,an r l .Fo screenin g process (prio in-reactoo rt r testing), heavy doses of 70 Mev iodine ions are being applied to samples of these materials, as a simulation of in-reactor fission fragment damage. The main objective of this screening process is to find a candidate that does not swell significantly. Alumina has been included in these tests as a reference because it is known that alumina swells significantly due to a transformation to the amorphous phase. Test results to date, and of other tests worldwide, are presented.

1. INTRODUCTION

The end of the cold war has led to the issue of the safe disposition of plutonium from dismantled weapons. Several option beine sar g studied, which include burning plutoniu Mixes ma d Oxide (MOX) fuel in thermal and fast reactors. The main shortcoming of this approach is the production of plutonium from uranium durin irradiatioe andgth X ,MO subsequently f no large th , e fissile plutonium content of the discharged fuel. The latter is especially true of reactors which operate wit hrelativela y high fissile inventor pooo t e r yneutrodu n economy. morA e definitive approac annihilato t s hi plutoniue eth negligiblmo t e dischargelevele th n si d fuel. This option involves burning plutoniu absence th mn i f uraniu eo preveno t m t plutonium formation n thiI . s approac levee hth f plutoniu o l e dischargeth n mi d fue s mori l e strongly dependen neutron o t n economy.

In this paper we present the CANDU as a candidate for plutonium annihilation, describe the basis CANDe foth r U advantage, illustrat advantage eth giviny eb resulte gth studief so identifd san y area f developmenso t

103 2. FISSILE REQUIREMENTS

The CANDU reactor with its excellent neutron economy can operate on core fissile plutonium absence th n uraniumf (i e o g k consequentl0 d inventorie5 )an s a w lo s sa y with thermal neutron flux levels in excess of 1015 n/(cm2-s) . On-power fuel management studies with CANDU fuel consistin f weapono g s grade plutonium, illustrate that plutonium annihilatione b rate n ca s maximize progressively b d y shiftin e fueth gl into higher neutron flux region e fissilth s ea s conten fuee th l f depleteso t . Fissile contensingla n i e % t pas 0 drop9 s y througsb hCANDa U fuel channel seconA . dfuee pasneedes th i l f so reduco dt fissile eth e conten dischargee th f o t d fue o negligiblt l e levels. Complete annihilatio initiae th f lo n fissile inventor , thereforeis y , achievable without having to reprocess and refabricate the discharged fuel.

This abilit CANDf yo Urelates i basie th co dt design feature usinf so g heavy wate moderatos ra r an dfuea l handling system that provide on-powen sa r refuelling capability. Both design features lead to low neutron absorption rates in the CANDU lattice due to the low neutron absorption cross sectio heave th f nyo absence wateth d f burnablran eo e poison reactivitr sfo y suppression. resulta s A , CAND operatee b n Uca d wit hvarietgrada w lo ef yo (lo w fissile content) fuels a s these provide sufficient reactivit CANDa n yi U lattice.

The extent of CANDU's fissile economy is illustrated in Table 1 where the fissile inventory for several reactor concepts is compared for 1 GW(e) cores. The inventory needed for CANDU is between 2 and 4 times lower. This is the case for a reactor that is fuelled with natural uranium.

Tabl e1 FISSILE INVENTORY GW(e 1 (te F )O ) CORES

FBR PWR CANDU CANDB UA 3 to 4 3 o t 2 1 0.05

r plutoniuFo m annihilation e fueth ,l consist a mixture plutoniu f th o s f o e m isotopea n i s neutronically inert matrix. (Several materials for the matrix have been proposed). The absence of U238 eliminate maie sth n sourc plutoniue th f eo m isotopes t als.I o eliminate maie sth n neutron absorber of the lattice resulting in a remarkable improvement in neutron economy. The loss of the neutron absorption in U238, which comprises 30 to 40 % of the neutron absorption in the CANDU lattice, reduces the fissile requirement of the CANDU AB (Actinide Burner) to 50 kg GW(e1 fo a r ) core. (Tabl. e1)

A lower fissile inventory requires a correspondingly higher operating neutron flux level to produc ratee eth d power highee Th . r operating neutron flu CANDxe basie th leve r th fo s i l U superiority in reducing the accumulation of the non-fissile minor actinides. In particular, the on-power refuelling syste ms usei movo dt fuee eth l into region f higheso rs fissil it flu s xa e content depletes during irradiation. Through proper fuel management of this type the last traces of plutoniu eliminatee b n mca d befor fuee dischargeds eth i l .

In order to appreciate the extent of the CANDU flux advantage, the thermal neutron flux levels reactoe th fuee r inth fo l r concept Tablf comparee so ar e1 Tabln di . Note2 e tha flue th tx level CANDe th comparabls i n i B UA thao e t som n i t e FBRs.

104 Table 2 NEUTRON FLUX LEVE FUEN LI L

Reactor Neutron Flux (n/cm2.s)) Type Thermal (0.625 ev) keV0 Fas50 )( t FBR - 0.5E+1 6- l.OE+1 6 PWR 8.0E+13 3.0E+14 CANDU 1.4E+14 0.7E+14 CANDU AB 5.0E+15 0.7E+14

understano T e annihilatiodth n process capture th , fissiod ean n cross section plutoniue th f o s m isotopes in thermal and fast energy spectra must be considered, (Table 3). In a fast spectrum, all the plutonium isotope fissionablee sar fissioe th t n bu ,cros s sectio relativels ni y small; abou3 t barns for 0.8 mev neutrons. In a thermal spectrum, only the isotopes with odd mass numbers are fissionable, but the fission cross sections are large; several hundred bams.

Table3 NEUTRON CROSS SECTIONS (barns)

THERMAL FAST ISOTOPE Oc <*r GC

The annihilation process in CANDU involves first, neutron capture to change from even to odd and then fission by thermal neutrons of the odd numbered isotopes. The high thermal neutron flux level provides an appreciable annihilation rate in spite of the neutron capture required first. Reaction rates (per nuclide plutoniue th f o ) m isotope CANDn si FBRd Uan f currenso t design are compared in Table 4. The reaction rates in CANDU are between one and two orders of magnitude higher.

105 Table 4 COMPARISO REACTIOF NO N RATES

(Relative Units)

FBR PWR CANDB UA ISOTOPE capture fission capture fission capture fission

Pu238 1.25E+16 4.32E+16 1.44E4-15 2.70E+18 9.00E+16

Pu239 l.OOE+14 l.OOE+16 2.15E+16 5.98E+16 1.35E+18 3.74E+18 Pu240 l.OOE+16 2.32E+16 1.45E+18

Pu241 I.OOE+16 2.86E+16 8.09E-H6 1.79E+18 5.06E+18

Pu242 7.50E+15 U2E+15 9.50E+16

6.96E+15 1.57E+16 4.35E+17 9.80E+17 Pu243

. FUELLIN3 G CANDU WITH PLUTONIUM

The absence of U238 has a major impact on the fuel management strategy that is used in the annihilation option fonnatios .A Pu23f no eliminated9s i reactivite ,th lattice th f yeo drops rapidly with fuel burnup e refuellinTh . g rate require maintaio t d n criticalit s significantli y y higher compared with the reactor that burns natural uranium. The refuelling rate is kept within the capabilit Fuee th lf yo Handlin g Syste adjustmeny mb initiae th f o lt fissile conten bundlese th f to . An increas fissiln ei e conten reduco t t refuelline eth g rate, however s detrimentaha , l effectst I . requires reactivity suppression which raises the fissile content of the discharged fuel. It hardens the neutron spectru increased man s Pu240 formation t reduceI . neutroe sth n flux level which increase accumulatioe sth f non-fissilno e actinides. fuee lTh management strategy use , thereforedis balanca , e betwee fissile nth e contene th d an t refuelling rat obtaio et optimue nth m benefit fro fuee mth l handling capability tha availables i t . The optimization includes the option of shuffling the fuel between bundle positions and/or between channels. lattice Th e reactivity drop with irradiatio initian a r nlfo Pu23 eacn 9i contenhg CAND6 7 f o t U fuel bundl s showi e e excesFigurn ni Th . 1 es reactivity average 4.5o t s % oveday0 f 15 ro s irradiation. requiree Thith s i s d valu sustaio t e n reactor operation e rat f fuellineo Th . e th g reacto achievo rt e this fuel lifetim withis ei capabilite nth f currenyo t fuel handling technology.

Tabl e5 BUNDLE PLUTONIUM CONTENT

ISOTOPE CONTENT Pu239 "68.5 Pu240 4.23 Pu241 0.26 Pu242 0.02

106 Wit hbundla e plutonium inventor , (isotopig 3 7 showx f yo cmi Pu23e Tabln ni th , 9e5) content at discharge from a once-through (no shuffling) cycle is reduced by 87 %. The isotopic mix of the discharged bundle is shown in Table 6.

Table 6 ISOTOPI DISCHARGEF O X CMI D BUNDLE

Isotope g Pu239 9.2 Pu240 15.2 Pu241 3.3 Pu242 2.7

Usin once-througe gth shufflingo h(n channe0 ) cycl48 a n eli (950 MW(e)) CANDU annuae th , l reduction of Pu 239 is 1.5 te. The fuelling rate required for this cycle is six fuel channels per FPD. The bundle fissile content is low enough to refuel the channel completely. It is estimated that the Reactor Regulating System is capable of handling this rate of refuelling.

K-i nf i n i t y

120 180 240 300 Irradiation

Pu 239 content

'75.5 +100.9. 8Q 8 Q .. * 122g 5 .1

f transuranlo o 5 35 x n I cmi from PWR spent fuel

Figure 1. Lattice Reactivity Depletion

107 3.1 The Once-Through Cycle

The plutonium remaining in the discharged fuel can be recycled after reprocessing and refabrication into bundles. This is a major cost component of the annihilation process.

The once-through fuel cycle minimizes cost by avoiding reprocessing and refabrication of the fuel. The key requirement for this cycle is to achieve a bundle refuelling rate that maintains the fissile content of the core close to the minimum value of 50 kg. There are two approaches envisage achievdo t e involvethise On . redesigsa Fuee th lf nHandlino g System .les e Thith s s si approacheo desirabltw e th t usei f es o s a technology tha currentls i t y unproven.

In the second approach, the fuel bundle is passed through the reactor once in the outer (low flux) channels that are located towards the core edge and then again through the central channels where the neutron flux is higher. The higher flux compensates for the lower fissile content durin secone gth d pas minimized san reductioe sth powen i r density. This typ fuef eo l .shuffling is required to annihilate the plutonium remaining in the bundle at the end of the first pass through the channel. This approach exploit on-powee sth r refuelling featur f CANDUo e additionae On . l t mosa pas r t s(o two sufficiens i ) reduco t t plutoniue eth m conten dischargee th f o t d bundlo et levea l tha comparabls i t lossee th o est experience fuen di l reprocessing.

The additional pass would entail reservation of specific fuel channels in the core for the irradiatio seconf no d pass bundle increasn a fuellin e d th san n e i g machine expecte s i dut t I y. d that some modifications wil requiree b l desigFuee e th th lf o d t nHandlino g Syste achievmo t e this capability. However, the design concept of the system will remain unchanged.

4. REACTOR DYNAMICS AND CONTROL

There are two major reactivity feedback effects that are attributed to U238 in CANDU. Coolant voiding fro mfuea l channel reduces resonance absorptio U23n n i increased 8an fase sth t fission rate. This produces positive reactivity. maiU23e th ns 8i contributo negative th o t r e feedback from doppler broadening of the cross section resonances on fuel temperature increase. As expected, both effects are reduced in the absence of U238.

Tabl e7 REACTIVITY COEFFICIENT CANDA R SUFO LATTICE

Bundle Plutonium Content Natural 76 g 100 g 122g (Current CANDU) Fuel Temperature -1.1 -1.1 -1.1 -4.5 Coefficient (micro-k/De) gC Coolant Void Reactivity +2.6 +2.3 +2.0 +11.0 (milli-k)

108 The coolant void reactivity coefficient is near-zero. The fuel temperature reactivity coefficient is negative and near zero, (Table 7). The power coefficient of reactivity is negative or zero throughou fuee e poweth t lth lifes rA . coefficien f reactivito t s neayi currentle r th zer r ofo y operating CANDU reactors, implicatiothero n s e i Reacto e th n no r Regulating Syste usinf mo g this typ f fueleo .

Table 8 TEMPERATURC EFFEC° 0 10 F TO E RIS PLUTONIUN EO M REACTION RATES (xlOOO)

Isotope Chang Neutron ei n Change in Neutron Difference Yield Absorption Pu239 -0.15 -0.074 -0.076 Pu240 0.0 0.135 -0.135 Pu241 -0.029 -0.0012 -0.028 Total -0.179 0.060 -0.239

The negative value of the fuel temperature coefficient in the absence of U238 warrants some explanation plutoniue Th . m isotopes hav negativea e contributio lattice th o net fuel temperature coefficient. This behaviour is due to the shift, on fuel temperature rise, in the temperature of the thermal neutron spectrum and the resulting change in the cross sections of Pu239 and of Pu241.

Table 8 shows the behaviour of the reaction rates of plutonium isotopes with fuel temperature rise. Pu239 and Pu241 reaction rates decrease with fuel temperature. But the drop in neutron yiel larges di r tha droe nth neutron pi n absorption.

There is a significant reduction in the delayed neutron fraction relative to the natural uranium CANDU. The value for this lattice is between 3.0 and 3.8 mk compared with 6 mk. However, the neutron generation time for the lattice is significantly longer (2 ms) and the discontinuity betwee sub-prompe nth super-prompd an t t neutron transien relativels i t y small.

5. DEVELOPMENT REQUIREMENTS

This study f hasyeto t uncovere s ,,no a neutroniy dan c issues that woulda prevene us e th t plutonium-based non-fertile fuel in CANDU. It is expected that with current fuel handling technology e fuellinth , g strategie f thio s s e thae e envisageb typus ar tf fuen e o e ca th l r dfo implemented. This makes this plutonium annihilation strategy feasible with the current CANDU concept. maie Th n develoo tast s ki pneutronicalla y inert matrix tha compatibls i t e wit actinide hth x emi in a reactor environment at power densities that are currently achievable. If the CANDU is to operat t currena e t power densities e fasth ,t neutron flux level, which e affectth e liff th o es CANDU plant remains unchanged.

109 6. FUEL DESIGN

Selectio f Fueno l Inert Matrix Materia r Carrieo l r Material

e objectivTh fino t suitablda s ei e inert matrix materia r carrieo l rs materiaIt . Pu holo t le dth function is to dilute the Pu, since pure Pu or PuO2 would be much too highly fissile, and to not produce new Pu during the irradiation (as UO2 does). The following sections identify the properties required for such a carrier material, list and assess a selection of materials that appear the most promising, and discuss work begun at CRL and elsewhere on this program.

6.1. Properties Require Carriea r dFo r Material

Neutronics: All elements in the carrier material must have low neutron absorption.

Melting Temperature: A high melting temperature is required so that fuel operating temperatures beloe ar w approximatel meltine th f o g5 temperatury0. e (Kelvin). This ensures that thermally activated processes, such as fission gas diffusion, are acceptably slow. A high melting temperature also ensures that there is an acceptable safety margin between fuel operating temperature and melting temperature, to accommodate potential off-normal excursions.

Compatibility with Coolant and Clad: Chemical compatibility with water at coolant temperatures is a requirement. Compatibility with the clad is also required, although the clad need not necessaril zirconiue yb zirconiua r mo m alloy.

Phase Stability: Phase transformations as the fuel cycles up and down in temperature are unacceptable if there is a volume change associated with the transformation or if the material loses structural integrity. Similarly losa , crystallinityf so , thus formin amorphoun ga s phases a , the fuel is irradiated, is unacceptable if it leads to similar problems. This latter point is one of the aims of the test program outlined below.

Thermal Conductivity: High thermal conductivity is advantageous, since this allows low fuel operating temperature (for a given element power rating). This provides increased safety margins carriee th n i r material. Thermal conductivit preferablleast d ya higan s ta , has y higher, 2 thanUO is required.

Solid Solubility with Pu: In order to disperse the Pu on an atomic scale in the carrier material, the carrier should be able to dissolve 1-3% Pu. In contrast, MOX fuel with small panicles of PuO2 in a UO2 matrix have been used successfully in LWR power reactors, but reactor operators conservative b o t hav burnue d th eha n ei p fuele limitth ,f so unti paniculate effece th th l f o t e naturfuee th l f undeeo r irradiatio establisheds ni . Note, however matri2 ,abilitys UO thaxe ha th t dissolvo t e PuO onls i t 2yi ; thafabricatioe th t n processe productioe s th use r dfo f (U,Pu)Ono 2 hav t beeeno n sufficiently advance produco dt ea perfec t solid solution. Similarlye b y ma t i , acceptabl produco et dispersioea f finno e particle candidate PuOf th so n i 2 e carrier materials considered here.

Irradiation Properties: Good irradiation properties woul requirede db demonstrated an , y db extensive in-reactor testing.

6.2. Candidates

A list of candidate carrier materials for the Pu includes Zr02, BeO, MgAl2O4, CeO2, SiC and ThO2. The last material listed, ThO2, is included because no new Pu is produced, even though

110 its irradiation produces new fissile material, U. Thus, use of ThO to hold the Pu would 233 2 achiev goae eeffectivelth f lo y destroyin , howeverIf . gPu subsidiara , y goa lproduct werno o et e any new fissile material, ThO2 would not be an acceptable candidate.

Since ther considerabla s ei e bas f irradiatioeo n experienc ThOn eo t 2wouli , d clearle th e yb simples candidatee th carriea f s o t a furthee b e rt qualifwilo I t st materia us . no lrr Pu y fo r fo l discusse followinge th n di .

Other candidates considered were Alp suicidd an 3 e dispersion fuels. Alumina (A1s 2O3wa ) rejected because it has long been known that it becomes amorphous with a volume change of whe% 30 n usefuela s da . Silicide dispersion fuels were rejected becaus l matrie A th e f xeth o dispersion fuels has too low a melting temperature, 60CTC, for safety considerations. metald an Alloyu sP werf o s e considered r examplfo , r alloyeZ e n apparentldA wit . Pu h y promising candidate is Si, as it is cheap, abundant, and has good thermal properties. Also, there is irradiation experience with uranium suicides (larger amounts of uranium that required of plutonium for this purpose). This candidate is retained as a representative of metallic alloys and suicides.

6.3. Properties of the Candidate Materials

Melting temperatures, thermal conductivities, and heat capacities of the candidate materials are give Tabln ni . e9

Table 9 MELTING TEMPERATURES AND THERMAL CONDUCTIVITIES OF CARRIER MATERIALS

MATERIA Melting Thermal Conductivity Heat Capacity1 L Temperature (W/m-K) (J/cm3-K) (J/g-K) <°C) 100'C 1000°C

ZrO2 2715 1.9 2.3 2.55 0.46 BeO 2530 220.0 20.0 3.07 1.02

MgAl204 2135 10.7

CeO2 2600 10.9 2.55 0.36 SiC 2700 85.0 2.15 0.67 Si 1410 108.0 1.65 0.71

U02 2878 8.75 3.21 2.58 0.24

Notes:

The heat capacity per unit volume is more relevant, for calculating temperature increases for a given inpu heatf o t , tha unitr npe weight becaus volume fuee th th l f elemeneo t wil e constanlb t whateve carriee th r r material.

Ill Zirconia has a high melting temperature and good neutronic properties (absorption cross section of 0.184 B for thermal neutrons) as long as hafnium, a persistent impurity is carefully removed. Commercia nucleat lbu grad, rcontainHf r gradeZ essentialls % i r 3 eZ 1- s y pure (-5 Hf)m 0.pp

Thermal conductivity increases with temperature, in contrast to most materials. However, it is probably too low, being even lower than for UO2, which has thermal conductivity values of 9 W/m-K at 100°C and 3 W/m-K at 1000'C.

It is insoluble in water, and should be compatible with Zircaloy sheath. Phase stability is a potential issue. Zirconia undergoes a phase transformations with large associated volume changes at 900-1000°C (monoclinic to tetragonal). This phase transformation causes crackin would gan d eventually disintegrat fuele eth , after repeated temperature cycling throug phase hth e transition temperature. However, addition f yttriumo s , calcium, magnesium or possibly other materials will stabilize the high temperature cubic phase of zirconia. (This cubic phase is the fluorite structure, the same as UO2 PuO2, and ThO2.) Approximately, 8 at% of the additive is required for full stabilization, but smaller amounts provide partial stabilization. fracture Th e toughness that this stabilization provides makes this materia intenseln a l y studied materia generae th n i l l ceramics field. Thus e databasth , s goodi e therma Th . l neutron absorption cross sections of these additives are not large (Y, 1.28 barns; Ca, 0.43 barns; Mg, 0.63 barns) o whils , e providin n extra g a degre f complexito e n fabricatioi y d possiblan n n i y irradiation, maintains this materia gooa s a dl candidate.

Thus, zirconia* poos r thermal conductivit moss it s yti serious drawback fuee lTh .woul d have to run at lower powers than urania. Other, more acceptable candidates are sought vere Th y high thermal conductivit f BeOyo , combined wit hhiga h melting temperaturee ar , strong advantages for this candidate. The conductivity does decrease rapidly with temperature, but is still high at 1000°C and the fuel operating temperature is determined more by its low temperature conductivity (above coolant temperature) tha higs nit h temperature vervalues ha y t I . low neutronic absorption (0.008 barns for thermal neutrons) and is a neutron moderator. It does not undergo phase transitions (hexagonal ove whols it r e temperature).

A potential drawback to BeO is its toxicity. The hazard is highest when using powders, as fabrication would probably require. However, toxisincs i u ceP already extre th , a toxicitO Be f yo shoul acceptablee db . Unknow whethes ni materiae rth l would for msolia d solution wit. hPu woultoxicite O Th Be df yo make laboratory testing more difficult.

Spinel (MgAl2O4) is a candidate suggested by the Europeans. The French, supported by the Transuranium Institute at Karlsruhe, Germany) are actively considering it for an inert matrix (carrier) material for both LWR and Fast Breeder Reactors, and irradiations are underway in the fast reactor PHENIX. An advantage is its perceived resistance to swelling in-reactor. The argument is that for swelling to nucleate, a number of vacancies must congregate together. Furthermore vacancieO d an , , stoichiometricthern si AI , e musMg e b t proportions, i.e. ration ,i s of 1:2:4. Sinc probabilite eth y that this will happen randoml smalls yi materiae th , resistans i l t to swelling. However, its thermal conductivity is only marginally higher than that for UO2. Melting point is sufficiently high, 2135°C. Thus, it remains a promising candidate as long as thermal conductivity is not an overriding concern.

1 CeO considereds i 2 becaus s crystait e l structur s ei identica UOo t l s goo2. ha Sincd2 eUO properties, excep r thermafo t l conductivity, and, especially, good irradiation performances i t i ,

suggested by Hj. Matzke, Transuranium Institute, FRG.

112 thought that this material could have similar good properties thermas It . l conductivit abous yi t 10.9 W/m-K at 100°C, slightly higher than that for UO2, and its melting temperature is very high, 2600°C. Because of its close similarity with UO2, this is a very promising candidate.

Silicon is abundant, cheap, and has good thermal properties. Also, there is irradiation experience with uranium suicides (with larger amount f uraniuo s m tha f plutoniuo n m require r thidfo s purpose). Silicon is generally relatively inert, although it is attacked by dilute alkalis. Most acids do not affect it. Presumably, control of the coolant pH, together with the use of a Zr sheath would neutralize concerns about coolante attacth y kb .

The use of suicide fuels (U3Si and U3Si2) for CANDU was investigated in the 1970's by AECL. They wer t selecteeno d because they softene swelled dan d abov certaiea n temperature, which depended on the bumup, and there was a reactor control problem, associated with the low heat capacity. These concerns would re-addressee neeb o dt puree th r r dfo silicon-plutoniu m fuel.

Silicon carbide is a material with a considerable industrial base of use, and, thus also, a large data bas propertiesn eo , especially compare otheo dt r candidates considere dverya aboves ha t I . high thermal conductivit higa d hy an meltin g temperaturew . lo Bot e ar h , elementsC d an i S , neutron absorbers wilneutroa C s d l a als an ,t noac moderator. Although compatible with water at low temperatures, it must be tested for compatibility with water at reactor coolant temperatures, 300°C and above. Fabrication methods have been developed for industrial application, although these may require modification to incorporate Pu. As long as this material compatibls i e with wate t higa r h temperatures excellent n appeari ,a e b o st t candidate.

6.4. Irradiation Experience

The only program to test carrier materials in-reactor, with a fissile component to introduce fission damag carriee th o et r material Bettia s swa , progra earle mth y[1-3n i 1960'] seleco st materiata l fofull-cora r e loadin Shippingpore th f go t test reactor. They tested A^Oj, ZrO2, Zr02+CaO, BeO and ZrSiO4, all containing large amounts of UO2 (>20 wL %). Only ZrO2 behaved

satisfactoril themr yfo : A1O became amorphous after onl verya y short tim reactorn ei O Be ; 3

exhibited high swellin2 g (from fissio s bubblga n matriO e Be formation xe th materia d an ) l "suffered considerable fission fragment damage". The authors of the Bettis reports suggest that e cubith s ci crystat i l structur f ZrOo e 2 which provide resistancs it s damageo et . This wa s supported, for example, by the results on the candidate material ZrO2+CaO+UO2, which existed as a two-phase material, one phase cubic and one non-cubic. The cubic phase showed resistance to damage, whereas the non-cubic phase did not. A12O3, BeO, and ZrSiO4 are all non-cubic and did not test well. However, CeO2, MgAl2O4, and SiC, candidates listed above but not tested by Bettis l foral , m cubic crystal structures. Therefore, ther reasos ei expeco nt morr to thaf e eo on t these candidates wil acceptable b l carriea s ea r CANDU n materiai u P r fo l.

The Bettis results show that ZrO might make a good carrier material for our purposes.

However, it is the least desirable 2 of the suggested candidates from a thermal conductivity consideration. Therefore prograa , f developinmo gcarriea rCANDn i materia u P s r Uha fo l begun, focusin candidatee g th firs n o t s CeO2, MgAl 2SiCd O4.an ,

French irradiations of selected candidate materials in PHENIX and a thermal reactor, although not containing fissile material, represen majoa t r investment towards findin carriega r material, although their primary purpose is not to burn Pu, but rather to burn selected undesirable actinide wastes, separated from fuelspenR .LW t

113 6.5. AECL Test Program o selecT a carriet r material froe abovmth e candidates, accelerator e use b r initia fo dn ca sl screening, as described below. This must be followed by a substantial in-reactor irradiation program (with dissolved fissile material, althoug necessarilt hno initially)u yP . Simultaneously, an out-reactor properties-determination program mus undertakee b t establiso nt h that there ear no other deleterious properties of the candidate material, such as reaction with water or the selected clad material.

Usin Chale gth k River Laboratories (AECL) Tandem Accelerator iodinV Me e 0 beaa , 7 f mo ions, incident upon the candidate materials, are used to simulate fission fragments. This will provide a first indication regarding their suitability for use as a fuel. The work has only begun. dateo T , spine alumind an l a (single crystals) have been tested onlt t roobu ,y a m temperaturen I . the alumina specimen, swellin apparens gwa spine e Th t l sampl currentls ei y being analyzed. Also furnaca , bees ha en built thao s , t specimen testee b t temperaturen da sca s more relevano t operating samplefuelC Si . s have been obtaine testingr dfo .

SUMMARY

Because of their high neutron economy, CANDU reactors can operate with minimal fissile content This is especially so in the absence of uranium, i.e. when separated plutonium is used as fuel. The low fissile requirement and the on-power refuelling capability of CANDU can be exploited to achieve high annihilation rates and minimal fissile content in discharged fuel. This minimizes reprocessing and refabrication cost.

The absence of U238 reduces the coolant void reactivity which is the major positive feedback reactivity componen existinn i t g CANDU reactors.

The well-thermalized spectrum maintains a negative fuel temperature coefficient by minimizing resonance absorptio Pu239n ni .

These features give CANDU reactors a strategic role in plutonium annihilation.

A progra develomo t psuitabla e inert materia hol o plutoniut le dth munderwas i t AECya L

114 PLUTONIUM BURNIN THERMAA GVI L FISSION IN UNCONVENTIONAL MATRICES

C. LOMBARDI, A. MAZZOLA* Departmen f Nucleao t r Engineering, Politecnico di Milano, Milan, Italy

Abstract plutoniue Tu m coming from dismantled warhead thad san t already stockpiled from com- mercial fuel reprocessing have raised many proposals for their burning in a safe and eco- nomical manner. This paper concerns the utilization of current PWRs partially fed with a non-fertile oxide-type fuel, the rest of the core being still fed with standard U-235 enriched fuelunconventionae Th . l fuel consist PuC>f so 2 dilute inern a n dti matrix, which is highly radiation resistant and scarcely neutron absorbent. The matrix may be ZrC>2,

A1O, MgO or their proper combinations. To verify the reliability and validity of the 3

results2 , two different computer programmes were been utilized: first single pin cell calcu- lations were performed using the WIMS-E program to assess the physical and neutronic features of the new fuel, then the CASMO-3 program was employed to study the fuel assembly in its actual layout. Commercial PWRs operating in a once-through cycle scheme can transmute more than 97% of Pu-239 and more than 71% of total initially loaded civi weapons-gradr o l e plutonium dischargee Th . d plutoniu conditioy an n i ms ni definitely proliferation resistant hige hTh . initial reactivity cause higsa h initial power peak, and a continuously decreasing power generation during the irradiation. The rela- tively high thermal conductivity of A1O3 compared to UÛ2 one and the use of burnable

poisons should allow fuethi facw o t ls ne e this latter effect2 lesA . s pronounced power peak tacklede b o t s . ha UC>n Reactivit i L 2 rodEO t a sy coefficients are absolut,n i e value, slightly lowe reactoe onesth 2 t r thaU0 r,bu e dynaminth c behaviour shouldiffere b t d-no t froen mMOX-fuelle operatine % thath 30 f o to t 0 g2 d cores conclusionn I . t seemi , s that once-througe th h cycle burnin plutoniuf go PWRmn i non-fertila svi e matrice intern a s si - esting solution for both weapons-grade and civil plutonium disposal, worth further inves- tigation.

. INTRODUCTIO1 N

recene th n I t years e excesth , s plutonium coming from dismantled warhead s prosha - motenucleae th n di r communit extensivn ya e debate: several solution disposinr sfo f go weapons-grade plutonium have been so far proposed and discussed from the viewpoints of safeguards, proliferation resistance, environmental safety, technological background, econom timd yan e schedule [1-4] resule Th man.f o t y published studies seem thae b to st fission options such as Light Water Reactors (LWRs) with mixed-oxide (MOX) fuels, Liq- uid Metal-Cooled Reactors (LMRs) with metal fuels, High-Temperature Gas-Cooled Reac- tors (HTGRs) with coated particle fuels Fasd an ,t Breeding Reactors (FBRs) withoue th t fertile blankee mosth te efficienar t effectivd an t e way disposo st surpluf eo s weapons- grade plutonium thin I . s framework worte b y tako ht ma e t situatioe i stoc, th f ko n about the present utilization of plutonium in commercial power plants and about its overall role in.the fuel cycle. Plutoniu always mha s been viewe advantageoun a s da s material, suitabl substituta s ea e for uranium-235 and therefore many studies have been carried out aiming at improving conversioe th n rati nuclean oi r reactorgenerato t s a o ss e plutoniu mt higa he ratth r efo

* seconde t OECD-Haldeda n Reactor Project 173, X P.O, N-1751BO . , Halden, Norway

115 following utilizatio FBRsn i . These reactor designee sar obtaio dt positivna e breeding gain, thus allowin gbettea r utilizatio uraniuf no thae m or LWRsn i fact n energe I . th , y extracted from a given uranium quantity results to be amplified by approximately a fac- hundrede on r to . Nowadays, on the other hand, the present situation points to the need of identifying solu- tions for plutonium burning in a safe and economical manner. At present, the introduc- tion of FBRs has been shifted into the far future because of economic, technical and political reasons. Besides, the considerable decreasing of uranium prices, the lower than expected uranium e highedemanth d r an davailabilit f naturayo l uranium resources rende e concernrth s about fissile material supplie longeo sn criticaa r l issue. Moreover, large quantities of plutonium recovered from reprocessed 110% fuel have been already stockpiled and over the next decade, under already ratified agreements, significant quan- tities of weapons-grade plutonium are expected to be available (about 100 metric tonnes). All these considerations, together with proliferation concerns, indicate that the pluto- nium question migh face e differena tb n di t way. Therefore t seemi , s possibl affiro et m tha even a t r growing plutonium production should not be any longer the goal, and hence the already stored quantities, those continuously produced from operating reactors and those obtained from weapons dismantlement have highesburne te ob th t a t t possible rate wit bese hth t economic benefit, avoiding, during their elimination, any further plutonium production. This proposal is primarily referable weapons-grado t e plutonium, havin meeo g t more th t e stringen proliferation no t d nan safety requirements, but has to be considered suitable for civil plutonium as well.

2. A NON-FERTILE FUEL CONCEPT

A variet differenf yo t technical solution plutoniur sfo m burning have proposer beefa o ns d and investigated. This paper concerns specifically the utilization of conventional PWRs partially fed with a new non-fertile oxide-type fuel. This combination could be an attrac- tive optio dispositior nfo plutoniumf no utilizatioe Th . heavf no y water reactor undes si r stud wells yrelevana d ,an t results wil presentee lb furthea n di r paper. proposee Th d solutio bees nha n suggeste Authore th y db successivel d 199n si an 2 t ypu forwar mann di y occasions [5-8] t seemI . s that other people share this proposal, particu- larl Japayn i n (see here below). To avoid any plutonium production during the irradiation, uranium-238 is completely absent in this special fuel. The fissile component, in PuO% format, is spatially diluted in an inert matrix which is highly resistant to radiation, scarcely neutron absorbent and has good thermo-physical properties. Chemical formats of the non-fertile matrix would be A12O3, ZrO2, MgO or their proper combinations. If needed, ThO2 may also be considered to improve the fuel neutronic characteristics, but in this case, for the proliferation resist- ance of the spent fuel, it may be necessary to add some uranium-238 to dilute the ura- nium-233 coming from breeding of thorium-232. These fuels have substantially different characteristics tha usuae nth l MOX fuels becaus differene th f eo t compositio particd nan - ularly of the absence of fertile nuclides. Recentl yJAERa I research group publishe resulte dth s o studfa whicyn i h this oxide-type fues beeha l n considere dworts i [9]t I briefl. o ht y summariz maie eth n results proe -Th . posed oxide fuel tailor-made ar s e multiphase fuels consistin mineraf go l compoundsd an , have been produce reactioe th y db n between plutoniu somd man e additives. Froe mth phase relations of ceramic materials and their chemical properties, it seems that a two- phase mixture of a fluorite-type phase and alumina (suitable for stabilizing the spent fuel) has favourable characteristics for plutonium burning. It also seems that the fluorite- type phase suc thoris ha full d aan y stabilized zirconi acceptable aar hoss ea t phasef so plutonium because of the high solid solubility of the actinide elements and fission prod- ucts irradiatioe th , n stabilit chemicae th d yan l stability spene Th . t fuels finally obtained

116 from both systems PuC^-ThC^-A^O PuOd an s 2-ZrO2-Al2O3 will become mineral-like waste forms which would have chemical stability, high neutron emission rate, high heat- ing rate and 'be quite radioactive, making it unattractive for subsequent use in weapons. The JAERI investigation demonstrates that this solution is technologically viable and could combine a high plutonium burning rate with a chemically stable form of the waste. Whatever the origin of plutonium, several technologies are fit for fuel fabrication: for example co-precipitation of fissile with matrix elements from solutions in a process lead- ing to mixed oxides, and absorption, selective or not, of fissile ionic compounds of suitable porous carriers, followed by heat treatments or similar techniques. It is worth noticing that the new oxide-type fuel should have a higher thermal conductiv- ity than the standard UC>2 one. This improves the heat transfer process from the fuel to the coolant, thus leading to lower centre-line temperatures. The fuel temperature may be kept either low to limit fission gas release and then the rod internal pressure and the cladding stress corrosion, or equal to the values of the usual fuel so the new plutonium- type fuel could sustain a higher power generation. Besides, a high thermal conductivity would allow the fuel to have short time constants, which is an important characteristic during transients. Single pin cell together with preliminary assembly-level neutronic calculations have been recently presented [8] and are here only summarized. This paper describes a further developmen neutronie th f to c stud presentiny yb g more detailed assembly-level results.

3. REACTOR CORE DESIGN

Reactors involved in plutonium incineration would be PWRs operating in a once-through cycle scheme plutonium-pooe Th . r spent fuel, handle high-radioactiva s da e waste, could be buried under deep geological formations without further reprocessing. Three different core configuration considerede b y ma s : core th totalls e i . 1 witd yfe h non-fertile fuels; 2suitabla . e fractio e cor th loades ei f o n d with special assemblies containinw ne e gth oxide-type fuels residuae th , l part being stilwitd lfe h standar assemblies2 dUO ; core 3 th .entirels ei y loaded with assemblies containin plutonium-oxidd gan bot2 hUO e fuel rods in proper proportions. firse Th t solution woul vere db y interestin attractivd gan e becaus t woulei d alloe wth larger plutonium consumption rate. Nevertheless, the dynamic characteristics of the inert fuel could lea considerablo dt e modification currene th f so t control systems design. factn I plutoniue th , m delayed neutron fractio lowes i nß r (0.21 plutonium-239r %fo ) than uraniue th (0.65e mon uranium-235)r %fo absence th d uranium-238f an ,e o , strong reso- nance absorber, decreases the fuel temperature coefficient (in absolute value). Moreover, e plutoniuth m nuclear parameters (for example d epithermafasan t l cross sections) reduce absolutn ,i e term voie ,th d reactivity coefficien moderatoe th d an t r reactivity coef- ficien wells a t . secone Th large d th solutioe experiencus n nca e of operating MOX-fuelled cores suita.A - ble arrangement of the special fuel assemblies in the core and a proper reloading strategy might reduc powee eth r peak flatted san flu e nth x radial profile. Anyhow locae th , l behav- iour during both operationa accidentad lan l transient carefulle b o t s ysha investigated. thire Th d optio easiese n th seem e studyb o t o st . Some plutonium-typ e30%o t rod 0 )s(2 are inserted in the standard fuel assembly without any change to the present reloading strateg currenf yo t PWRs overale loca.Th d an l l reactor behaviour woul drivee e db th y nb standar fueonld 2 an ly d UO limite d perturbation woul inducee db inere th y td b fuel . partiae Th l loadin plutoniuf go m fuel envisages a , alss i solution, n doi 3 coherend an s2 t wit hscenaria whicon i hplutonium-loadea d fraction core ofth e could bur plutoniue nth m produce previoua n di s fuel standare batcth y hb d UO2-loaded fractio equilibriue th t na m cycle.

117 The third solutio choses stude n wa investigatthir o th t nf fo s y o i analysism ai e Th . whether a high plutonium consumption rate may be achieved in current PWRs without any major modification to the reactor design. The resulting spent fuel should be pluto- nium-poor and proliferation resistant. The reference reacto 3.2n a %s ri enriched AP600-type one reactoe .Th r main parameters summarizee ar Tabln di . eI

TABLE I: REFERENCE REACTOR MAIN PARAMETERS pellet diameter 8.2mm cladding outer diameter 9.5 mm pitchpin 12.6 mm

pU02 95%ofT.D. average linear heat rate 13.5 kW/m assembly layout 17X17 fuelno. rods assembly1 264

reference Th eassemble fueth s i l y define Tabln d i whic n i , erodI 6 h5 s ove4 totara 26 f lo are substituted with the new-fuel rods (see Figure 1). Two plutonium isotopic composi- tions were considered: civil plutoniu military-gradd man e plutonium e plutoniuTh . m compositions are detailed in Table II. The plutonium mass content is an important parameter. Assumin reference th s ga e valu uranium-23e eth 5 mass containe standa n di - ard 3.2% enriched UC>2 fuel (Fissile Uranium, FU), we adopted fissile plutonium masses (plutonium-239+plutonium-241) in each rod equal to 0.71 FU, 1.0 FU and 1.3 FU respec- tively. This interval was chosen according to the following considerations relevant to civil plutonium: 0.7 implieU 1F s tha totae tth l plutoniu enableU F m 0 mas1. s ; equas si FU o lt a direct comparison with the same initial fissile mass; 1.3 FU is one of the values adopted fuelX . MO a n i

TABL : PLUTONIUEII M ISOTOPIC COMPOSITIONS

isotope military-grade civil Pu-239 % wt 3 9 58 wt% Pu-240 6wt% 24 wt% Pu-241 % wt 8 0. 13 wt% Pu-242 0.2 wt% 5wt%

In any condition the remaining empty volume was completely filled with A^Os, regard- les technologicaf so l choices t tendin bu ,conservativ e b o gt termn ei neutrof so n absorp- tions. A variant envisages the use of burnable poisons under the form of gadolinia (Gd2O erbir 3)o a (Er2O3).

118 4. STUDY METHODOLOGY

To verify the results reliability and validity, two different computer programmes were uti- lized t firsA . t singl celn epi l calculations were performe usiny db WIMS- e gth E program [10] and its 69-groups '1986' library [11] to assess the physical and neutronic characteris- tics of the new fuel. The results are detailed in Reference [8]. In this paper the CASMO-3 code (version 4.7) [12employes ]wa studo dt fuee yth l assembl actuas it yn i l layout. CASMO is a multigroup 2-dimensional transport program for burnup calculations on fuel assemblies prograe .Th m handle geometrsa y consistin cylindricaf go l fuel rod varyinf so g composition in a square pitch array with allowance for fuel rods loaded with gadolinium and erbium. Nuclear data are collected in a library containing microscopic cross sections in 40 energy groups (E4LTJB4 library). Neutron energies cover the range 0 to 10 MeV. Since we limit our study to assembly-level fuel depletion calculations, we adopt a reactiv- ity-based cycling model [13 determino ]t equilibriun ea m cycl whole-core eacr th efo f ho e compositions under study. A realistic model of a particular reactor involves a detailed description of the neutron and fuel behaviour in position, energy and time as well. In fact, flux and power distributions vary between the fresh and the old fuel and such effects should be taken into account with a detailed analysis covering spatial and energy depend- ence. However e essentiath , l feature fuef o s l behaviou investigatee b n ca r usiny db ga simple model neglecting this detail, assuming the time is the only independent variable. The simplifying assumption is that fuel batches of different exposures combined in the same core contribute to the overall reactivity in proportion to the reactivity in similar sized reactors, operating in a 1-batch cycle, loaded with fuel having an uniform exposure. Then, considering a reactor with M fuel batches, each having a different burnup history, one might introduc partiae eth l infinite multiplication facto batcf ro h(£/i : )as

where k^ is the infinite reactivity in a 1-batch operated reactor subject to fuel burnup T*. Then,_the M-batch operated reactor in this model has an overall infinite multiplication factor k:

M / \ M l

k-infinitL e equilibriuEO th e f yo Th m cycl s assumeewa d equa 1.06o t l 6tako t e into account both the leakage and additional parasitic absorptions. With this value the stand- d 3.2ar % enriched AP600 core reache sdischarga e equilibrium burnu f 33.po 1 MWd/ e cyclth kgHMj e9 timt Le .e measure conveniena n di t unit. Effective Full Power Days (EFPD) were chosen to measure the irradiation time instead of the usual MWd/kgHM because differene ofth t heavy metal conten conventionae th n ti speciae th d llan fuel rods. above Th e conditio overale th expressee n cyclnb h o lL fn nt reactiviteca EO : e das th r yfo

M

i= 1

beginnine Atth nexe th f tg o cycl e (BOCn+1 longese )th t irradiated batc removes hi d dan a fresh batch is loaded. If we assume convergence, i.e.

119 „ e = „ n —>o=

independen , thenn sufficientlf r o t fo , y discharge largth , en e burnu equilibriue th f po m cycl 6*=MQs ei SS. Therefore conditioC EO e nth , becomes:

M *(«e„) =1.066 n= 1

which can be used for determining 9*.

5. SINGLE PIN CELL CALCULATIONS

preliminare th n I y phase, detaile Referencn di e [8]were w , e intereste assessindn i g some neutronic features of this new fuel, considered as a single rod separated from the assem- bly. Features of the standard AP600 fuel rod are kept as a reference. WIMS-e Th E singl calculationn epi s wer same e th carrien eo fuet dlou compositions here adopted with the exception of a further variant implying the addition of ThO2 filling 50% free ofth e volum rode depletioe th f .eTh o n calculations were done wit assumptioe hth f no imposing a thermal flux equal to that obtained in standard UO2 fuel irradiated at con- stant power. This means to assume in these calculations that the presence of the special rods induces only small perturbation of the flux distribution inside the fuel assembly. The obtained results are here summarized. new-type 1Th . e fuel seem havo s t capabilite eth almosn a f yo t complete eliminatiof no plutonium-239 (>97%), both for civil and military-grade plutonium. The plutonium- poor spent fuel contains large fractions of plutonium-240 and plutonium-242, thus ren- dering it completely proliferation resistant. 2.. Plutonium fuel rod shows a high initial reactivity, followed by a steep decrease versus EFPD (Figure 2). 3. Temperature and void reactivity coefficients are much lower, in absolute value, than for standard fuel, reaching positiv military-gradr fo e L valueEO t sa e plutonium (Fig- . 5) d an ure4 , s3 4. Thoria bearing fuel behaves in a more similar way to UC^ fuel, but it shows a lower reactivity (Figure 2). Some concerns about the proliferation potential of the spent fuel e uranium-23th o t exise du t 3 build-up e initiaTh . l reactivity coul e increaseb d d through a larger initial plutonium inventory, but in this case, as well as in the original one, the fuel reprocessing should be necessary for the best exploitation of all the fissile material. Being this complex strategy out of the scope of our proposal, the thoria fuel will not be further considered in our study. 5. The power generation in the plutonium rods steeply decreases during the fuel life (Fig- Thi. 6) se aspecur deemes bettetwa e b do t r investigate assembly-levey db l calculations also considering the use of burnable absorbers.

6. ASSEMBLY LEVEL CALCULATIONS

Single pin cell neutronic calculations showed that the proposal seemed viable and promis- usefua s a r l plutoniufa in s ga m burnin therman gi l reactor concerneds si . Howevere th , extendee analysib o t s t assemblda swa y leve orden li tako rt e into accoun heterogee th t - neity effects of the lattice, which were not included in these calculations. For this scope the CASMO-3 program was applied to the actual assembly geometry shown in Figure 1.

120 Neutronic calculations refer only to an infinite lattice without any soluble poison in the coolant. Depletion calculations were obtaine keepiny db g constan e assemblth t y power alon irradiatioe gth n time. First of all a thoroughly comparison between WIMS-E and CASMO-3 results was carried out. It was found that the agreement between the two different programs is fully satisfac- tory instancer Fo . Figurn i , k-infinite th e7 y versus EFP botr Dfo h programme shows si n for the AP600 and the special assembly which contains the plutonium rods having a plu- tonium contenthin I . s latteFU t 0 equar1. caseo t l , WIMS-E results were obtainey db weighin reactivite gth y values resulting from singl celn epi l calculations proportionallyo t relative th e fractio speciaf no standard lan d rods spitn .I thif eo s approximatio thad nan t on the thermal flux (see paragraph 5) the results are coherent and show the same overall trend simila.A r agreemen alss o twa reactivit e founth r dfo y coefficients whicr fo , dife hth - ferent geometry conditions and assumptions may play a more significant effect. CASMO-3 result showe followine sar th n i g figures .displa 9 Figur d infinite an yth e8 e multiplication factor versus EFP differenr Dfo t specia compositionsd ro l thesn .I e cases the presence of 208 standard rods over a total of 264 and the correct description of the assembly flux distribution render the k-infinity curves of the various assemblies closer to reference th (see eon e also Figur sho1 1 evolutio e Figure . wd th e2) generan e th 0 s1 f n o - ated power (normalized to the average one) during the irradiation for the two rods having peae th (plutoniukL factoBO t (uraniua rL mEO rod d m)an rod) peae .Th k factor ranges from 1.7 to 1.3 at BOL to 1.25 to 1.14 at EOL depending on fissile plutonium concentra- tion. While the initial peak in plutonium rods might be accepted if its thermal conductiv- ity were relatively high, this is not the case for the standard rods at EOL. This aspect appears important, becaus t coulei d affec possibilite th t operato yt t higea h assembly power. Fuel cycle managemen greaf to shoule tb helt dthino o p t s concern, because eth present out-in fuel cycle strategy implie oldeo s tw tha r e batchetth locatee sb cene th -n di tral part of the core, which is a high-flux zone. On the other hand the plutonium fuel should determin morea e flatten flux bot radian hi axiad lan l directions decreaso t o s , e eth maximum assembly power. This will be investigated in the near future. The burnable poison insertio l plutoniual n i m rodenvisages swa reduco dt initiae eth l peak factor. This solution was studied only to dampen this effect and not for the long-term reactivity control as usually done. Obviously this latter possibility is still applicable in this particular context burnable Th . e poisons were gadolinia (Ga.Q) r erbio a (E^Os). stude Th y requirements wereinitiae th ) i : l infinite reactivit three-batchee th f yo s core should not be lower than 1.066; ü) burnable poisons should be completely burnt within firse th t cycle orden ,i avoio rt penalt y overalde an th yn i l discharge burnup. Some results showe termar n i Figure5 n i 1 s o ot f 2 k-infinits1 normalized yan d power generation ver- EFPDs su erbie Th .a clearly appear viablt no sa e solution. Erbiu mburns i t very slowly along the whole fuel life, because of its relatively low absorption cross section (Figure 12). spitn I substantiaa f eo l reactivit firse th tf o cycl yd epenalten (aboue EFPD)th 0 t ya 40 t , the initial power peak undergoes only a limited reduction (Figure 13). On the other hand gadolinia appears convenient as it is evident from Figures 14 and 15, which show that the reactivity maximum pass from 1.3 initiae 1.2th 7o t d 4l an peak from 1.5e 1.32o th t r 5fo unpoisone poisonee dth casd wite ean don h 1.2 %gadolinif o a respectively burnable Th . e poison percentage referree sinerar e th to d t matri x weight (density around 4200 kg/m3). uraniue th r importanAn a sfo mL peaEO t tka countermeasur e reductioth e s ei th f no nominal fuel power density. Therefore r choicou , assumo et s referenca e e reactoe th r AP600, whic characterizes hi lowea y db r power density than conventional PWRs, seems wele b lo t justified. Reactivity coefficients concerning moderator voids and temperature (and then density) are shown in Figures 16,17 and 18. The void reactivity coefficient is not constant with the void fraction and then the coefficients for two different void percentages are given. These plots show that the plutonium rods, in general, worsen the reactivity coefficients render- ing their value less negative. However limitee ,th d percentag plutoniuf eo m rods renders

121 the various curves not too different from the standard uranium ones. It can be concluded that the insertion of these special rods should not change appreciably the reactor dynamic behaviour. The isotopes evolution during the irradiation is displayed for plutonium content 1.0 FU in Figures 19 and 20 both for military-grade and civil plutonium. Peculiar characteristics of compositiond ro l al r . fo showe L IV splutoniuar d TablEO n i an d I ean II mL rodBO t sa

TABLE III: FUEL INVENTORBURNUL EO D PMILITARYAN R YFO - GRAD (WTU E P INITIA F %O ) LHM

0.71 FU U F 0 1. 1.3 FU EPPD 1128 1209 1294 Pu-239 0.833 wt% 1.037 wt% 1.22% wt 2 Pu-240 5.388 wt% 6.91% wt 9 8.147wt% Pu-241 4.65% 8wt 4.935 wt% 5.121wt% Pu-242 5.30% 5wt 4.921 wt% 4.610 wt% Burnt Pu-239 99.10% 98.88% 98.68% Burnt total Pu 83.81% 82.19% 80.90%

TABLE IV: FUEL BURNUP AND EOL INVENTORY FOR CIVIL PU (WT INITIAF %O ) LHM

0.71 FU U F 0 1. 1.3 FU EPPD 1105 1175 1248 Pu-239 0.812 wt% 1.116wt% 1.465 wt% Pu-240 7.958 wt% 10.174wt% 11.929 wt% Pu-241 5.77% 7wt 6.266 wt% 6.67% 7wt Pu-242 9.801 wt% 9.260 wt% 8.798 wt% Burnt Pu-239 98.60% 98.08% 97.47% Burnt total Pu 75.65% 73.18% 71.13%

Fuel burnu practicalls pi y identica thao t lstandarf o t (1182 d7UO EFPD t plutoniu)a m conten , wit FU hslighta 0 equa1. t o improvement l military-gradr fo t e plutoniuo t e mdu limitee th d presenc evee th nf e o plutoniu m isotopes t varieI . s almost linearly wite hth plutonium content, wit hcoefficiena t approximatel EFPD/0.8 2 r d yfo an equaU 4 12 F o t l civil and military grade plutonium respectively. The burnt fraction of plutonium-239 is ver ycasey higan ,civi r hn arouni fo military-grad r l plutoniu% fo d98 % 99 d man e pluto- nium totae Th . l plutonium burnt fractio civinr rangefo l % plutoniu76 s o frot 1 d m7 m an from 80 to 84% for military-grade plutonium. Therefore the final composition of unloaded

122 plutonium, which contains from 66 to 72% of plutonium-240 and plutonium-242, for mili- tary-grade and civil plutonium respectively, is such to guarantee in any condition the lack of proliferation potential. This fact allow chooso t s plutoniuse u eth m conten basie th n sto of an overall optimization study. These values might be modified at core-level, and this will be verified in the near future.

5. CONCLUDING REMARKS

The proposal of burning civil and military-grade plutonium in special non-fertile rods uni- formly dispersed in PWRs assemblies in a fraction of about 20% were studied by single pin assembly-leved celan l l calculations maie Th .n characteristic neutronic aspectt a s assembly-level are as follows. 1. The new fuel seems to have the capability of an almost complete elimination of pluto- nium-239. On average from 97 to 99% of it is burnt at EOL. The total burnt plutonium fraction ranges from 71 to 84% of initial loaded. The plutonium-poor spent fuel con- tains a major fraction of plutonium-240 and plutonium-242 (from 66 to 72%), and then definitels i t i usablt yno weapo r efo n purposes. 2. The slightly higher initial reactivity can be coped with by the usual long-term control solutions and/or by the insertion of burnable poisons in plutonium rods, this latter solutions having als scope oth reduco et e initial power peaks. 3. Reactivity coefficients, which are much less negative for single plutonium rods than standare th values2 onle dUO ar , y slightly lower when evaluate t assembly-levelda , becaus prevalene th f eo t fractio uraniuf no m assemblye rodth n si . The dynamie nth c behaviou reactoe th f ro r shoulappreciable b t dno y affected. 4. The power distribution between plutonium and standard rods in the same assembly changes appreciably durin fuee gth l life. While initial peak plutoniun si m rods might be probably accepte theio t e r drelativeldu y high thermal conductivity, which improves fuel performance case standarr th efo t , thino s si d rod t EOLsa . This aspect appears important also for its effects on DNB limits, but it is dampened by the present trend to reduce power density in PWR cores. Then the initial choice to assume as reference reacto AP600e rth , whic characterizes hi lowe% 25 r a powe y db r density than conven- tional PWRs, is well justified. More pronounced flattening effects of plutonium fuel might decrease the maximum assembly power value. Insertion of limited amounts of gadolinia might face the initial power peak in the plutonium-type fuel. The present choice to have a uniform dispersion of plutonium rods in the standard UO2 assembly appears reasonable under many aspects solutioa t bu , n which envisages alsoa uniform dispersion of standard uranium assemblies and plutonium-type assemblies is presently under study. These neutronic results confirm the viability of the proposed solution, and the need to examine closely the power peak problem also by core-level calculations. Then the open question is the technological assessment of the non-fertile fuel matrix. Available refer- ences reveal that this assessment is in progress at JAERI. However, other research pro- grammes in the world are devoted to such materials both for nuclear and non-nuclear applications. As for the first ones let us remind many actinides burning proposals which envisag adoptioe eth thesf no e materialse latteth r rfo ones A . s many activitien i e sar progress also in our country.

123 rods

UO2 rods

Water-filled rods

1/8 symmetry

Figure 1. Assembly layout

ooo Pu02(0.7)-Al203

civil Pu A A Pu02(1.0)-AL203

milit. Pu Pu02(1.0)-Al203(50%)-Th02(50%) o « o Pu02(1.3)-Al203

500 750 1000 1250

EFPD

Figure Single2. cellpin infinite multiplication factor versus EFPD

124 ————2 U0

O Pu0 O 2-Al203 A A A Pu02-Al203(50%)-Th02t50%)

250 500 750 1000 1250

EFPD

Figure Single3. cellpin temperature coefficient versus EFPD

——— U02 civil Pu Ap/p0=50% O O O Pu02-AL203 milit. Pu A Pu0 A 2 -AlA 203(50%)-Th0 Q 5 _ 2(50%)

.-•is'

-200-

-250-

-300- 250 500 750 1000 1250

EFPD

Figure 4. Single pin cell 50% void coefficient versus EFPD

125 50f

UO, civil Pu Ap/p0=10%

000 Pu02-Al2Q3 mi 111. Pu A A A Pu0 -Al 0 (50%)-Th0 {50%) 0-- 2 2 3 2

-150-

-2004 250 500 1000 1250

EFPD

Figure Single5. void cellpin 10% coefficient versus EFPD

000 Pu02(0.7)-Al203

civil Pu A Pu0 A 2(1.0)-Al203

militu P . Pu02(1.0)-Al203(50%)-Th02(50%)

> Pu0< > < 2(1.3)-AU0 O 3

Vm /V= 1.9f 2 h t impose$ d

250 750 1000 1250

EFPD

Figure 6. Single pin cell fission rate per rod unit length versus EFPD

126 GOO WIMS-E A A A CASMO-3

250 500 750 1000 1250

EFPD

Figure WIMS-E7. CASMO-3and infinite multiplication factor versus EFPD

—— — AP600

COO U02+Pu02(0.7)-Al203

U02+Pu02(1.0)-Al203

U02+Pu02(1.3)-Al203

IOOO 1250

EFPD

Figure Assembly-level8. infinite multiplication factor versus EFPD military-gradefor plu- tonium

127 ———— AP600

000 U02+Pu02(0.7)-Al203

U02+Pu02(1.0)-Al203

U02+Pu02(1.3)-Al203

1250

EFPD

Figure Assembly-level9. infinite multiplication factor versus EFPD civilfor plutonium

———— AP600 (min d max.an . )

U0D C 2 +Pu0O O 2(0.7)-Al203 U02 rod

1.6- À U0 A A2+Pu0 2(1.0)-Al203 Pu02-Al203 rod ..

O o G U02+Pu02(1.3)-Al203

0) o Q_ 1.2- • TJ

_ c0 a> o>

1000 1250

EFPD

Figure 10. Assembly-level normalized generated power versus EFPD for military-grade plutonium

128 ———— AP600 (mm and max)

ooo U02+Pu02(0 7)-Al203 U0d ro 2

16- A A A U02+Pu02(1 0)-A1203 Pu02-Ald 20ro 3

o U0 o O2+Pu0 2(1 3)-Al203

_ (D

O O. 12- •D

_ c D) 8-

civiu P l

4-

250 500 750 1000 1250

EFPD

Figure 11. Assembly-level normalized generated power versus EFPD for civil plutonium

14-

civil Pu Pu02U 3)-Al203-Er203(15%)

Pu02(1 3)-Al203-Er203(12ï) 13- Pu02(1 3)-Al203-Er203(6 0%) j*^ Pu02(1 3)-AL203-Er203(3 0%)

Pu02(1 3)-Al203

12-

11-

first cycle second cycle

250 500 750 1000 1250

EFPD

Figure 12 Effect of erbia poisoning on the infinite multiplication factor

129 US-

Pu02(1.3)-A I203-Er203(15%)

Pu02(1.3)-Al203-Er20g(12%)

1.5-• Pu02(1.3)-Al203-Er203(6.0%)

Pu02(1.3)-Al203-Er203(3.0%)

Pu02(1.3)-Al203

L 1.25- • 01

O o. -o Qj "H 1_ c(U ::: 9

.75-- Pu02-Ald 20ro 3 civil Pu

U02 rod

250 500 750 1000 1250

EFPD

Figure Effect13. of erbia poisoning normalizedthe on generated power

0 Pu000 2(1.0)-Al203-Sd203(0.6%)

A A Pu02(1.0)-Al203-6d203(1.2%)

Pu02C1.0)-Al203

c 4c-

1250

Figure 14. Effect ofgadolinia poisoning on the infinite multiplication factor

130 1.6- civil Pu

O O CD Pu02(1.0)-Al203-6d203(0.6%)

A A A Pu02(1.0)-Al203-6d203(1.2%)

Pu02C1.0)-Al203

1.2- o Q. _0 "io_ c m O)

U02 rod

Pu02-Ald 20ro 3

.4- 250 500 1000 1250

EFPD

Figure Effect15. of gadolinia poisoning normalizedthe on generated power

-200- AP600

U02+Pu02(civil)+Al203 U0 +Pu0 (milit.)-t-Al 0 E 2 2 2 3 ü a. -o -220-

< -240-

-260- Ap/p0=50% Pu content 1.0FU

-280- 250 500 750 1000 1250

EFPD

Figure 16. Assembly-level 50% void coefficient versus EFPD

131 -10CH

AP600

-120- U02+Pu02(civil)+Al203 U02+Pu02(milit.)+Al203

E o o. -140- Q. Q.

< -160 Pu content 1.0FU

Ap/p0=10%

-180-

-2004 250 500 750 1000 1250

EFPD

Figure 17. Assembly-level 10% void coefficient versus EFPD

-504 250 750 1000 1250

EFPD

Figure 18. Assembly-level moderator coefficient versus EFPD

132 o Pu-23 o o 9 A A A Pu-240

civil Pu milit. Pu

1000 1250

EFPD

Figure Pu-23919. Pu-240and evolution versus EFPD plutoniumfor contentFU 1.0

Pu content 1.0FU

e» 0 Pu-2400 1 "S A A A Pu-242 c o c 0) Ü c o Ü oCL

250 500 750 1000 1250

EFPD

Figure 20. Pu-241 and Pu-242 evolution versus EFPD for plutonium content 1.0 FU

133 ACKNOWLEDGEMENTS

The WIMS-E and CASMO-3 programs were used at the OECD-Halden Reactor Project ENEL-ATe (Norwayth t a d )an N offices (Italy), respectively. Cooperatio thesf no e organi- zation makinn si g these programmes availabl greatls ei y acknowledged.

REFERENCES

] M.R[I . BUCKNER, "Compariso f Optionno r Plutoniufo s m Disposal Reactors", U.S. D.O.E. Report WSRC-TR-92-554, Savannah River Site, Aiken, SC 29808 [2] M.R. BUCKNER, P.B. PARKS, "Strategie Denaturinr sfo Weapons-Grade gth e Pluto- •nium Stockpile", U.S. D.O.E. Report WSRC-RP-92-1004, Savannah River Site, Aiken. SC 29808 [3] D.F. NEWMAN, "Burning Weapons-Grade Plutonium in Reactors", 4th Annual Scien- tifi Technicac& l Conference, June 28-Jul y19932 , Nizhni Novgorod, Russia [4] C.E. WALTER, R.P. OMBERG, "Disposition of Weapon Plutonium by Fission", Inter- national Conferenc Technologd ean y Exhibitio Futurn no e Nuclear Systems: Emerging Fuel Cycles and Waste Disposal Options (Global X93), September 12-17 1993, Seattle, Washington [5] E. CERRAI, C. LOMBARDI, "Burning Weapon-Grade Pu in Ad Hoc Designed Reac- tors?", Proceedings of International Symposium on Conversion of Nuclear Warheads for Peaceful Purposes, Rome, June 15-17 1992, vol. II [6] C. LOMBARDI, A. MAZZOLA, "The Uranium Exploitation and the Plutonium Dilemma", Energia Nucleare, no 1,1994 [7] C. LOMBARDI, A. MAZZOLA, "Weapon-Grade Plutonium: Annihilation Via Thermal Fissio Unconventionan ni l Non-Fertile Matrices", International Congres Convern so - sion of Nuclear Weapons and Underdevelopment: Effective Projects from Italy, Rome, 4-5 July 1994 [8LOMBARDI. ]C . MAZZOLA,A Non-FertilA " , e Fuel Concep Plutoniur tfo m Burningn i Thermal Reactors", presented at the EHPG Meeting, 30 October-4 November 1994, Bolkesjo, Norway [9] H. AKIE, T. MUROMURA, H. TAKANO, S. MATSUURA, "A New Fuel Material for Once-Through Weapons Plutonium Burning", Nuclear Technology, Vol. 107, August 1994 [10] J.R. ASKEEW, M.J. ROTH, "WIMS-E, A Scheme for Neutronic Calculations", AAEW- R 1315, June 1982 [II] M.J. HALSALL, C.J. TAUBMANN, "The 1986 WIMS Nuclear Data Library", AAEW- R 2133, September 1986 [12] M. EDENIUS, B. FORSSEN, "CASMO-3, A Fuel Assembly Burnup Program", Studs- vik/NFA-89/3, Rev , Studsvi2 . (1992B kA ) [13] J.D. LEWINS SimplA " , e Teaching Mode Transienr fo l t Fuel-Cycle Analysis", Ann. Nucl. Energy, Vol. 10, No. 9, 1983

134 REACTOR PHYSICS CHARACTERISTIC POSSIBLF SO E FUEL MATERIALS FOR PLUTONIUM-INCINERATING LWRs

. KASEMEYERU , J.M. PARATTE . CHAWLAR , * Paul Scherrer Institute, Villigen, Switzerland

Abstract

A range of mixtures of plutonium and different burnable poisons (BPs) have been consid- ered from the viewpoint of providing a possible LWR fuel without uranium. The reactor physics feasibility of such materials for enabling a considerably more effective incineration

of plutonium than possible with PuO2/U02 (MOX) fuel has been investigated in terms of burnup reactivity swings, temperature and void coefficients and control absorber worths. The potential for reduction of total-Pu has been shown to be ~ 3 times greater than with MOX, the difference in possible reduction factors being much more striking in terms of quantite th f fissile-Puyo bees ha n t I foun. d tha e specifith t c e effect givear a f P o snB crucia determininn i l burnue gth controllabilitd pan y characteristic e hypotheticath f o s l LWR fuel considered suchs A improve.n a , d overall performanc resuln eca t from usinga suitable mixtur f BPseo , rather tha individuae non l type. Further influence th ,e th f o e "inert matrix" employed for the Pu-BP fuel can also be significant - on the void coefficient, in particular e nee thus Th .dha s been indicate r optimizindfo e Pu-Bgth P fuel composi- inere th td matrixtioan terms n ni BP f bot s.o e Finallyhth scopine th , g charactee th f ro present investigations needstressede b o st varioue Th . s results hav l beeeal n obtainen do the basis of unit cell and simplified fuel assembly calculations, and there has been little explicit consideration of materials aspects as such.

1 Introduction

availabilite Th plutoniuf yo mquantitien i beyonr sfa d projected need relativela s si y novel situation. With LWRs likel continuo yt dominato et nucleae eth r energy s sceneha t i , become importan consideo t t r their potential rolreducinr efo therebd an g- y helpino gt regulat large th ee- plutonium inventories world-wide. Thus, whil "self-generatede eth " mod f plutoniueo m recycl LWRsn ei ,f PUÛ2/U0o involvin e us e gth 2 (mixed oxider o , MOX) for constituting 30 to 40% of the core, is corning into broader practice in Europe [1], it does not represent a possible solution in the above context. This is because about the same amount of new plutonium is generated as that consumed. Designs for 100%- MOX core beine sar g investigated effectivn a d ,an e reductio plutoniuf no m would certainly be achievable with these [2]. The buildup of new plutonium from the uranium present in e MOXth , however, remains disadvantageous fro viewpoine mth f inventoro t y reduction and also enhances the need for multiple recycling.

The present paper extends investigations reported recently on the physics feasibility of LWR plutonium fuels without uranium, [3, 4]. Mixtures of plutonium and different in- dividual burnable poison scombinatiopossibla e (BPsf welth s o a - s e a s l- )e us BP f no have been considered. Without specific regard having been given to materials aspects, the physics performanc thesf eo e hypothetical fuel s assessei s termn di f burnuso p reactivity

"co-affiliation: Swiss Federal Institut Technologyf eo , CH-1015 Lausanne

135 swings, temperatur void ean d coefficient controd an s l absorber worths muce Th . h higher potentia reducinr lfo g plutonium inventor compares yi d with 100%-MOa tha r fo t X core, als e isotopiae sar o th c compositio dischargee th f no d plutoniu amounte th d m minoan f so r actinide generated) s Cm (Np , ,Am . Further influence th , e "inerth f eo t matrix bees "ha n investigated - particularly with respect to reactivity coefficients.

2 Calculational Models

The various results being presented were all obtained on the basis of unit cell and/or simplified fuel assembly calculations eacn I . h case BOXEe th , R code which forms parf o t the LWR code system, ELCOS, was used [5].

2.1 CelAssembld an l y Descriptions

Since compatibility with current-day pressurized water reactors (PWRs) is an important goal unie th , t cells considered have bee PWR-typf no e with common characteristics sa given in Table 1.

Table 1: Common Characteristics for the Unit Cells Fuel diameter: 9.13 mm Hot full power (HFP) conditions:

Clad(material): zircaloy Fuel temp.: 600 °C inner diam.: 9.30 mm Clad temp.: C ° 5 31 outer diam.: 10.7m 5m Water temp.: 300 °C

Square pitch: 14.30 mm Pressure: 158.r 8ba

Boro watern ni : none Power density: 103 MW/m3

Six individual BPs, as well as one particular BP-mixture, were selected for the present study hypotheticae th , l fuel materia eacn i l h case consistin particulae th f go mixeP B r d with Pu02 and dispersed in an "inert matrix". The plutonium density assumed was 0.8 g/cm3, i.e. close to the values which occur in LWR MOX fuels, and the inert matrix fo r e casemosth f so t calculated (all, except those discussed additionall Section yi n 3.5) 3 was A1203 of density 3.0 g/cm . The BP density, in each case, was adjusted to yield ~ a &C1.1 f O o 3t beginnina f lifgo e (BOL) witP (Xe-freeHF h ) conditions. Table2 gives the various poison density values used, the single BP-mixture considered being 40/60 atom% boron/holmium. The isotopic composition for the BPs corresponded to naturae th l compositio eacn ni h case (boron, rhodium, indium, europium, holmiud man hafnium). In order to provide comparatory values for the assessment of performance indices, two reference LWR unit cells have been considered, viz.

235 3 (1) with U02 fuel of 4.3 w% U enrichment and 10.4 g/cm density and

136

3 (2) with MOX fuel of density 10.2 g/cm, 3 containing 0.69 g/cm plutonium of fissil% w e71. 2enrichment uraniue th , m enrichment % beinw f 0.2g5o . isotopie Th c composition assume plutoniue th r varioue dfo th mn i s Pu-BP fuels approx- imated that obtained from LWR U02 fuel with a burnup of 42 GWd/t and 5 y cooling time [2]. This composition is compared below with that considered for the reference MOX cell:

238pu 239pu 241pu 242pu Pu-BP: 2.7 54.5 22.8 11.7 8.3 MOX: 1.8 59.0 23.0 12.2 4.0

For the relative evaluation of control rod worths, a single fuel assembly configuration has been considered. This consisted of 20 control rod positions in a 15 x 15 assembly, as indicate Fig.ln di .

Tabl : e2 Densit d Isotopiyan c Compositions Assume e Burnablth r dfo e Poisons 3 A1d an 2 g/cm00 u 33. densitie(P d fue,e an respecth 8 l n 0. zonsi e -ear tively eacn i , h case)

Cell Type B densitP y BP isotopic cimposition (BP) g/cm3 atom% atomic weights B 0.0188 20.0/80.0 10/11 Rh 0.467 100 103 In 0.27 4.3/95.7 113/115 Eu 0.0408 47.8/52.2 151/153 Ho 0.38 100 165 Hf 0.325 0.2/5.2/18.6/27.1/13.7/35.2 147/148/149/150/152/154

B+Ho 0.15 8.0/32.0/60.0 10/11/165

2 Calculationa2. l Method Datd san a

The present analyses were carried out using the BOXER code which forms part of the LWR code system, ELCOS [5]. Both two-dimensionaceld an l l transport/depletion cal- culations can be performed by BOXER, the former in 70 energy groups and the latter after collapsing to a reduced number of broad groups. Eight were used for the current fuel assembly calculations.

An important featur cele th l f calculatioeo treatmene th ns i resonance th f to e energy range between 1.3 and 907 eV, self-shielding of the cross-sections being achieved via a two-region collision probability calculatio ~ 750 n ni 0 energy points. Resonance interaction effects e therebar y taken into account explicitly e resonancth ., AboveV 7 e e90 cross-section s are interpolated in the more usual fashion, i.e. as functions of temperature and dilution employing appropriate tables.

137 Zone 1: homogenized fuel cells Zone 2: homogenized guide tubes, with or without control rods

guide tubes (zircaloy): inner/outer diameter: 12.4/13.8 mm control rods (Ag/In/Cd ~ 81/14/5 w%): density: 9.9 g/cm3 m m 9 8. diameter: clad (SS-304): g/cm9 7. 3 density: outer diameter: 10.m 1m

layerO ZonH^ , : 0.55me3 m half-thickness -100.1mm

Figure 1: Layout (one quarter) assumed for the calculation of control rod worths for the 15 x 15 fuel assembly.

date Th a librar yBOXEe useth r dfo R calculation generates swa d fro Joine mth t European File, JEF-1 [6]. Considerable refinement of the burnup treatment was undertaken recently

contexe th n i f actinido t e transmutation studie presene sth [7]d ,tan library contain4 3 s

248 actinides explici5 ,5 froCmd Tmo an t,ht 232 (plu pseudo2 s ) fission productsr Fo . currenthe t study, datgenerateawas additionadfor l nuclides correspondinBPs the gto considered.

Validation of BOXER calculations, using its JEF-1 library, has been carried out in the past largel basie th H^O-moderated f so n y o criticals ,UO s [8] orden I . enablo rt emora e relevant

assessment for the present calculations, some of the 7.5% Pu/;ss MOX experiments in the PROTEUS-LWHCR programme [9] have also been analysed. Satisfactory agreement was obtained with experimental results, reflecting adequate accuracies for calculating individual reaction plutoniu e rateth n si m isotopes [4].

3 Calculational Results principae Th l results discusse varioue th r dfo s Pu-BP fuels pertai reactivite th o nt y varia- tion with burnup (which is a constraint on the maximum core-residence time and hence on the effective plutonium reduction achievable), the changes in individual nuclide densities, reactivity coefficients and control absorber worths, as well as effects of the inert matrix on the physics characteristics.

3.1 Reactivity Variation with Burnup

Because of the different absorption cross-sections and burnup chains involved, cell cal- culation varioue th r sfo s Pu-BP fuels yielded large difference reactivite th n i s y variation with burnup. Fig.2 shows the results for the seven cases considered and compares these wit correspondine hth cellX g svariation MO (Sectio reference d th an r 2 nsfo 2.1)eUO .

138 It is seen that relatively short core-residence times would be achievable with Rh, In, Ho or othee th n r O hand . indicates BP a ,H s a f Section di n 3.3, individual reactivity coefficients for severa thesf o l e Pu-BP fuel foune quite b ar s o edt favourable.

, ther sligha Eu s ei d Witt an increas hB reactivitn ei y initially befor fale eth l occurso s , that upto ~ 1500 irradiation days (corresponding to ~ 53 GWd/t for a current-day PWR) would be possible - from reactivity loss considerations alone. A relatively flat reactivity variatio reportes nwa d earlier also with erbium l thes[4]al n e.I three cases, however,

600 800 1000 1200 1400 Irradiation [ d ]

1.4 1.4 (b) MOX Pu-BP, 0.0188g/cc B 1.3- h 1.3 Pu-BP, 0.0408g/cc Eu Pu-BP, 015g/cc B+Ho

1.2- -1.2

c -1.1 X

1.0- -1.0

0.9- L0.9

0.8 I ' I -0.8 0 40 200 0 80 100 0 060 1200 1400 Irradiation [ d ]

Figur^ variationk : e2 s with burnu f Pu-Be H th r d pPfo an unio H t, cell In wit) , s(a hRh as the BP and (b) with B, Eu and a mixture of 40% B and 60% Ho. Results cell R e showsar LW r comparison. nfo X MO reference fod th r an j eUO

139 there are certain difficulties with respect to the reactivity coefficients - which provides the incentive to consider mixing BPs for achieving a more favourable overall behaviour. caseo H , % depicte60 - B Fig.2(b) n di % 40 e , Th provide illustrativn sa e example th n ei current study.

3.2 Nuclide Density Changes

Apart from change Pu-isotopese th r sfo productioe th , minoe th f nro actinidem A , Np s- and Cm - has also been considered, with 1500 irradiation days assumed in each case. All nuclide densities have been normalized to the initial amount of plutonium, i.e. that at zero burnup. Sinc e resulteth s obtained were very similae variouth r fo rs Pu-BP fuels, two representative cases (for which the 1500 irradiation days would indeed be feasible) are depicte , viz4 Tablen di d . Pu-Ean Pu-B-Hos3 d uan eacn I . h case, comparatory results are given for the reference MOX cell.

Table 3: Total-Pu and Pu-Isotopics after 1500 Days of Irradiation

Final Pu-Isotopics [w%] Cell Final/ Initial 240 Type Pu (total) 238pu 239pu Pu 241pu 242pu MOX 0.697 2.6 39.7 29.7 17.9 10.2

Pu-Eu 0.315 5.6 7.6 32.8 19.9 34.1 Pu-B-Ho 0.320 5.5 7.3 32.7 19.7 34.7

Table 4: Nuclide Densities for the Minor Actinides after 1500 Days of Irradiation

MA Composition [w%] Cell Total MAs Type Initial Pu 237Np A1 m24 A3 m24 242Cm 244Cm 245Cm Others MOX 0.055 6.0 19.5 40.2 4.3 25.6 3.2 1.2

Pu-Eu 0.063 0.0 6.7 45.9 4.3 37.6 3.9 1.4 Pu-B-Ho 0.067 0.0 7.0 46.5 4.5 36.9 3.7 1.6

As indicated above total-Pe th , u reduction achievee Pu-Bth r Pfo s aboud i fuels % 68 t , ~ 62% being fissioned and 6 to 7% being converted into MAs. This corresponds to a plutonium "incineration" rate of ~ 1.14 kg/GWd, i.e. a figure quite comparable to that being aimeadvancer fo t da d concept Frence th n si h CAPRA program with fast reactors total-Pa fue X s i l MO u e ~ 0.3e resul( th reductio r Th [10]% 8fo t . 30 kg/GWd) f no ,

with the conversion to MAs only slightly lower at 5.5%. If one considers the changes in Pu-isotopics in conjunction with the total-Pu reduction, one finds that thPue239 content e Pu-Both f PX fuel MO reduces i s e mucs a th y s 95%n i hda b s compare% a , 47 ~ o dt case.

140 As regard minoe sth r actinides absence th , uraniuf eo m account therr sfo e Nbeinp237 o gn

produce e Pu-Bth n di P fuelsface Th t . tha muca t h lower amoun 241f o t A mfouns i o dt occur than in the MOX case (even though its production is mainly viPua241 decay) is due

to the thermal neutron flux in the Pu-BP cells becoming several times higher towards the

242 end of the irradiation periodA. m241 is then converted tAom much more effectively, largs it eo anlattert e fissioe dth turn du i , nn- cross-sectio fissioo t n- n productse Th . higher thermal neutron flux also results in more captures Puin, whereby the fraction

of 244Cm produced is larger than in the MOX case. Qualitativel242 y speaking, the longterm radiotoxicity of the discharged fuel due to all the higher actinides - including the Pu- isotope s- woul severae db l times Pu-Be loweth n ri P fuel MOXe s th tha r nfo .

3 Reactivit3. y Coefficients

An initial assessmen f safet o t controllabilitd an y y characteristic made b comy n eb ca s - paring reactivity coefficient controd an s l absorber worthPu-Be th r Psfo cells considered, cases X e relatiorelativTh MO . e referenc d nthoso e th t an user r definin2 efo dfo U0 e g a reactivity coefficient of type x (Doppler or void) with respect to the variation of the corresponding parameter p for the cell is:

with ki denoting either k^ or fce//, and with

Tabl givee5 s values calculate Dopplee th r void fo ran d coefficient t BOLa s formee .Th r s beeha n considere terme n fc«di th ,f chang so increasn a r efo fuen ei l temperature from 600°C to 1000°C. For the void coefficient, moderator voidage ranges of 0 to 10% and 0 100o t % have been considered separately case th partiaf eo n I . l voidag 10%)o t s 0 i e( t ,i adequat consideo et 100o t e coefficien0 th %r r voidFo e changtermn i . ,tth k^ f so n ei 2 however, fce// variations consideree neeb o dt d since change migration si n area (Me ar )

quite different for the cells with and without uranium. This is indicated by the comparison of M-value2 s mad Tabl n ei relativ A . e6 e assessmen e voith df o tcoefficien 100o t 0 % r fo t voidage without consideratio leakagf no e effects would clearlpessimistio to e ye b th r cfo Pu-BP cells, considering thaMe th 2t-change e almos2 sar U0 te twicth r largs efo a s ea casesX geometricaanA .d MO - correspondin 2 l bucklinm~ 1 3. f go g approximatele th o yt dimension 100a f so 0 MW(eassumes corR wa ededucin- r )PW d fo C^(Qe g100%th o t ) result f Tablso . e5

It is seen, as mentioned earlier (Section 3.1), that the Pu-fuels with single BPs, which have reactivity coefficient comparablf o s e magnitud reference d thoso eth t an r 2 efo eU0 MOX cells, correspon e caseth so dt with relatively large reactivity losses with burnup. The B+Ho BP-mixture considered, on the other hand, appears to combine the favourable features of a flatter reactivity variation and acceptable reactivity coefficient values quite effectively.

141 Table 5: Reactivity Coefficients for the Various Unit Cells at Zero Burnup

/°roo Cell ^Doppler C&i( 10%o 0t ) C%d(Q to 100%) Type [10~4/%void] [10-4/%void] U02 -2.4 -15.2 -77.1 MOX -3.1 -18.5 -43.3

B -1.0 - 3.9 - 1.6 Rh -1.3 -6.5 -12.3 In -1.4 -8.6 -0.8 Eu -1.2 -7.1 -1.6 Ho -2.5 -20.0 -21.6 Hf -2.1 -15.6 -3.1

B+Ho -1.7 -12.2 -13.4 Table 6: M-value2 s at (a) 0% and (b) 100% Moderator Voidage for Several Unit Cells

2 2 2 Cell KOO Grit. Buckling [m~ ] [cmM ] Type (a) (b) (a) (b) ) (b ) (a

UO2 1.375 0.644 65.3 -6.76 57.5 525 MOX 1.205 0.848 37.9 -2.707 55 54.0

B 1.134 1.486 21.8 4.3061.5 1130 Ho 1.133 1.127 22.5 1.416 90 58.9

B+Ho 1.176 1.339 29.0 3.2460.8 1048

Fig. valuevoie L 3th d f indicateBO s o coefficient e th w sho s change with burnup Pu-Be ,th P fuels being represented by the B-f-Ho case. Particularly for the 0 to 100% void coefficient, the change to more strongly negative values is much more pronounced for the Pu-BP cell than for the reference uranium-containing fuels. It should be mentioned that the variation of Doppler coefficients with burnu relativels pwa y l casessmalal r lfo .

3.4 Control Absorber Worths

The considerable reduction of control absorber worths (soluble boron and control rods), as the fraction of MOX assemblies is increased, is a major constraint in designing a PWR core with fue100X l %[11]MO . This results fro largee mth r absorption cross-sectionf so Pu-isotopese th , relativ uraniumo et similaA . r situation woul expectee db systemr dfo s with Pu-B consideree b P o fuelst t this bu ,conjunctiosha n di n wit e correspondinhth g control requirements.

142 reactivite Th y burnupo swint e gdu indicates a , d earlier, coul significantle db y lower fo r a suitably chosen Pu-BP fue e referencl th tha r casea nfo eUO . Tabl compare7 e o tw s other types of control requirements, viz. that for equilibrium xenon and the cold-to-HFP reactivity swing. Here, too, the control needs for the Pu-BP cells are generally lower than reference foth r e fuels. Also show correspondine th Tabl n ni e ar e7 g resultsolubl) (i r sfo e boron worth and (ii) control-rod cluster worth (cf. Section 2.1). Both types of control absorber effects for the Pu-BP cases are seen to be very similar to those for MOX.

(a) U02 MOX 2 Pu-BP, B+Ho o -10- --10 # •x t « t o

-15- --15

0 -20 -•20

-25 -25 0 80 100 0 60 0 0 12040 0 0 14020 0 0 Irradiation [ d ]

(b) U02 m ~ -20 5 --20 o PU-BP, B+HO

i -30- --30

o - -40 --40 "••*•-.. ? -50 --50 0 o

ü .70- -•70

-80 l ' t r l ' l ^ I ' l ' l ' T •80 0 80 100 0 060 120 0 40 0 140 0 020 0 Irradiation [ d l Figure 3: Variation with burnup of void coefficient values for voidage ranges of (a) 0 to 10% and (b) 0 to 100%

143 Table 7: Comparison of Typical Control Requirements and Control Absorber Worths

Control Requirements Control Absorber Worths Cell Equilib. Xe Cold-to-HFP Sol. Boron Control Rods

Type (Afcoo^o) (Afcoo,%) (A&oo/lOOO ppm,%) (A*e//,%) U02 -3.1 -5.7 -7.2 -32.6 MOX -1.6 -7.8 -2.9 -20.8

B -0.9 -0.7 -2.2 -18.0 Rh -1.1 -2.9 -2.3 -18.4 In -1.2 -4.1 -2.4 -18.6 Eu -1.0 -1.8 -2.4 -18.2 Ho -1.4 -7.9 -2.6 -20.0 Hf -1.3 -6.6 -2.6 -19.6

B+Ho -1.2 -4.6 -2.5 -19.3

The above considerations have been made for BOL and clearly need to be extended to accoun representativr fo t e average core conditions. Tha controe th t l absorber worthr sfo the Pu-BP fuels increase quite strongly with burnup is indicated in Fig.4. This is again linked to the increase in thermal neutron flux with irradiation, particularly towards end- of-life with fuel cross-sections reduced considerably. e casClearly th f reactivit eo n i s a , y coefficient assessment, realistic whole-reactor calculations are needed to evaluate features related to controllability - both in terms of reactivity balance and power distributions.

3.5 Effects of the Inert Matrix The various results presented for Pu-BP fuels have all, till now, been with A^Os consid- ered as the inert matrix (Section 2.1). The latter's influence on the physics characteristics

s beeha n currently investigate calculatiny db g cases with certain possible alternatives, viz. BeO ZrCd Be mos,CU an Th . t significant effects observe de reactivit werth n eo y

coefficients,4 particularly the 0 to 100% void coefficient, as indicated in Table 8.

Tabl : e8 Compariso f Reactivitno y Coefficient Value r Pu-Bfo s P fuels with 40% B - 60% Ho as BP and Alternative Materials as Inert Matrix

Inert Density /~1 ^Doppler5 C£a(0 to 10%) C££( 100%o 0t ) Matrix [g/cm3] [io- /°c] [10~4/% void] [10-4/% void] A1203 3.0 -1.7 -12.2 -13.4 BeO 2.5 -2.0 -12.6 -12.1 U B4C 2.0 -1.7 -13.0 -13.3 ZrC 5.8 -1.6 -13.0 -15.1

There are two individual effects which contribute to the influence of the inert matrix

on C**/ d o 100%)(Qt e moderatin e th firs s Th i t. ge full effecth yn i tvoide d condition, whic shifn e harca hth t d neutron spectru msignificantlo t y lower energies resultinn gi

144 increased resonance capture in the BP and a lower k value. The second - often partly compensatin influenc ge - effec migratioth e s th i t n eo n area which determine leakage sth e characteristics. Fig. Tabld 5an illustrate9 individuao tw e eth l effects.

(a) —— U02 •—- MOX I -" - Pu-BP, B-t-Ho -•3 a a o o S -4 --4 c

--5

0 , ai -6 - --6

2 o W -7 H

200 400 600 800 1000 1200 1400 Irradiation [ d ]

-15 U02 MOX Pu-BP, B+Ho -20- -*-.,______

Q .25- --25 e •o o

-30- --30 o O

-35 --35 200 400 600 800 1000 1200 1400 Irradiation [ d ] Figure 4: Variation with burnup of (a) soluble boron worth and (b) control-rod worth

Again, realistic whole-reactor models clearly need to be considered for assessing the in- fluence of the inert matrix on void reactivity. Which type of inert matrix is ultimately preferable will f courseo , , als e determineob variouy db s material properties including irradiation behaviour.

145 10 10 10 10- 10 10 10 10 10 10 10 10 Energy (eV)

Figure 5: Neutron spectra for Pu-BP fuels with 40% B - 60% Ho as BP and A12O3 and Be iners Oa t matrix

) 100Tabl(b %d : e 9 Moderato an fc« M d % 2,an -value0 ) (a t rsa Voidag Pu-Br efo P fuels Alternativd an P B s a eo MaterialH % 60 - s Iner sa B witt% Matrih40 x

2 2 2 Cell KOO Grit. Buckling [m~ ] M [cm ] ) Typ(b e ) (a (a) (b) ) (b (a)

A1203 1.176 1.339 29.0 3.24 60.8 1048 BeO 1.188 1.182 38.4 4.1149.0 444 U B4C 1.170 1.186 31.2 3.25 54.4 573 ZrC 1.168 1.271 29.5 2.90 56.8 934

4 Conclusions The various results presented in this paper indicate that, from reactor physics consider- ations alone, it should be feasible to employ LWR plutonium fuels without uranium for achieving considerably more effective Pu-incineration than possible with MOX.

A homogeneous mixtur suitabl a PuC>f eo d an 2 y choser combinatioo , nBP BPsf no , could yield a relatively flat reactivity-vs.-burnup curve as LWR fuel and result in a total-Pu reduction of as much as ~ 1.1 kg/GWd as compared to < 0.4 kg/GWd for a 100% MOX core termn I . f fissile-Puso e differencth , possibln i e e reduction factor s muci s h more striking. As regards the production of the minor actinides - Np, Am, Cm - this would be significantly lower, relativ total-Pe th o et u reduction achieved.

Qualitative consideration of controllability and safety-related characteristics for an LWR core fueled with possible Pu-BP mixture bees sha n mad compariny eb g reactivity coeffi-

cients and control absorber worths with corresponding values for reference UO2 and MOX bees fuelsha n t I foun. d tha , wittHo h individua d whican h hs sucR acceptabls BP lha e

146 value reactivite somr th sfo f eo y coefficients coul obtainede db associatee ar , d with rela- tively large reactivity losses with burnup - whilothe e u th E ern r hando O .suc s B s hBP ,a enabling long core-residence times - would not yield sufficiently negative Doppler and/or void coefficient values. Acceptable solution providee b y usinsy ma d b gcombinatioa f no BPs, a 40% B - 60% Ho mixture having been suggested currently.

The influence of the inert matrix employed for a given Pu-BP fuel is, in general, small, but there could be significant effects on reactivity coefficients - on void coefficient char- acteristics for total voidage, in particular. Thus, not only material properties, but also such physics effects could determin ultimate eth e choic inerf eo t matrix.

Control absorber worth void an sd coefficient value Pu-Br sfo P fuels have been founo dt vary considerably with burnup. Clearly, realistic whole-reactor calculations are needed to evaluate features related to controllability and safety in a more quantitative manner.

REFERENCES

[1] Schlosser, G.J., et al., Nucl. Technol. 102 (1993) 54 [2] Bouchard , ProcJ. , Topica.S IntEN . l Mtg Nexn o . t Generation LWRs (TOPNUX'93), La Hague (1993) [3] Akie, H-, et al., Nucl.Technol. 107 (1994) 182

[4] Paratte, J.M. appeao t , , ChawlaAnnal n i r t e . , NuclR. , . Energy.

[5] Paratte, J.M., et al., Proc. Jahrestagung Kerntechnik, Travemuende (1988)

[6] Rowlands, J., Tubbs, N., Proc. Int. Conf. on Nuclear Data for Basic and Applied Science, Sant (1985)e aF .

[7] Kadi, Y., et al., Proc. Int. Conf. on Future Nuclear Systems (GLOBAL'93), Seattle (1993) [8] Pelloni, S., et al., Nucl. Technol. 94 (1991) 15

[9] Chawla, R., et al., Kerntechnik 57 (1992) 14

[10] Stanculescu, A., et al., Proc. Int. ENS Nuclear Congress (ENC'94), Lyon (1994) [11] Bergeron t al.e , , J. , Proc. Int. Conf Physicn o . Reactorf so s (PHYSOR'90), MarseiUe (1990)

Next page(s) Lef7 t 14 blank PHYSICAL AND TECHNOLOGICAL ASPECTS OF CERMET FUEL. APPLICATION OF Pu BURNING IN WER REACTORS

V.M. DEKUSAR, A.G. KALASHNIKOV, E.N. KAPRANOVA, A.D. KARPIN, I.S. KURINA, V.V. POPOV Institute of Physics and Power Engineering, Obninsk, Russian Federation

Abstract

The VVER-500 reactor was taken as an example to consider changes in light-water reactor neutronics and fuel cycle features when transferring to plutonium (weapon-grade composition) cermet fuel in all core assemblies. Technology problems of fuel fabrication are discussed.

INTRODUCTION

Nowaday e mixeth s d oxide fuel (MOX-fuel s user i )plutoniu fo d m utilizatio R reactorsPW n i n . Ther s onli e y abou % plutoniu5 t m oxidn i e light-water reactor fuel composition, the rest is depleted uranium oxide. Becaus thif eo s reaso intensive nth e proces uranium-23f so 8 conversion into plutonium-239 takes place simultaneously e witplutoniuth h m isotope transmutation withi e reactoth n r run .t reduceI s plutonium burn-out efficienc d stimulatean y e interesth s consideration i t f opportunitno o t y e fueus l with lower amoun uranium-238f to . One such fuel may be the cermet fuel as plutonium oxide and uranium oxide mixture dispersion in zirconium matrix or plutonium- oxide and magnesium oxide mixture dispersio n zirconiumi n e lasth t n I cas.e th e fertile material is absolutely absent, which gives an opportunity to achieve the most high plutonium burn-up. The cermet fuel composition has the number of advantages as compared with oxide one in the problems of reactor unit safety under the normal oprational conditionals and in the event of of design-basic accidents with the main piping rapture. It shoul e noteb d d also that cermet fuee allowus l o carrt st ou y neutronics parameters optimization without any changes of fuel rods or fuel assembly design.

1. TECHNOLOG CERMEF YO T FUEL MANUFACTURE

The utilizatio f cermeo n t fuel (tha s sinterei t O U particled a n i s Zr-based allow matrix R reactorWE n )i s (water moderated water cooled power reactor) enables the effective settling of several problems, making it possible: o ensurT 1. e indestructibilitth e f cermeo y t fuel elements under designed coolant-loss accident circumstances owing to fuel element cladding temperature reduction to 500 C during in the accident (including the maximum designed accident e absenc s pressurd owinth ga ) an o f t go e e under the cladding; 2. To ensure the possibility of reactor exploitation in the dynamic regime without fuel elements claddin e o damagecorrosionab t o e t gdu d l cracking under stress owing to absence of contact of agressive fission products with the cladding; o T 3kee. p withi e fueth nl composition practicall l fissioal y n product casn si f sea eo claddinr o l g failure.

149 Conceptio cermee th f to n fuels utilizatio R reactorWE n i ns beesha n discussed in detail in the work [1]. The above specified advantages of cermet fuels as well as higher effectivit f plutoniuyo m burn-ou cermea n i t t fuel element mad possiblt i e e to begin in our Institute of Physics and Power Enginering, being now the State Scientific Centre, some technological wor n creatioo k f cermeo n t fuel d cermean s t fuel elements o settlT . e probleth e f cermeo m t fuel utilizatio reactorn i n s intende r plutoniufo d m burning-out e technologth , y

;ïém'eÈts;wtft - - ; u'tïïïzaffpiiofMOXfuel',,,„'.'*'?/:*". *"'''; Nitrate solution Nitrate solution

recycling Mixing

Ammonium precipitation

Mixed ammonium salf U/M£/n_Pu Mother liquor

Calcination

Reduction additionao T f lo extraction U and Pu/Pu/ Powder UO2+PaO2 /MgO+PuO2/ Spheroidization by the slip Depleted solution casting method

Sintering Int wakee oth s storage

Spheroidal fuel UO2+PuO2 /MgO+PuO2/ Depositio Zr-basef no d allow coating

Charging into a Zr-allow shell

Fillingby molter allo -baser w(Z d allow) Fuel element manufacturing

Figure 1.

150 Table I The neutronics and fuel cycle parameters VVER-50e th r fo 0 reactor with various fuel types.

Reactor versio tern i nf fue o m l Parameter UO PuO +Ur O+Z PuO O+U 2 2 2 2 2

1. Power, MWt(e)/MWt(t) 640/1800 640/1800 640/1800 2. Load factor 0.8 0.8 0.8 3. Refuel ling factor 5.43 5.43 5.43 Materia. 4 l flo unclosen wi d cycle natura- l uranium, t/year 98.5 11.2 6.6 - weapon-grade plutonium, kg/year - 349 417 5. Enrichment with fissile isotopes, % 3.8 3.5 6.3 6. Burn-up, MWt*day/kg 45.3 45.3 74.4 7. Average conversion ratio 0.56 0.65 0.52 8. Annual plutonium unloading, kg/% 113. 4/ - 211/60 157/37.6 9. Spent fuel radiotoxicity increase as compared to make-up one *) - only of plutonium s me -i t 3 4. 6 time2. s - of plytonium, minor actinides and decay products - 0 time1 s 7.3 times 10. Potential biological hazard, 3 11 ia m H20 5. mo 1.2*10 1.2*10" 11. Reactivity change, total/margin lifetimee forth % , 16.2/8.0 16.3/7.8 17.4/9.6 12. Radial power peaking factor BOL/EOL 1.38/1.31 1.50/1.42 1.62/1.56 13. Temperature reactivity worth coefficient e BOLth ,n i s 10~ 41/K - of the coolant, cold 0.45 1.01 0.97 MPUM 0.14 -0.03 -0.09 at -power -2.66 1 1 . -3 -2.55 - of the fuel, cold -0.27 -0.29 -0.28 MPUM -0.24 -0.26 -0.23 at -power -0. 18 -0.20 -0.20 14. Relative absorber rod efficiency (BOL) 1.00 0.81 0.91 15. Effective delayed neutron fraction, 10~3, BOL/EOL 6.22/5.86 4. 13/4.31 3.55/3.93 16. Isotopic composition of unloading plutonium% , Pu-238 2.0 1. 1 1.3 Pu-239 48.8 37.9 26.2 Pu-240 27.3 33.4 37.3 Pu-241 13.8 17.0 17.5 Pu-242 8.0 10.6 17.7 After 10 years storage.

f manufacturo e followinth f o eo type tw gf cermeo s t fuel compositions i s being now worked out: a) a composition containing depleted or natural uranium as a fertile material and including in it plutonium dioxide and uranium dioxide (MOX) in allow-baser Z a d matri alloy)r Z x+ (Pu O ;U O- £ c.

151 b) a composition containing no fertile material and including plutonium dioxide and magnesium oxide in a zirconium-based matrix allow)r Z + .O Pu (Mg - O The volume ratio of zirconium alloy in the both cases is about 50 per cent, and the volume ratio of plutonium dioxide that is needed to maintain the power output of a reactor is 2.8 and 3.5 per cent, respectively. Fuel of both kinds in powder form can be obtained by the ammonium precipitation from nitrade solution with subsequent calcination and reduction. From these powders e sli y meanth b ,p f o castins g methodr fo , example, spheroidal particle e obtaineb n ca s d thereafter subjecte sino t d - tering. The coating of the matrix material can be deposited on .spheroidal particles by the ionplasma spraying method, for example. Coated spheroidal particles are charged then into thin-wall zirconium shells and impregnated with molter Zr-base alloy. Fuel rods obtained in this way go then to fuel element fabricating. The flow diagram of the cermet fuel manufacturing process in presented in fig.l. Cermet fuels obtained in this process possess high thermal conductivit highes i A ( yr tha Wat0 r grad1 t pe tr centimetre pe e d goo)an d dissolutio e nitrith n i cn acid, tha s veri t y importan n cas i tf spen o e t fuel element processing.

NEUTRONICE 2TH . FUED SAN L CYCLE FEATURES REACTOFOE RTH R WITH VARIOUS FUEL TYPES

REACTOE 2.1TH . R VERSION WIT MIXEE HTH D FUEL

Tabl I showe e calculatioth s n result f fueo s l cycle parameterd an s neutronic e VVER-50sth oner fo s0 reactor wit X fued cermeMO han l e on t l corloadeal en i dassemblies . This reactor versio s comparini n g wite th h conventiona O fuelled(U l ) reactor e reactoTh . r versions witX fued MO han l 2 cermet e weapon-grad th e fuebasear ln o d e plutonium e isotopiTh . c composition of weapon-grade plutonium was chosen as follows [2] :

238pu/239pu/240pu/241pu/242pu/ _ 9/93. 34/6_04/0 . 58/0. 047 (in weight percent) e zirconiuTh . m conten n cermei t t fues equai l0 5 l volume percent. The plutonium content in MOX fuel and cermet one was chosen to preserve total power output. calculatioe Th n results e mako allofollowint th es u w g conclusions. e greates1Th . t amoun f plutoniuo t m (260k r 62.4s o initiag it %f o l loading) is burnt-out in the reactor with cermet fuel. The same values for MOX fuee 138k ar l40%d gan . f cermeo e tus fuee Th l increases annual plutonium consumptio% 20 y b n and decreases radiotoxicity increase of plutonium and minor actinides in spent fuel afte year0 1 r storagf so e lif 30%y b e . 2. In the reactor with cermet fuel temperature reactivity worth coefficient coolane termn th i s f so t temperatur minimue th t ea m power level under monitoring and above it are close to that in the uranium fuelled reactor. The control and safety rod system efficiency is also close to that e uraniuith n m fuelled reactor. These circumstances simplif e reactoth y r safety maintenance with substitutio uraniuf no m fuel with plutonium one.

2.2. REACTOR WITH THE FUEL WITHOUT ANY FERTILE MATERIALS

To evaluat e maximuth e m feasibilit f o light-watey r reactorr fo s plutonium burn-out consideratio e reactos giveth nwa o nt r with cermet fuel

152 in form of plutonium and magnesium oxides mixture in the zirconium matrix (weapon-grade plutonium) calculatione Th . s showed that when using such fuel compositio s realli n t withoui y U possiblt o achievt y e plutoniuth e m burn-up fraction more than 80%. However here serious difficultiee sar encountered followinge th .s a The e .yar 1. The deficit of the fertile materials causes reactivity change within the lifetime to be increased approximately twice as compared to that in the traditional WER reactor. It requires early in the run a corresponding increase borith cf o eaci d concentratio n wateri n . This causes a positive temperature reactivity worth coefficient of water in process of the reactor heating-up and even when the reactor is at-power. This is intolerable for the reactor safety reasons. It is useful to increase refuelling facto orden i rdecreaso rt' e margin reactivite th r fo y lifetime. 2. It is difficult to provide the necessary power density flattening due to a high fuel burn-up. The required reduction of boron concentration in water and power distribution flattenin e provide b e rod th n s ca gy b witd h burnable poisons (RBP). However then they must be installed not only into made-up fuel assemblies, but also into assemblies being in operation for several years i.e., there shoulpossibilite th e db replacinf o y P duringRB g refuelling. 3. The absolute value of Doppler temperature reactivity worth coefficient decreases very significantly. For the first loading in the beginning of the run its value is equal to -0.2*10 —5 l/C0 . To increase this coefficien f speciao e us tl absorber requireds si . 4. The most considerable reduction of the effective delay-neutron fraction value (this fraction is 2.5*10~3 in the reactor with the initial loading) takes plac thin ei s reactor.

CONCLUSION

1. The use of cermet fuel instead of MOX one allows: - to increase plutonium consumption by 20%, - to decrease summary radiotoxicity of plutonium and minor actinides in discharged fue 30%y lb . 2. Temperature reactivity worth coefficients in terms of the coolant temperature and the control and safety rod system efficiencies are close to those in the uranium-fuelled reactor. This simplifies the reac'tor safety maintenance with substitution of uranium fuel with plutonium one. 3. The use of cermet fuel without any fertile materials results in considerable neutronics parameter changes for the worse. However the reactor with such fuel deserves a careful study as one that allows to provid r maximuefo m plutonium burn-out.

REFERENCES

[1]. V.POPOV, A.RYSHKOV, V.SHARAPOV, V.TROYANOV, V.SPASSKOV, O.NIKISHOV. A cermet fuel application for PWR safety improvment. Proceedings of International Conference on Design and Safety of Advanced Nuclear Power P7-7Plants, II l , ,vo Tokyo, .Japan, (1992).

[2]. GILLET T.C., DENNING R. S. , RODIHALG J.L. Shielding calculation technique e desigth f r plutoniuo nfo s m proceeding facilities. Nuclear Technology. 31(2), (1976) 244.

Next page(s) Left blank 153 FUEL WITH CARBON MATRI BURNINR XFO G PLUTONIUM IN THERMAL REACTORS

E.A. IVANOV, G.N. KAZANTSEV, I.Y. OVCHINNIKOV, P.P. RASKACH SSC IPPE, Obninsk, Russian Federation

Abstract It is offered to use the matrix fuel for effective plutonium burninthermae th n i gl pressurized water reactor. The fuel with carbon matrix is plutonium (or another heavy metal) dioxide into manufacturing porous graphite block. This fuel composition being e thermath applie n i ld pressurized water reacto re neutroth lead o t sn spectrum shift and, therefor, changes the temperature reactivity coefficients and efficiency of burnable absorbers. Such technology allow us to place into the fuel rod only plutonium without uranium and thorium. The technology of reprocessing and recycling of such fuel is very simple. facilitye th t Therno ,e whicear h use thin i d s technological chain that once can be used in the double purposes.

INTRODUCTION. We offer for solution of the plutonium utilization problem to apply the matrix composition as fuel. The fuel with carbon matri plutonius xi m dioxide whic places hi d into manufacturing porous graphite block. Such fue suggese lw o t use into the fuel rods in the thermal pressurized water reactor as WER-type. The useful of carbon matrix is that one can be reprocessed very easily. Technological properties of such matrix allow us to avoid of high chemical reprocessing and isotopic separating f elemento y An .n thi i s s technological chain (reactor, fuel reprocessing, production of the fuel rods etc.) . can't be used in the plutonium production purpose. Thus, this technology can't have double purpose. Fuel rode manufacturear s d with graphite matrix have p betweenoga t n fuel bloc d claddingan k e thermaTh . l conductivity of the matrix more higher then of ceramic one. Hence e operatinth , g conditio f o plutoniun m dioxide particle n suci s h fue s mori l e easy the n conventionai n l fuel. Ther e somar ee variant o product s e ceramic fuel into carbon matrix. We know nf them o e abou firso Th .s tw develope i tte th n i d ASPRIIM (the author G.G.Bayburin) e seconTh . s developei d d e IPPth in E (thOvchinniko. Ya e . I authord an e var s G.N.Kazantsev etc.). Proe neutronith m c physics poin f vieo t w this fues ha l some benefits too. Graphite matrix being installed into thermal reactor fuel gives new neutronic properties of burnable absorbers graphite th d .s soliAn ei d moderatord ,an

155 varyin f it'o g s concentratio n changca reactivite e th ew n y feedback coefficients.

1. TECHNOLOGY. There are two technologies of production of fuel with graphite matrix e therma ,th e use b whicn i dn l ca hreactor s for the plutonium burning. One of them was developed by IPPE's specialists. The initial components of such matrix fuel is manufacturing graphit d ceramian e c particles (uraniur o m plutonium dioxide etc.) . The more convenient size of dioxide particles is nearly 0.6-0.8 mm, but the permissible one may be another. The specialists in the IPPE developed the technolog productiof yo e cristallith f no c plutonium dioxide particles through electrolization of molted heavy oxides or heavy metals. e fueTh l being manufacturin y sucb g h procedurs ha e unizotropic physical properties (thermal and electrical resistance, experimentalls strengthi t i d )an y obtained that ther unizotropis ei c shift under irradiation. Thermal conductivity in this fuel about by six times higheceramie th rn i cthe e fuelnon . e reprocessinTh f suco g h fue le ver b hav yo t esimple . After removing of the cladding, the reprocessing may be carried out by milling of matrix and by following separating of carbon and heavy dioxide particles into gaseous stream or into liquid with specific weight about 3-5 kg/L. In opposite to the high chemical reprocessing the main procedure reduces to simple mechanical separation. Low strength of carbon matrix make available to except p betweethga e n fuel bloc d claddinan k n fuei g l rods. Therefor e thermath e l conductance b e y betweema y he n increase. This fact and relatively high thermal conductivity of carbon matrix ensure more lower inner fuel temperature then one in the conventional WERs. e usin Th f suco g h technology produco allot s u w e fuel pins without uranium and thorium, and to place more easily into fuel absorber anothed san r transuranium elements. There e somar e proble f i reprocessem d thorium dioxidee Th . reprocessing of irradiated thorium dioxide, been used in such fue mors li e easy. There is technology, which is developed in ASPRIIM (G.G.Bayburin). This technolog n powde s baseno i y n o rd proceduree Th . initial components here are porous manufactured graphite bloc d liquian k d substation e maith n d procesAn . s hers i e impregnatio graphitf no e bloc liquidy kb . The reprocessing of such fuel also very simple. This fuel has isotropic properties and also has high thermal conductivity. Thi s higi s h chemical n alstechnologca e b o t i t bu y very preferable n vare heavca th y . e yw Usin metae on gl density into fuel, which should be very useful for thermal reactor with high burn-up fuel management scheme.

156 Both of technologies (from the IPPE and from ASPRIIM) e b use dn ca without significant restrictionr fo s manufacturin e fue th e therma lth f o grodr fo ls reactors. They are similar from the neutronic physics point of view. beind An g appliethermae th n i dl reactor they haven't double purpose (to burning and to accumulating of the plutonium). Both of them allow us to create fuel composition with plutonium without fertile material. Both of them allow us to vary the concentration of heavy component up to 50%.

2. NEUTRONIC PHYSICS. Graphite being install e intpressurizeth o d water reactor's core is solid moderator and it also has influence n spectruo f slowing-dowo m e nneutro th neutrons o t n e Du . spectrum shift the properties of integral fuel burnable absorbe alse ar ro changed. Thus r uraniufo , R m WE fue n i l the burning rate of integral fuel gadolinium become lower d becoman e e fissilalmosth f o te e equamaterialon o lt . Changing ratios between the water and the fuel concentrations and installing solid moderator, we change the value of void coolant reactivity feed-buck coefficient. Everywhere it's possibl o adjust e t such parameterf o s burnable absorber o thas s t fuel temperature reactivity feedback coefficients als e negative reactivitar oth d an e y margin is very small [1]. e everIth n y core loading variant being considered late reactivite rth y coefficient e negativar s d smalan en lo absolute values. Some problem y occuma sf plutoniui r m dioxid s usei e d without fertile materials r (thoriucase)ou n i .m This problem are also solved by choosing of resonance absorber (suc rhodiums ha , rheniu wolfrad man m etc.). The numerical calculation shown that positive contributions in temperature reactivity coefficients are brough y fissiltb e materials (U-233, U-235, Pu-239, Pu-241) and some fission products (Xe-135, Sm-149). The negative contribution - by non fissionable heavy isotopes (U-238, Th- 232, y Pu-24b othe y d b Rh-10 Pu-242d an r0d an 3an ), resonance absorbers (W, Re etc.). Without fissile material breeding the reactivity bias should be too big, and only burnable absorber can compensate its satisfactorily. In the binary moderated systems with high enrichment by fissile isotope the negativity of the void coolant reactivity coefficient s i determines y b d concentratio propertied nan resonancf so e absorbers. It's a reason to use the rhodium as burnable material with or without boron. The variants of core are considered later in this report are not optimized. We want to show that application of the carbon matrix to dilute the fuel into rod may be usefu r solvinfo l g some problem plutoniue th f o s m burning. Thus, e examplereactoth f o rs core characteristics demonstrates only principal possibilit e e usinth th f f o go y fuel based upon carbon technology d it'An . s possible that parameters of reactor cores will be further improved.

157 3. REACTOR CORE CHARACTERISTICS. We considered some variante nucleath f o sr reactors like WER-type with the plutonium fuel into carbon matrix. Desig f reactoo n d parameteran r f coolano s t don'e b t changed. In our investigations we modified only number of fuel assembl d numbean y f fueo r l rods into assemblye Th . outer diameter of fuel rod is non modified, cladding material - Zircalloy. Height of the core and conception of contro e WERl th s the a bodien . i ye ar Maintainins e th g coolant characteristics (pressure, temperature, drop in pressure and difference between inner or outer coolant temperatures) without changin e defingw e speciath e l power d temperaturan e distribution into fuel rodd assembliesan s . o savT e hole powe muse rw t increas outee eth r core radiuf si decreasing water par latticn ti e {i.e. rod-to-rod step)d .An if rate of plutonium burning is fixed the permissible volume part of plutonium into fuel rod must be decreasing with decreasing of rod-to-rod step. Therefore the system can be critica lcertaio t onl p yu n relative ste f latticeo p , which is defined by the plutonium burning rate. The fuel after irradiation isn't been able to use in military purposes uraniume th d .An , which produce n thii d s reactor contains nearly 20% of U-234 and a lot of U-232 (about 900 ppm) . Thus, after even the first cycle this fuel can't be interested for terrorists. We considered such reactors cell which can be used as a partial e loadinconventionath n i g l WER-type reactors. Ther e somar e e limit n containino s f fissilo g e material fraction into fuel rod. Some limits are connected with the strength and reliability of the fuel rod, and they are well- know d haven'nan t some differenc conventionar efo froe on m l reactors. And there are some limits, which typical only for plutonium transmitters and burners. We should to increase (ipossiblet i f specifie )th cplutoniue ratth f eo m burning, and also the part of the plutonium at the end of cycle have to be small. e firstth e consideren w ,I d WER-type core wite th h electric power 1000MW, which fueled only weapon-grade plutonium into the graphite matrix. The fuel free from fissile material have positive temperature reactivity coefficients o overcomT . e erhodiu us thi e w sm into carbon matrixes. Numerical calculations gives that in all variants e reactivitth y margi s aboui n te reactivitth 2-5 d %an y coefficients: upon fuel temperatur s aboui e t -0.01 -0.02 pcm/K on the low level and -0.05 -.1 pcm/K on the full power; and upon coolant temperature (including changing of the density) is about -0.1 -0.15 pcm/K and -0.2 -0.3 pcm/K relatively [1] . Really, we can adjust absorbers so that reactivity feedback wil e morb l e negative t isn'i t t bu , necessary. Parameter plutoniuf so m reactor presentee sar n di table 1. We considered the core operating without partial reloading. Here fuel after one year long irradiation is remove d reprocessedan d . Equilibrium cycle means that fuel after reprocessing is mixed with fresh weapon-grade

158 Table 1. The main parameters of the core with plutonium loading without fertile materials. Equilibrium cycle.

1 1 2 the core radius, m 1.556 1.311 the number of the fuel 198 146 assemblies the assembly dimension "on24.36 22.89 flat", cm Number of rods in the397 331 assembly lifetime, eff.days 350 350 the electric power, MW 1000 1000 absorbers Rh+Gd Rh+Gd material balanc year epe r in out in-out in out in-out Pu-239g k , 1096.84 b91.78 705.06 1121.2 407.23 713.88 Pu-240, kg 203.85 158.55 45.29 208.84 162.3 9 45.86 Pu-241, kg 079.87 275.34 4.529 278.06 273.4'7 4.586 Pu-242g k , 239.44 239.44 0.00 211.94 211.4 9 0.00 plutonium and loaded into the core. In calculation model in the fuel reloading intcore assumee oth ew d thagaseoul tal s left from fuel and time between reprocessing is three years. It's shown that instea equilibriuf do m compositione sar different but values of the plutonium consuming in this reacto e equalar r t meanI . s that e corth choice f o e parameters should be based upon the reactor safety criteria.

Table 2. The main characteristics of the core loaded by Pu andmixture.Th Equilibrium cycle.

I 1 2 ß the core radiusm , 1.556 1.311 1.147 the numbee fueth lf o r 198 146 120 assemblies the fuel assembly24.36 23.89 23.10 dimension "on flat", cm the number of rods 397 331 271 the electric power, MW 1000 1000 1000 the time between 300 300 300 reloading, eff.days typ f reloaeo d 1/3 in-out-in 1/3 in-out-in in-out-i3 1/ n material balance per year in out in-out in out in-out in out in-out rh-232, kg 1936 1845 91.06 1936 1844 92.07 1936 1843 93.65 Ü-233, kg 60.34 60.34 0.00 ß9.06 59.06 0.0 58.43 58.43 0.0 U-234, kg 12.82 12.82 0.00 12.02 12.02 0.0 12.01 12.01 0.0 Pu-239g k , 379.2 66.9 312.3 B80.4 69.9 310.5 381.2 72.6 308.6 Pu-240, kg 43.32 23.26 20.06 44.73 24.79 19.94 46.12 26.29 19.83 Pu-241, kg 61.44 59.44 2.006 ^8.85 56.86 1.994 56.66 54.68 1.983 Pu-242, kg 32.24 32.24 0.00 29.02 29.02 0.0 26.45 26.45 0.00

159 Later we considered the same reactor but with the mixed plutonium-thoriu e maie tablth th n m 2 et A loadin . ] [1 g parameter f suco s h reactor d balancee an fuelsth sf o s materials are demonstrated. The lifetime of such cores was choose very shot -only from methodical considerations A . mentioned above, the optimization of the technical characteristics of such cores hasn't been an our aim. As in the case with free plutonium here isn't obtained the strong dependent of material balance from the type of reactor cells. Thus, the such core may be optimizing upon anothey safetan d ryan criteria too. We considered the problem only of the weapon-grade plutonium burning. And in the main we tried to minimize the plutonium concentratiof cycleo n i d n thiI en d . san e th n ni the other case we supplied that fuel is used without chemical and isotopic separation of heavy metals once from another. Thus, uranium in the variant with thorium have equilibrium concentration.

CONCLUSION

There are two variants of carbon matrix fuel technology. Both of them are useful to produce fuel rods. The fuel based on carbon matrix have some benefits. In thi ss possibl' fuet i l o e loat plutoniueth d m dioxide and combined it with any heavy dioxide (thorium in e uraniub y d som ma an mr case e d ou minoan , r actinides). Operational conditions for ceramic particles will more easy the n conventionai n l fuel. Suc s morhha e fue d lowero l r inner temperature then ceramic one. reprocessine Th sucf go h fue vers li y simple. Matrix composition supply reliable containment of gaseous fission products e temperatur.Th f fueo e l particles relativel w andlo y , hence, gaseous will contai y heavb n y ceramic itself. If gaseous will leave from fuel particles they'll kept by pores in matrix. There somar e e other e benefitsucth h f o sfuel . Graphite matrix being installed into thermal reactor fuel give w neutronine s c propertie f burnablo s e absorbersn I . addition this, graphit s solii e d moderator d varyinan , f o g it ' s concentration we can change the reactivity feedback coefficients. In such fuel there are more of variable parameters, i.e. that parameters of this system more flexible and, hence e optimizatioth ,e corth e f o n characteristics with such fuel is simplified. And, it's very important that the every facilities from the technology chain, whic s preseni h t above can'e b t applied for accumulation of the weapon-grade nuclides. The all stage f proceso s s wil e reliablb l e protected against terrorism.

160 REFERENCE

[1] E.A.IVANOV, A.G.MATKOV "Application of the solid moderato e fueth l n i rassembl e tanth kf o ytyp e reactoo t r increase the reactor stability on the low power", Computationa d Experimentaan l l Validatio f Nucleao n r Power Safety and Fuel Cycle Investigations (Proc. 8-th Top. Meetin reacton go r physics problems, Moscow c "Volga"s/ , - 5 , 9 Sept.,1993), MEPhI, Moscow, volp 129 , 2 ..

Next page(s) Left blank 161 SESSION 4: MOLTEN SALT SOME IDEAS ABOUT HYBRID SYSTEM CONCEPTS

P.A. LANDEYRO ENEA, CRE Casaccia,

A. BUCCAFURNI . SANTÏLLA , I ANPA

Rome, Italy

Abstract

The incineration of minor actinides is a non sense if the problem of plutonium, that is the most important of the actinides, is not resolved. The concept analysed in the present paper is an extension of the ones developed for minor actinide burning mose Th . t promising hybrid system concept considers targefued an l t bot liquidss ha . From the results obtained, it is possible to discard solid and sodium targets. The solutions adopting composite target seem, at present stage, the most promising, but still remain productio probleu e P sth e th levemf a o acceptablt t na lno burninu P a n ei g system. necessite th o t f electricae yo Du l energy production blankee th , t using plutonium dissolve moltedn i n salt permit reaco st highe e hth r thermal efficiency. 1. INTRODUCTION A hybrid system is composed of a particles accelerator, a target, for neutron production, a blanket, for transmutation or fissile material production, and their respective interfaces (i.e. the window between accelerator and target). These systems were studie pas e fissilr th tfo n di e material production lase th t yearn I . s these ideas were reconsidere minoe th r rd fo actinid elong-lived (MAan ) d fission product destruction as alternative to the traditional final disposal of nuclear waste. s plutoniuA mose th ts m i importan e actinideth f o t s mass) a (abou e % th , 95 t incineration of minor actinides is a non sense if not resolved the plutonium problem. Plutonium is an excellent nuclear fuel, the most convenient solution should be the burning in a nuclear reactor to produce electrical energy. But in the case of very big stockpiling, it seems interesting to burn a part in less efficient, but less time consuming systems, such as thermal hybrid systems as demonstrated by the Academy of Science of the United States. The concept analysed in the present paper is an extension of the ones developed for MA burning. The most promising hybrid system concept considers both fuel and target as liquids.

. PLUTONIU2 M presene th n I t work, plutonium meanactinidee th l sal s presen dischargee th n ti R dPW standard spent fuel excep uraniue tth m isotopes. Standar spenR dtPW fue define2 s i l UO s da fuel enriched at 3.3% irradiated at 33000 MWd/t

165 TARGE. 3 T The target can be solid or liquid. In the first case, it is possible to consider two accident scenarios: loss of coolant and loss of beam. In the first case if the beam is not switche quitf dof e soo targee nth t will secone meltth fase n i ,d th t quenchin provokn gca ea strong thermal shock. The molecular dynamic calculations seem to demonstrate that long-range order disappears in the areas travelled through by very energetic particles; this means that in these zone e materia th stargee th f to l becomes glass. This result shoul confirmee db e th y b d behaviour of the tungsten window at Paul Scherrer Institute (PSI) [1]; which, after a thermal shock is broken into very small parts, practically like powder. previoue Th s considerations see mindicato t convenience eth liquif eo d targets. In the case of a heavy metal flowing target, the protons do not penetrate very much majorite targete neutronth e th mose th d f o f Th .an yproduceo e t m sar c firse 0 th 2 t n di interesting solution to flatten the neutron flux shape is to use composite targets: a low mass number liquid, such as lithium or sodium (primary target), surrounded by a secondary target mad heavf eo y metal (suc leas h a uranium)r do . A set of calculation for pure materials was carried out considering uranium, lead, tin,

sodiu berylliumd man . Uranium targets produce some amoun plutoniuf o t s mspallatioa n product; lead produce long-livee sth masPbw d;206 lo s number targets hav neutrow elo n yield verd an y high fractio f neutro o hignn i w he lo nar energ e th MeV) 0 yo 2 t rag> e E e.( Du neutron yield possibls i t ,i discaro et d sodiu spallatioms a n target A secon calculatiof o t dse s performenwa compositr dfo e targets Pb-Bed ,an U-Be , with the central beryllium zone at different radii. Low energy neutron (E < 20 MeV) distribution in the target is significantly improved. The neutron yield is drastically reduced and the production of Pu isotopes remain high in the case of uranium targets. The neutron yiel U-Br defo targe tdiamete witm berylliuc e h2 th r rfo m zone (22.94 quits )i e similae th o rt yield correspondin leago t d (23.35). A f calculation o thir t se d s performewa s a targed t containin e fuelgth . Thas i t plutonium dissolve wer1 LiFn di same g/ econcentrationTh u .0 eP 10 , 50 , fuee th 10 l f so considered. The maximu 7 m neutron yield achieved is 2.34 at 100 g/1; this means that direct spallation on fuel is not convenient. Taking into account all the parameter considered the most convenient solution to use lead target.

BLANKE. 4 T neutroa r Fo n flux greater tha lO^n/Ccm0 n 1. impossibls i traditionae t i th ) e s 2 us o et l system: solid fuel and cooling water. In fact some blanket proposals consider liquid fuel: actinide dissolved or dispersed in heavy water. The most recent blanket concepts are based on plutonium or minor actinides dissolved in molten salts, particularly 7LiF - 9BeF, used in the Molten Salt Breeder Reactor ExperiencRidgk Oa ee Nationath t ea l Laboratory. A serie f MCNso P cell calculations were carrie above considerint th dt ou a e u P e gth mentioned concentrations and: dissolve- 1 heavn di y wate moderated ran heavy db y water, 2 - dissolved in 7LiF and moderated by heavy water, 3 - dissolved in 7LiF and moderated by graphite.

166 The highest Kinf values correspond to the case of Pu moderated and dissolved in D2O; the minimum one for Pu dissolved in 7LiF and moderated by graphite.

In any case the minimum Kinf value is 1.45550; this means that the three sets of solvent-moderator studie equivalente dar . The outlet temperature of the molten salts can reach values of about 640 °C without increasin MPa 0 pressure e 1 efficience gt th a . Th C ° comparisohavs e a ,0 f w yo e30 O 2 D r nfo the Thermal Cycle for molten salt can reach 42%, against 33% for D2O. e increasTh f efficienco e y means thae samth t e amoun f electricao t l energs yi produced reducing the mass of fission products, particularly the long-lived ones. analyses i ] [2 n dI another important advantag moltee th f eo n salt blankete thath s i t reprocessing. Hybrid systems having lead target wit effectivn ha e radiufissile th f eso 0 zon25 f eo cm and a height of 500 cm reach the criticality for a Pu concentration of 7 g/1 and the corresponding neutro 1 I0n 1. l5flu n/(crcfs i x. Wit s) h that characteristics preliminary calculations demonstrate possiblthas r yeari t pe i t u . P bur o ef t o ng abouk 2 t55 5. CONCLUSION From the results obtained, it is possible to discard solid and sodium targets. The solutions adopting composite target seem t presena , t stage moste th , promising t stilbu ,l remains the problem of the Pu production at a level not acceptable in a Pu burning system. necessite th o t f electricae yo Du l energy production blankee th , t using plutonium dissolve moltedn i n salt permit reaco st highee hth r thermal efficiency. hybrie Th d concept considered results suitabl plutoniur efo m burning.

REFERENCES

[1] M.Victoria, private communication. [2] K. Furukawa, et al., Thorium Molten Salt Nuclear Synergetics, Journal of Nucl. Sei. and Technol 1157-1172 .2 8 (1990)

Next page(s) 7 Lef16 t blank RATIONAL Pu-DISPOSITION FOR U-PRODUCT1ON BY THORIMS-NES (THORIUM MOLTEN-SALT NUCLEAR ENERGY SYNERGETICS)

K. FURUKAWA Institute of Research and Development, Tokai University, Kanagawa

K. MITACHI Departmen Engineeringf o t , Toyohashi University of Technology,Toyohashi

Y. KATO Department of Chemistry of Fuels Research, Japan Atomic Energy Research Institute, Tokai, Ibaraki Japan

S.E. CHIGRINOV Radiation Physic Chemicad san l Problems Institute, Academ Sciencef yo , Minsk, Belarus

A. LECOCQ European Associatio Molten-Salr nfo t Reactors Development, Orsay, France

L.B. ERBAY Departmen Mechanicaf o t l Engineering, Osmangazi University, Eskisehir, Turkey

Abstract

Plutonium should not be simply burned out and be used to produce a new fissil , becausU r esolvin fo eOO e Qglobath g l energy/environmental problem nexe th tn i scentur yhuge-siza e Fission Industry shoul estabe db - lished preparing huge fissile materials t establishee seemI b . o t s y b d THORIMS-NES [Thorium Molten-Salt Nuclear Energy Synergetics] composed of simple Molten-Salt Fission Reactors [MSR: FUJI-series] and 233U producing Accelerator Molten-Salt Breeders [AMSB] .facil h ThesT w -ne e ities wi 11 be developed by uti 1 izing/ el iminating the "excess Pu" separated from Pu weapon-heads and the spent-fuel of U-Pu reactors. A semi-real scenari r proceedinofo g suc he firs worth n nexe tki th hal tf o centurf s yi presented. THORIMS-NES will have big advantages not only in safety and radiowaste issues t als bu n nuclear-proliferation/terrorisi o, d man economy guaranteeing low R & D cost.

1.INTRODUCTION dispositiou P e Ith n n issue r targetou , n finai s l stag: on e t shoulpu e b d (A) nearly complete elimination of the world's Pu stocks [1], in parallel achievin e followingsth g :

169 (B) economical utilization of energetic potentiality of Pu (and TRU) as an useful producer of neutron or fissile materials, not simply burning out, (C) real establishment of new/rational huge-size nuclear industry to solve the global energy/environmental problems in the next century.

The real meaning of the target (C) will be explained in detail in Chapter 2. To establish this (C), the target (B) should be proceeded in the best way. Such an idealistic Pu disposition work might be real ized by the modification of our Th-breeding synergetic system: THORIMS-NES ("Thorium Molten-Salt Nuclear Energy Synergetics [2], which is composed of [I] simple fission power stations [Molten-Salt Reactor (MSR) : FUJI-series], [II] fissile-fuel producers by spalla- tion/fission reaction GeV-proto1 f so n Accelerator Molten-Salt Breede AMSBr: d ,an [IIIy procesDr ] s plants. Detailed explanation of this new technology modifying the original THORIMS- NES in general has been given in another paper of this meeting by Lecocq et al. [3] . A brief explanation will be given in Chapter 3 too, in which especially a little detail of AMSB and A-plan will be mentioned. Its real application in global strategy of next century will be discussed in Chapter 4.

2. FUTURE NUCLEAR ENERGY DEMAND

The next 21st century will be a transient period from fossil fuel age to solar age through nuclear energy era, as a global and especially local climate could not accommodat excese th e s heat emission several times bigger thapresene th n t arti- ficial heat generation. Therefore heat-emissioe ,th n type energy technologies (even nuclear fusio r satellito n e electric-generation )e utilize b coul t no ds majo a d r 22ne th oned n i centurys . Such advanced prediction wil understooe b l d fro illustratione mth Fig.1n si , basically dependin d extendinan g g Marchetti's predictio e futurth n o ne energy [4, 2, 5]. If we tentatively accept the same global energy growth rate of 2. 3%/year e past necessarth e th ,s a y fissio OOP. 2 n "- energ TWe-YeaP OO e y, th wil"1 n i r e b l next century. Thi "50s i s 01,000- " times larger tha pase nth t (peaceful) fission energy production of only "2" TWe-Year. (In here, we have to recognize that even such huge nuclear energy will not be enough to solve the COp Greenhouse effect as . ) show] [4 n Fig i n) (C .1 Suc hhuga e total amoun steea d tan p increase correspondin abouo t g year0 1 t s f doubling-timo e nucleath n i er energ e achievey b e presen wilt th no ly b d t U-Pu solid-fuel cycle system suc s LWRha , HTG severao d LMFBRt Ran e du ,l difficulties connected with (a)breeding abi1ity safet) (b , y [including severe accidents], radio-waste) (c s [including productio f trans-Uraniuo n m elements] nuclear) ,(d - prol iferatio d -terrorisan n m [including Plutonium-elimination] d (e)publian , c d institutionaan l acceptances related wite technologicath h l simplicity, flex- ibi1ity and economy in the global applications (cf. the left column of Table n referenci 1 . ) ] [3 e Therefore, we have to prepare an essentially new technology.

170 PHILOSOPHW NE 3. : YTHORIMS-NE S ---THORIUM MOLTEN-SALT NUCLEAR ENERGY SYNERGETICS---

3. 1. General principles

Already, to solve the several energy problems shown in Fig. 1. a new philosophy name THORIMS-NEs a d bees Sha n propose e publicth t o I t ddepend -e followth n o s - ing three principles [2, 5]. [I] Thorium utilization, [II] Application of molten-fluoride fuel technology, [III] Separatio f fissilno e producing breeders (process plants—AMSB:Accelerator Molten-Salt Breeder) and power generating reactors(utility facilities—MSR: Molten-Salt Reactor).

artificial heat emission

1850 1900 2050 -200 y Historical Trend Primary Energy Consumption TW (B) in Energy Substitution

J^^=^^ M. Nakisenovic1984 -^~ 0.1 1850 1900 1950 2000 COa Yearly Emissions (C) 2065

1850 1900 2000 2100 Fission Energy Production^*,. (D) TWe 2000 TWe- Y Y • e TW 0 90 1 =10,000• e 0GW

• 5

2000 2050 2100 2150

Fig. 1 Global Future Energy Prediction. a furthe [(As i ) r extensio f Marchetti'o n s estimat f historicao e l trend in energy substitution; (B) the prediction of yearly growth rate in the past primary energy consumption; (C) the COg yearly emission from several fossi e nucleath l) fuelsr(D fissiond an ; - energy ] productio . ) (B & n) base(A n o d

171 These faci 1 it les followed by chemical dry-process plants will be easily coupled each other preparing a simple "Molten-Salt Fuel Cycle" with a minimum handling and transportation thin I . s fuel-cycle several radio-waste specie. sP . includinF d an U gTR wil e destroyeb l d gradually withou y separatioan t n works [6].

ooo 3.2. Modification for Pu-burn and U-production

. I Genera. 32 . l approach

The modification of THORIMS-NES might be established following on the three plans: (1) D-plan: Pu (and Trans-Uranium elements [TRU] ) fluoride separation by Dry-proc- ess from the spent solid-fuels accumulating in the world. The technological basis has been examined by French, Russian, etc. . (2) F-plan: Pu-burnin 233d U-productiogan MSy nb R [FUJI-Pu mose Th t . ]rationa l safe technolog effective th r fo y e Pu-incineratio napplication a wil e b l f MSno R fertilize Thy db . The detai f neutronio l c examinatio reportee b 1 1 Mi y nacdit w t b e h i n thii s. al meeting [7]. (3) A-plan: Pu-burnin 233d U-productiogan AMSB-Puy nb n parallei , l wit- F h plan. AMSB-Pu was mostly examined and is being improved by the cooperation of Chigrinov's group, Minsk [8].

. 2 FUJI-P . 2 . 3 u A simple rational design of MSR named FUJI has been proposed including a fuel-self-sustaining type FUJI-II already [9], and will be explained including Pu-burning version: FUJI-P réfn i u . [7]. The significant advantage of MSR in burning dismantled weapons fuel has been excellently examine ORNy b d L peoples [10]. Soon after dismantling weapon headf so rtop- * easiln ca ) yU conver r Pfluorido (o t ut e dilutin othey gb r fluorides. This salt mixture itself is fairly resistive to nuclear-proliferation/diversion. If nece- ssary, it wi 11 be improved more by spiking by high gamma active fluorides (including Co. Ru, Cer denaturiny o b . Pu r o ) g y Ub U . 60 106 144 238 235 238

. AMSB-P3 . 2 . 3 u

The basic ide f AMSao inventes Bwa n 198do 0 dependin e "singlth n o ge fluid type Molten-Salt target/blanket concept"[11], whic s significantli h y simpld an e practica structurn i l operatiod ean n [Fig . Th.2] e target/blanket vesse simpla s i l e pof 4.5o t n depthmi n diametei m 7 . d Protoan r n beam wil injectee b l n offi d - center position of molten-salt vortex. Therefore, several serious technological prob- lems related with (i)material compatibility and radiation-damage, (ii)heat removal, (iii) spallation chemistry, and (iv) target shuffling (uniform continuous reaction mostle ar ) y abl solvo et thiy eb s design concept excep proton-beae th t m injection port engineering. Dependin e initiath n go l developmental effor f MSo t R basic technology, A-plan developing AMSB-Pu wil proceedee yearw b lfe sa delan i d y from FUJI-PuA . typical compositio f target/blankeno t salts i s 7 233 239 LiF-8eF2-ThF4- UF4- PuF3 = 64-18- [ 18- (X+Y) ] -X-Y mol«: [X, Y = 0~0. 7]

172 Fig Schemati. .2 c set-u f single-fluid-typpo e Accelerator Mo-1 ten-Salt Breeder (AMSB). [system-size] : 4.5m in diameter, 6m in depth.

1 rProton beam; 2:Salt inlet; 3:Salt outlet; 4:Reactor vessel ; 5, 6:Graphite: 7:Primary loop; 8:Heat exchanger; 9:Main salt pump; 10:Throttle valve; 1 1 :0verf low line; 12:Storage tank; 13:High-pressur salt inlet ; 14:Gate valve ; 15:Vacuum line; 16: Vapor trap; 17:Duct; 18:0rifice; 19:Focussing magnet ; 20:Bypath for salt processing; 21:Pump; 22:Process facility; 23:Filter.

The most importan f AMSB-Po tD R& itemfollowings[12e ur th wil fo se b l : ] 6 , ,13 (A) 1 GeV. 100-300 mA proton Linear Accelerator: In the target/blanket salt system a 1 ittle lower voltage, 1 GeV, wi 11 be convenient, although it should be optimized in final design. Proton will penetrate into salt more than 1 m, where is a most in- tense reaction zone, and such situation is effective for heat removal. As a long term program, more economical non-monochromatic beam accelerator development shoul e encourageb d n industriaa s a d l machine than Linac. (B) Injection port engineering: It is amost serious unclear item. However, the vapor of salt might be mostly condensable on duct wal 1 (cf. Fig. 2), applying several additive techniques such as electrostatic col lection. Gaseous species in molten salt should be carried away to be separated in outside of reactor core. Such practice might be solved by the real beam test increasing its intensity stepwise.

173 ) Accurac(C f neutronio y c calculations e target/blankeTh : t salt syste f AMSmo B contains several kin f nucleo d i including light ones. Therefore, neutronic cal- culation is not easy and low accuracy in reaction products yield and in heat gener- ation rate yet. Furukawa and Kato unsuccessfully aimed to proceed the experimen- tal analysi f largso e target salt block gettin N (no helwe SI gth f pPSIo ) groun pi w againo t n bu plannin1981, undet i gcooperatioe rth n with Russian group wilt I . l be valuable developmenth r fo e f spallatioo t n theore generaln i m , too. ) Reacto(D r chemical aspects: Several chemical issues relating with "spallation chemistry" has been successful ly examined [12, 13,6] depending on chemical basis of R developeMS ORNy b e chemicad Th L l processing procedure f salso t wil more b l e flexibl r substantiallou n i e y subcritical system e transmutatioTh . f hazardouno s radioisotopes could be proceeded by the minimum separation work simply circulating through target/fuel salt cycl THORIMS-NESn i e . (E) System engineering design optimization: The size of target/blanket system is important relating with [a] radiation damage of reflector graphite and reactor vessel wall, [bj inventory of fissile, fertile and reaction products for resulting the most important optimal reactor performance d finallan , tota ] [c yl economy. Its optimization connecting with salt composition and the operation purpose of this facility shoul improvee db d dependin progrese th n go severaf so l technological items.

A SCENARI4. O OPENING THORIUM-ERA Y APPLYIN'B G PU-BUR 233D U-PRODUCTIOAN N N

4.1. Technological basis

To establish the targets (A), (B) and (C) mentioned in Chapter 1, the three plans, D-, F- and A-plan, have been proposed (cf. sec. 3. 2. 1.). If we proceed the above plans to burn Pu and produce oooU, these efforts would fully contribut o prepart e e reath el standard THORIMS-NES finalld an y automatically .

The conventional methods for Pu disposition are depending on solid-fuel con- cepts, which are not suitable for establ ishing the targets (A) (C). They are not ful ly eliminating Pu, or need repeated processing and fuel-refabrication or in a weak neutron economy. Such situation e illustratesar whicn i , h 3 Fig n advantage.i d f so the FUJI-Pu and AMSB-Pu are also shown [5]. e conservativTh e performance f FUJI-IIo s , FUJI-P d AMSB-P- an uPu n i u ooo burning/ U-production are presented in Table I. In future, FUJI-II and FUJI-Pu shoul understooe b d d that they would include IGWe-class superFUJ superFUJI-Pr Io u of higher performance than thos d AMSB-Pf Tablo ean , I e u would als e ablb o o t e improve more.

Table I. PRELIMINARY PERFORMANCE OF FUJI [PER IGWe]

AMSD AN B [1GeV-300mA]

233 Pu-invU-inv. 233 . Pu burn/U prodY . Elec. output FUJI-Pu 3 t 0. 86 t 0. 7 t 1 GWe FUJI-II 2 t self-sust. 1 GWe AMSB-Pu 5 t 5 t 0. 6 t t 1.0 1 GWe

174 CONCEPTUAL PRESENTATION OF FISSILE-MATERIAL CHANGE IN SEVERAL PU-INCINERATION REACTORS 1993,6 K,F.

Pu inventory Pu burned X*(W7777,1 initial Pu produc. charge 235 233 —(burned) u/ u 2"g produc.

added X : no addition 0 : addition Pu fertile addition blanket

(I) Solid-Fuel Fission Reoctors 1~3 years later

•PU partial 238U Pu-thermal LWR charge rPu reprod. F(B)R •PU full charge 238,, HTGR

FR • PU full charge HTGR

FR LWR . •PU full charge 232Th HTGR

•15 years later

•PU full charge f/f/f HTG? R

(II) Liguid(Fluid)-Fuel Fission Reactors

•PU partial 0 232-•Th charge

•PU full charge 0 232Th

(III) Sub-Critical Accel.(Spoliation) Reoctors (out of reac.)

•PU full charge 0 X

•PU full charge 0 232Th

ion iii; (ou f reac.to )

Fig. 3. Conceptual Presentation of Fissile-Material Change in Several Pu-Incineration Reactors.

175 4.2. An example of scenario to open THORIMS-NES era

Proven U-Pu cycle systems woul realizt dno energe eth y production predicten di e social) owinth Fig(D 1 o .t g, technologica d economicaan l l reasons. However, THORIMS-NES would be able to realize the several global-energy scenario applying the above D-, F- and A-plans [5, 11]. Here, one of the simple examples has been show d FigTabln an i n.4 [cf I I e. 14].

Tabl REPLACEMEN. II e U-PF TO U REACTOR TH-Y SB U REACTORS ELIMINATING PLUTONIUM [cf. Fig. 4]

Total U-Pu Cumu 1 at i ve D-Plan F-Plan A-Plan Energy Stations Pu Pu supp. FUJI -Pu; FUJI AMSB-Pu (GWe) (GWe) (ton) (t/lOY) (GWe) (fac. /10Y) 2000 300 300 0 1,20 2010 550 550 2,550 start start 2020 1,000 850 0 4,65 730 0 5 100; start 2030 1, 850 1, 000 5 7,42 0 4,89 0 20035 ; 300 2040 3,460 1,050 10, 525 4, 860 ; 1,810 0 650 2050 6,800 600 12, 825 2, 700 0 70 , 05 ; 500 2060 9,300 100 135 ,87 580 010 , 9 ; 0 100

New Pu 233y 233U Systems burn produc. residual (GWe) (t/lOY) (t/10Y) (t/lOY) 2000 2010 2020 150 430 350 250 2030 850 3,090 2,550 700 2040 2,460 1 3, 710 0 45 , 5 ,480 2050 0 6,20 0 3,45 5, 750 -200 2060 0 9.20 2. 580 4.000 -800 [/10Y : cumulative valun i e 0 year1 i st e date1th 1 ]

Tentativel totae th y l electric productio U-Pf o n u fueled power stationn i s future wil assumee b l 4 time s a ds large maximun i r m thapresente th n . Eveo s n low this will still produce more than 10 ton Pu (assuming 300Kg/GWe-Ynet) till 2050n appl heren ca presene I . e th y w , t Purex process plants ,might i too t t bu , not become significant due to the high capital/operation cost. As soon as possible, D-plan should be started to prepare more proliferation-resistant mode process, because a simple storage of spent-fuels will be non real solution. Separated Pu shoul d A-plan an e burne b d- F t soo. y ou db n In the stage before realizing D- and F-plan, dismantled weapon-heads Pu (and Pu from spent-fuels, too) should be converted to fluoride in the diversion protec- tion for t beinmye evet g no ndecide e fina R th dfuel-sal MS l t composition [cf[10]1 .. 2 sec. .2 . 3 Pu (TRU) disposition could be started from 2010 by F-plan, and from 2020 by A- pla paralleln i n formee Th . r activity wilmaximun i l e becomGW m 0 scale20 e about

176 10 20 x103 GWe (net electricity)

8- Total Nuclear Power -15

6- Pu separation/burn 10

4- Pu production

2- Nuclear Power (U-Pu) - 1

0 0 2000 10 20 30 40 2050 60

Fig. 4. A Scenario for Pu Disposition in Next Century [cf. Table II & Fig. 1]

2030, burning about 2, 600 ton Pu (TRU) or more. The latter will become 650 facili- ties in peak about 2040, burning about 10,000 ton Pu (TRU) or more. The duty of FUJI-Pu will be finished until 2040. However, it can be operated propes a r Th- U power stations (FUJI reactof o d ) en tilre lifth l e withouy tan 233 modificatio e technologicaTh . n l developmen f AMSB-Po t u wil e significanb l t among 2020 and 2040. The initial AMSB-Pu will be in lower grade not producing any outer electricity. Afterward its new version wi 11 produce excess electricity improving in'performance by near critical condition, which would allow an excess electric production of 1 or 2 GWe/facility [5]. Afte middle th r f 2040'o e s decade whicn i , woul u P h almose b d t eliminated, ooo AMSB-Pu should be gradually dismantled recovering U fissile, which is strongly effective for initiating a lot of FUJI power stations. Therefore, the main leading rol0 years 4 f AMSB-P o e- e perio,0 3 th althoug un f i o dwil e b l h AMSB wile b l continuously usefu radio-wastr lfo e (including minor actinides) incineratioa s na minor purpose [6] .

5. DISCUSSIONS

5. ADDITIV.1 E TECHNOLOGICAL RATIONALITY

5. 1. 1. CONTAINER MATERIALS: Fortunately, an easy manufacturable, weldable d higan h temperature resistant alloy compatible with molten fluoride s beeha s n developed basically depending on the effort of Dr. H. Inouye, ORNL. This alloy.

177 Hastelloy N [Ni- (1 5-18) Mo- (6~8) Cr in weight K], has been modified adding about e protectioth r e attacT fo f b 1n o surfacM %o k e [15]. Thi s experimentallswa y supporte y Kurchatob d v Institute [16]. This alloy doesn't need to be introduced in reactor core region suffering any severe irradiatio thermad nan l shock. [MSR cor occupies ei graphity db d fuelean - salt only. mose Th ]t thin alloy part wil aboue b l t half inch diametee r tubth n i e intermediate heat exchanger. Therefore reactoe th , r design woul muce db h simple and easy. In practice, the corrosion of Hastelloy N was surprisingly small even in the small experimental reactor MSRE due to the negligible contamination by air and moisture. Larger commercial power stations woul muce b d h safe corrosionn i r .

. FUJI-Pu2 5cooperatioa e . coul.w 1 t f I ge d : n with ORNL constructioe th , f no pi lot-plant: 7MWe-miniFUJ establishee b 1 1 i Iw d less tha yearn7 s (excep ceni tg 1 sin time) using reserved MSRE fuel-salt. The first FUJI-Pu might be operated among 15 years from now[2, 5].

. AMSB-Pu3 . 1 physicas . It 5 : l basi simpls si cleard ean . After preparin mA-0 1 g 1 V protoGe n Linac target/blanke. S . M , t experiments wil startee b l yearn i d s later. Except the development of high current Linac, a beam-inject port engineering wi 11 be most important R & D item. However, the vapor of salt might be mostly condensable on duct wall, applying several additive techniques such as electrostatic col lection. Gaseous species including T should be carried away by molten salt flow e separateb o t t outsida d reacto- of e r core. [cf] . 3 sec. 2 . .3

5. 2. SAFETY

MSR has essentially no severe accident, and subcritical AMSB is more safe. Therefore, engineering safety of THORIMS-NES is significantly high [2]. The other safety features wil showe b l [3]. y Lecocb nal . t e q

5. 3. Anti nuclear-proliferation and -terrorism:

In THORIMS-NES, 232U content in fissile 233U will become more than 500 ppm [4]. Therefore, the very hard gamma (2. 6 MeV) activity of daughter Tl in ooo explosive 10 Kg U will emit nearly lethal dose of 200 rem/hr at 50 cm dis- tance, and the protection shield will be required lead more than 20 cm in thick. hardese th s Id leasti tan t desirable materia safeguard/detecweapons r it lfo d ,an - tion is very easy.

Alternative Approach for diversion protection: FUJI-Pu woul locatee db safe-guarden di d sites inside countries with extensive nuclear experience. Ordinary FUJI without Pu and without processing could be outside those states (the inside/outside concept). If O'SO U content in 9^^ U is comparatively low in early stage of this fuel cycle, isolated UF^ would be instantaneously denatured to 12% 3U by natural U. This treatment wilt producno l y seriouan e s difficulties, because denatured 233U-fuel of FUJI is a fluoride mixture of 233U, 238U, 232Th in the composition of 0. 2, (0. 2 x 7 =) 1. 4 and 12 mol% respectively, which produces 239Pu about 10

178 ooo times smaller than U. Especially in the case of fuel-self-sustaining FUJI, Pu O^R conten becom1 1 i w te negl igible yearsamonw fe ga . (Sim r denatureila versioU d n f FUJo s i nominateI s FUJI-a d discussed Van d already [17) . ]

5. 4. Economy

Significant economical advantag f THORIMS-NEo e s beeha S n concludea s a d result of preliminary comparative examination between superFUJI and LWR. The main reasons are mostly depending on (1) negl igible fuel cost (only an addition of Th ooo and small amount of U in full life) comparing with the same order of magni- tude as the construction cost in LWR, (2) simple reactor core structure without fuel assembly and fuel handling machine (core internals are only graphite and small number of control rods), (3) very high thermal efficiency more than 46% by ultra- supercritical steam turbine, (4) simple safety system because of the higher safety smad reactor an l 1e R Idingi MS rbu f higheo ) , (5 etc d r capacit.an y factor becausf eo no necessity of refueling and vessel-opening works. (6) In this fuel-cycle system nuclear material transportatio continuoue th o t e du sver s fuew ni yfe l burningd ,an a verw additiofe y f fissile/fertilo n e materials only. AMSB will be much expensive than 1 GeV superFUJI. However, it will not be important, because its total activity in the scenario of Table 4will be only about 15 TWe'Y, less than 2, or 1 % of total 1,000 or 2,000 TWe-Y capacity.

5. 3. R & D PROGRAM

Its technological basis of F-plan has been prepared by the excel lent efforts of ORNL among 1947-76(15]. Therefore commercializatioe th , f FUJI-Pno smallen i u r size (100-300 MWe) will be performed in 15 years at least. Such public nuclear power stations should be simpler in configuration/operation/maintenance, and power-siz Fissios ea flexible ik Spar 1 no t l latioeno n Breeding power stations [MSBR or AMSB]. Depending on the above initial developmental effort of MSR basic technology, A- plan developing AMSB-Pu will be proceeded in about 10 years delay from FUJI-Pu. Its optimization connecting with salt composition and the operation pur- pose of this facility should be proceeded depending on the progress of several technological items. The time schedule of developmental program has been shown in Table III.

Table III. DEVELOPMENTAL PROGRA F THORIMS-NEMO DISPOSITIOU P R SFO N [ ——— > : R&D ; ———— : operation ]

2000 2010 2020 2030

miniFUJI-Pu ———————: FUJI-Pu —————- ——————— ;————— ———— (FUJI)- superFUJI-Pu ——————— (su oerFUJI)

- — —— AMSB proto AMSB-Pu >-., istfc+i; (hg— ) ——g —(s - )(AMSB )

179 . 6 CONCLUSIONS It has been demonstrated that a modification of THORIMS-NES accommodating Pu-burnable versions FUJI-Pu and AMSB-Pu might achieve full burn-out of Pu and openin U fue°° (pura f global- o g er cycleTh eh T l ) tilmiddle th l f nexo e t cen- OO'S tury. This strategy will be fundamentally public acceptable and economical due to the following reasons: (l)simple, continuous and flexible operation and mainte- nance works for Pu-disposition;(2) safe: no severe accidents; (3)anti nuclear- proliferatio d -terrorise stronan nth o gt e gammdu m a activit f accompanieo y d OO fuel/targen Oi U t sal THORIMS-NESf to ; (4)ver w radio-wasteyfe nearlo t e sdu y no TRU and few maintenance/operational works; (5) very few nuclear materials transportation; (6) small R&D cost owing to few items: (7) high economical poten- ) rea(8 tiall d establishmenan ; f idealistio t c Breeding Fuel Cycle, In conclusion, dispositio u P here e th , n issue shoul managee db d promisino t g solve the energy/environmental problems including world poverty symbolized as "North-South" problem in the next century. This will be a reply to the final proposal of David lilienthal [18]. We are expecting to start the worldwide debate for international col laboration aiming further improvements in this field.

ACKNOWLEDGEMENTS e authorTh s wis o exprest h s sincere thank o mant s y friend f USAo s , France, Russia, India, Switzerland, Canada, Turkey, Germany, Belgium, Japan and the other countrie r theifo s r kind cooperatio d encouragementsan n .

REFERENCES

[l] Comm. Int. Security & Arms Control, Nat. Acad. Sei. , "Management & Disposi- tion of Excess Weapons", Washington: Nat. Acad. press, 1994, pp. 2-3. [2] FURUKAWA, K. , LECOCQ, A. , KATO, Y & MITACH I. K. , "Summary Report :Thorium Molten-Salt Nucl. Ene. Synergetics", J. Nucl. Sei. Tech., 27 12(1990) 1157-1178. [3] Lecocq, A. , FURUKAWA, K. , "Rational Pu Transmutation for 233U", this Meeting. [4] MARCHETTI, C. & NAKICENOVIC, N. . "The Dynamics of Energy Systems and Logistic Substitution Model", RR-79-13, Internat. Inst r Applie,fo d Systems Analysis, (1979): MARCHETTI Nuclea, . ,C r Plant d Nucleaan s r Niches:0e th n Generatio f Nucleao n r Energy durin e Lasth gt Twenty Years, Nucl. Sei. Eng.,

90, (1985)521-526; MARCHETTI, C. , "How to Solve the C02 Problem without Tears" Worlh 7t , d Hydrogen Conf .Moscow, , Sept. (1988). FURUKAWA] [5 , CHIGRINOV . ,K , KATO& MITACHI . E , . ,,S Y. , "Accelerato. ,K r oqq Molten-Salt Breeder Power Reactor useful for Pu-burning and °U Production", Emerging Nucl. Energy Systems (YASUDA, H. , Ed. ) World Sei. , Singa pore (1994) 429-433. [6] FURUKAWA, K. , LECOCQ, A. , KATO, Y. and MITACHI, K. , "Radiowaste Manageme- Globan ni t l Appl f Thoriu.o m Molten-Salt Nuclear Energy Synergetics with Accel. Breeders", Specialists Meet Accel.-driven ,o n Transmutation Techr fo . Radwaste & Other Appli. , Saltsjobaden, Stockholm, LA-12205-C, (1991)686-697. [7] MITACHI, K. , FURUKAWA. K. . YAMANA, Y. . SUZUKI, T. & KATO, Y. , Neutronic

180 Exam. Pu-Transmu y Smal.b l Molten-Salt Fission Power St.thin i , s Meeting. ] CHIGRINOV[8 EVITSKAYAI K , . ,S PET, . , A LTSKY I , . ,RUTKOVSKAYAV & . ,K FURUKAWA "Calcu, . ,K . Metho f Energo d y Systems base n Higo d h Energy Particle d Nuclean i Accelerators", Emerging Nucl. Energy Systems (Proc h Int.7t . Conf., Makuhari, 1993) Worl) (YASUDA . Ed d . Sei. , H Singapor, . e (1994) 434-438;KATO, . ,Y FURUkAWA, K. , MITACHI, K. & CHIGRINOV, S. E. , "Fuel Trajectory in Accel. Molten-Salt Breed. Power Reac. System includ. Pu Burning", ibid. (1994)439-443. [9] FURUKAWA, K. . MINAMI, K. , OOSAWA, T. , OHTA, M. , NAKAMURA, N. , MITACHI, K. , KATO, Y. , "Design Stud Smaf yo Molten-Sall1 t Fission Power Station suitabl Coupr efo g lin with Accel. Molten-Salt Breeders", Emarg. Nucl. Energy System (Proc.4th Int. Conf. Madrid, 1986) (VELARDE, G. . Ed. ), World Sei. . Singapore (1987) 235-239; FURUKAWA, K. , MITACHI, K., KATO, Y. , Small Molten-Salt Reactors with Rational Thorium Nucl. Engineering & Design, 136(1992)157-165. [10]GAT , ENGEL & DODDS . , Molten-SalU . R L . . ,J H t Reactor r Burninfo s g Disman- tled Weapons Fuel, Nucl. Tech (19920 10 ., ) 390-394; ENGEL, GAT "Disman. . U ,R . ,J - tled Weapons Fuel Burnin Molten-Saln i g t Reactors", Global'93. Shuttle (1993). [1 13FURUKAWA, K. , TSUKADA, K. & NAKAHARA, Y. : "Molten-Salt Technology on the Spallation Neutron Facilities", Proc.4th ICANS, (1980)349-354; Single- fluid-type Accelerator Molten-Salt Breeder Concept Nucl. J , . Sei. Tech., 18(1981)79-80. [ 12]FURUKAWA, K. , KATO, Y. , OHOMICHI, T. & OHNO, H. , "The Combined System of Accelerator Molten-Salt Breeder (AMSB d Molten-Salan ) t Converter Reac- tor (MSCR) ", Thorium Fuel Reactors (Proc. Japan-U. S. Semi. Th Fuel Reactors, Nara, (1982) Atomic Ene. f SocJapao . n (1985) 271-280. [13] FURUKAWA t al.e ,. , K "Accelerator Molten-Salt Breeding System", First Int. Sympo. on Molten Salt Chemistry & Technology (Int. Conf. , Kyoto, 1983) J-302, 399-404, J-303, 405-408, J-304, 409-413, K-210. 497-499. [14] FURUKAWA , Kato ,CHIGRINOV& K. . ,Y "Plutoniu, . E . ,S m (TRU) Transmutation

OOO and U Produc. by Single-Fluid Type Accelerator Molten-Salt Breeder (AM- SB) ", Int. Conf. Accel.-driven Transmu. Tech. Appl. , Las Vegas, July (1994). [15]ROTHENTAL, M. W. , HAUBENREICH, P. N. , BRIGGS, R. B. , "the development Status f Molten-Salo t Breeder Reactors", ORNL-4812 (1972 ; ) ENGEL GRIMES, . R . J , , W. R. , BAUMAN, H. F. , McCOY, H. E. , DEAR ING, J. F. , RHOADES, W. E. , "Conceptual Design Characteristics of a Denatured Molten-Salreedert Reactor with Once-through Fuel ing", ORNL/TM-7207 (1980). ] [16 NOVIKOV , IGNATIEV . M . , B ,. FEDELOV V CHEREDNIKOV. & V , . I . V , . MoltenH . V , - Salt Nuclear Power Facilities: perspectiv problems& e , Energoatomizdat, Moscow (1990). FURUKAWA] 17 [ , LECOCQ . K , . .MITACHIA , & KATO . , K .. Globa Y , l Measurr fo e Energy & Environmental Problems by Thorium Molten-Salt Nuclear Energy Synergetics, Materials Sei. Forum, 73-75(1991) 685-692. [l8]Lilienthal, D. E.. Atomic Energy: A New Start, Harper & Raw, N. Y. (1980).

Next page(s) Left blank NEUTRONIC EXAMINATIO PLUTONIUN NO M TRANSMUTATIO SMALA Y NB L MOLTEN-SALT FISSION POWER STATION

K. MITACHI . YAMANAY , . SUZUKT , I Toyohashi Universit Technologf yo y Toyohashi

K. FURUKAWA Tokai University, Hiratsuka

Y. KATO Japan Atomic Energy Research Institute, Tokai

Japan

Abstract Neutronic calculations were made to examine the conceptual

feasibility of the small molten-salt reactor, FUJI-Pu, for effective plutonium diminution and effectivU e233 productio n to use thorium as energy resource. Thoug e reactoth h n thii r s repor s almosha te samth te features e reporteon e th d f o previously e authorth , s intende o consumt d ed an mor u P e to produce more 233y changingb U e reactoth - r operation mode, i.e. yearly isolatio f uraniuo n m isotope y applyinb s g fluorinatio d evaporatioan n n method. Burn-up characteristics were calculate 0 day90 sr operatiofo d t a n rated power by using ORIGEN-2. The results show that the reactor can transmute 980k f plutoniugo m isotope d producan s e 489k 233f o gU annually fo e electricitr th lGW f o e y generation. Hence e reactoth , - ef r n coula e b d ficien n u efficiena burneP t d an r t 233U producer, comparing with LWRs operated with MOX fuel.

1. Introduction

For the effective disposition of weapon-head Pu and reactor grade Pu, the applicatio f Th-o Un 233 recycle syste m e bests practicamighth It e .b t l d economicaan l realization depen n fluid-fueo d e higl th concepth o t e du s gamma activity of 233U fuel, which is highly efficient for the prohibition of nuclear proliferatio d terrorisman n . Among several fluid-fuel concepts e molteth , n fluorides fues wa l verified to be the only usable liquid fuel by the long effort of Oak Ridge

National Laboratory in 1947-1976. They aimed to establish a fission breed- ing power station MSBR. [1J However, it has several serious difficulties in design and global application in future. Therefore, our group is proposing more rational approach, named as THORIMS-NESC2], which will be reported by another papers presented in this meeting1^' ^. This symbiotic system uses simple size-flexible molten-salt fission power stations. Its significant characteristics are not only high safety and high simplicity in configura-

183 tion (no core-graphite exchange and no continuous chemical processing), but also few works in operation, maintenance and transportation due to the idealistic near-breeder character even smaller 150MWe molten-salt reactor, named FUJI- H C5]-[6]. Such small molten-salt reactors coul e purpose applieb dth u P r f o fo ed disposition. In recent two years, we have been examining this issue in several mode o evaluatt s d improvan e performance^s it e f theo me wilOn . l e reporteb d here, FUJI-Pu n whic i ,e initia th h l fuel sal s composei t f o d fully Pu fissile and Th fertile components. The reactor will be operated with successivd almosan , t Th yearl ed additioan yu P isolatio f o n 233f o nU by applying simple fluorination and evaporation method. This operation mode increases neutron econom d minimizean y U sburning, resulting higher

additio d consumptio d highean n an u rP productiof o n 233e f reactoo nTh U . r 233 will be able to realize by few R&D work and cost within 20 years and open a withoua er ne. h T wPu t

. Initia2 l core characteristics

2.1 Calculation method

A schematic diagram of the reactor in the present study is depicted in Fig.l e primarTh . y feature reactoth f o er cor s almosi e e sam th s FUJIta e - 11. It is a graphite moderated reactor in which the single fuel-salt con- taining fissile and fertile materials serves as both the fuel fluid and the blanket fluide reactoTh . r cor s dividei e d into three zones: Core,

Axial- /Graphite o• directio / ncontainin g boron ———————— t — jf —————————— co' /Vessel * Reflector 1% • C3 Blanket(r) •^r' r- * Blanket(z% 22 ) / co C3, OJ

OJ OJ Core • 20% 17% e 0.05 Radial- ' direction 1.57 3 6 0,3 3. ,0 0. 1 2. 65 Uni t :m

Figur 1 eReacto r mode e calculationl th use n i d e percentageTh . n i s the figure are the fuel salt volume fraction in each zone of the reactor.

184 Blanket- d Blanket-zan r e reactoe operateb Th . n ca r d without continuous chemical processing and without graphite replacement for the period of 30 years, i.e. the plant life. Ten group cross section data (6 for the fast neutron flux and 4 for the thermal neutron flux) are first calculated for all zones of the reactor by usin e SRAth g C code systeme nucleath d [san r] data file, JENDL-3, develope t Japaa d n Atomic Energy Research Institute. Then e effectivth , e neutron multiplication factor, neutron flux distributio d fissioan n n reac- tion distribution in the reactor were estimated by using two dimensional diffusion code, CITATION.

2 Semi-optimize2. d reactor core

It is desirable for the reactor to have higher fuel conversion ratio, smaller plutonium inventor d lowean y r neutron radiation damage A serie. s of survey calculations were conducte o obtait d e overvie e th nuclean th f o wr characteristics of the initial cores, widely changing the parameters such e concentratioath s f plutoniumo n e fuel-salth , t volume fractioe th n i n core e reactoth , r geometry, etc a result .s A a semi-optimize, d reactor cor e plutoniu th suc s a h m burner, FUJI-Pu s designee conditiowa ,th r fo d n of acceptable exposure on graphite, 6.3xl013n/cm2/s for the fast neutron flux (>50keV) e exposurTh . n Modifieo e d Hastello s limitewa N yo t d 2.3xlO e fasnn/cmth t r 2/neutrofo s n flux (>0.8MeV d 1.2xl0an ) 13n/cmr 2/fo s the thermal neutron flux (<0.18eV). These limitations on the irradiation cited from the reference[l] can ensure the reactor operation for about 15 equivalent full power years. Thi equivalens i s o about t 0 year3 t f reaco s - loae th d t a facto n to f 0.5ru ro r . The fuel-salt is a molten-salt mixture of the fluorides of lithium, beryllium, thorium, uraniu d plutoniuman m e compositioTh . e initiath n i nl cor s listei e isotopi Th n Tabli d. I ec compositio f plutoniuo n m supplied e reactotth o a typicas showi r r Tabln i nfo lr n e fuefo R l LW burne a n i d about 33GWd/ton of burn-up and 1 year cooling. Several important parameters of FUJI-Pu are listed in Table ffi. The reactor vessel is 5.3m in diamete d 4.7an r n heightmi e reactoTh .n generat ca r e 250MW f heatto h . The temperatur f fuel-salo e e reactoe outleth th f t o ta tr becomes 935K, capabl f generatino e e supeth g r heated steam with 25MPa pressure th d an e temperatur f 810Ko e e outpu.Th f electricito t y becomes 100MW y installinb e g the steam power generating syste t thermam ne wite th hl efficienc f 40%o y .

Table I

COMPOSITION OF FUEL-SALT ASSUMED IN THE CALCULATIONS

Composition LiF BeF ThF PuF 4 3

mol* 71.75 16.2 0 12.0 0.25

185 Tabln e ISOTOPIC COMPOSITIO PLUTONIUF O N M ASSUMEE CALCULATIONTH N I D S

Isotopes 238pu 239pu 240pu 241pu 242pu weight % 1.5 55.0 25.3 13.2 5.0

e compositioTh s typicaR fuei nPW lr burned-ufo l o t p 33GWd/ton and left for 1 year before reprocessing.

Tabll ff e PRINCIPAL DESIGN PARAMETER FUJI-PF SO u

Thermal capacity 250 MWth Maximum neutron flux 13 2 Net electric generation 100 MWe Graphite(>50keV) 5.7xl0 n/cm s Thermal% 0 efficienc4 y Metal (>0.8MeV) 2.3x10" n/cm2s 3 Power density MW8 6. th/m Fuel salt Reactor vessel 239Pu concentration 0.25 mol% Diameter/Height 5.3/4.7 m ThF4 concentration 12.0 mol% Core zone Volume in reactor 7.5 m3 Maximum radius 1.5m 7 Total volume 12.0 m3 Graphite fraction 83 vol% Flow rate 0.55 m3/s Blanket-r zone Temperature In/out 835/935K Thickness 0.3m 3 Fuel conversion ratio 0.58-0.64 Graphite fraction vol0 8 % Inventory Blanket-z zone 239pu+241pu g k 5 27 Thickness 0.40 m Th 17.3 ton Graphite Fraction n 78vo Graphite 167 ton

3 Distributio2. f neutroo n n flux

e distributionTh e fasth t f o neutros e initian th flu n i xl core ar e showe therma d thosth n Figi nan f .o e2 l neutron flu n Figi x . .3 Comparing to FUJI-n, the core volume of FUJI-Pu was made larger to lower the neutron o increast flud an e xfue th el conversio ne reactor ratith e f coro oTh .e s dividewa d into thre ee fuel-salzonesth d an , t volume fractio- ad s wa n juste n eaci d h zont onl no eo lowet ye pea th rk e distributiovaluth n i e f o n the neutron flux, but also to increase the fissile production in the blanket. 10cm thick graphite plate containing O.lmolfc of boron was placed outside of the graphite reflector to lower the neutron radiation damage on the reactor vessel. The boron consumed by the irradiation of neutron flux during the plant life is about 20% of that contained in the initial graphite plate. As shown in the figures, the neutron flux is made rather uniform in the core. The peak values are 5.7xl013n/cm2/s (>50keV) in the

186 rt O CD " h- p t r 3 D C P n t X " 3 3 3 CD CD H- O rt n cn 3" O (T > n CD •n B H- -r) (-" o 0"? l-1' c 1 CD c P )-f i-j X T! rt CD 2 1— ' ET CD

Thermal neutron flux (n/cm /s) p CD rt Fast neutron flux (n/cm/s)

rt- 3 ET bo h — *—* ^ . 2 3" CO CD j_y P o o CD 3 000 O P o 'S O M i-" rt cn Cb ET cn rt Oo CD 3" 0 rt 1 rt ET (O •0 P CD ^ I-J o . CD [bution s 3 p CD reacto r n P I-» o- -f <-f r«- r-4- "O o CD O C II II II II C» rt —' O O CD p rt O> -3 prpp i O O cn Ms O CD O C*l O) O O> CD 3 a> toc> ncr cn rt O H P 3333 3" o 3" 0 O ^ CD o CD CO Mj "J O ^"i CD Mj O >-S ET CD *ü I-" rt CD CD • P Cn 3- rt •~J P 1 cn CD O P 3 P P >— 9 CD rt 1 0" B Ms O 3 cn CD n'- CD P 0 rt rt en CO H- 3- CD rt Q O "O cn *~t u. cn CD rt O P 3" 3 cn i-S >-< P gj rt CD CD CD P P C H. 1 B t— O-> t-^ rt CD cn CD P rt 3 3 O CD rt 0 i-" 3 ho i-S ET CD rt P * CD c CD j . rt o rt cn h-' •t i-i ^ O. E P O rt X CD 3 ET cn 3 CD rt O P M, P rt h-' P 3 H-9 CD C X n n Cn X CD D"

i.stanc e — '» p CD rt *? 3 3" O CD CD H H- B C cn rt OJ P ET CD rt rt *-( O X P 3- O H" Ms 3 CD 3 P 3 O 1 I— O CD CD O CD B C CD 3 Pb rt 3 CD h-** rt i-J >-* rt >t oo en ET O O CD era i CD 3 B HJ VH core, 2.3xlOun/cm2/s (>0.8MeV d 0.3xl0)an 12n/cm2/s (<0.18eV e inneth t r)a surfac f reactoo e r vessel. These value e withiar se exposurth n e limit above mentioned, so that replacement of the graphite moderator and the reactor vessel will be unnecessary during the plant life.

^ Col laps Ing

groua p

Figure 4 Procedure to calculate 1 group cross section data from JENDL-3.

3. Burn-up characteristics

Calculatio1 3. n method

Burn-up characteristics of the reactor were calculated by the procedure illustrated in Figs.4 and 5. Using RESENDD program^, 1900 group cross section data of 136 nuclei were first reconstructed from JENDL-3 and the cross sections were composed int 1 grouo p cross section data with 1900 group neutron spectrum obtaine y meanb d f celo s l calculation adaptine th g collision probability routine SRAth Cn i ecod e system s reportea , n i d reference [10]. Then, the fission and the capture cross sections of ac- tinides were replaced by those calculated by the SRAC code system to count the self shield effec y resonanb t t nuclei. With thes 1 groue p cross sec- tion data burn-ue ,th p calculations were carriey usinb t gou d ORIGEN-2[11^. e molten-salIth n t reactor e fuel-salth , s kepwa t t homogeneous it y b s circulation throug e coreth h , pump d heaan st exchangers e composith t Bu -. concentratioe th tiod an n f fissilo n e material d fissioan s n products always changee fuel-saltth n i d , because nucleath f o er reactio e cord th an en i n the additional fuel supply from outside of the plant. In order to simulate the continuous change of the composition and the concentration of the fuel- salt e burn-uth , p tims dividewa e d int a serieo f shoro s t intervals "day", e fissioanth d n reaction rate s calculatewa s y CITATIOb d r everfo N y zones of the reactor. Then, the irradiation and decay calculation were carried out by using ORIGEN-2 for every time intervals a-* defined as follow.

188 INPUT

SRAC USER LIB. 10? groups SRAC a •Collision 1 1 i l i b a b o Pr •CITATION

A-ef f I outpu o m r e tTh Reactioe ai n day zone

ORIGCN2 burn-up

Delà arrangement

'Conversion ratio/

Figure 5 Procedure to calculate burn-up characteristics by using ORIGEN-2.

dcor; = day x Wcore/W total

with cross section f Coro s e zone,

dbl_; = day x Wbl_r total with cross section f Blanket-o s r zone,

day = dbl_r*

= r WcorV + e

where Wcore is the thermal output of Core and Wbl.r is the thermal output of Blanket-r e irradiatioTh . d decaan n y calculation were performed with total e fuel-salth mas f o s t under total thermal outputÏ Ï , total • The calculations were carried out for time interval d.cor e * with cross sections of Core, for dbi-r* with cross section f Blanket-ro s , etc e calculatio.Th n procedurr fo e the time interval "day" was repeated for 900 days run time of the reactor. Burn-up calculations were r FUJI-Pcarriefo t ou ud runnino tw n i g operation modes, namely, Case- d Case-2an 1 . Case-1 simulate e conth s - tinuous reactor operation at rated power for 900 days without any chemical

189 . , U 1 . , 1 i i , i , , i , i i

o -o 104

D O» 103 ^_ _ ^ 0) : u P ^-i— , 0) 1 — „i ..-.——

Th f\ ——.-——-.-—— ,—— -r-, 102 : , i , i , , III, , 1,1,- c1 ( 6 3( 30 )0 9( Burnup ti me (days) Figure 6 Time-changes of plutonium and thorium fed in the reactor. Broken curves indicate the time-changes in Case-1, and solid curves the ones in Case-2.

processing. Case-2 simulates the reactor operation with uranium removal at every 300 days in batched way by applying simple fluorination and evapora- tion method afte 5 day6 r s fuel-salt cooling.

3.2 Feeding rate of Plutonium and Thorium

With the plant operation, the fissionable plutonium isotopes and thorium decrease, and the fission products accumulate in the core, causing lower neutron multiplication factor. Hence e fissilth , d fertilan e e materials should be fed to keep the steady reactor operation. The feeding rates of Pu and Th required for this purpose are depicted in Fig.6. The broken curves shoe feedinth w g rat e solin Case-th i e d an 1curve e oneth ss in Case-2. In Case-1, the feeding rate of Pu is about 2600g/day at the beginnin d decreasean g o about s t 400-500g/day after aroun 0 days5 dn I . Case-2, much amounu shoulP f e supplieo tb d n everi d0 day 30 yn orde i s o t r compensate fissil et abou a diminutio d t fe e core h th e needT b n .i o nt s 120g/day after aroun 0 day botn 30 di s h cases. e burn-ue fueTh th l f o salp t cause e neutroa changs th n i en flux dis- tribution. The profiles in Case-2 are depicted with two dotted chain curves in Figs.2 and 3. The profiles of the fast neutron flux after 900 days operation become lower than those of the initial core by 17% in the centra e lcore th zon ,f o ewhil profilee e thermath e th f o sl neutron flux are lower than those of the initial core by about 40% throughout the core.

190 3.3 Inventory of heavy metals

The changes of the fissile inventories in the reactor are shown in Fig.7. The fissile materials in the core are 239Pu and 241Pu at the beginning e reactooth f r operation. These isotope e consumear s d wite reactoth h r run, while 233s producei U d from protactiniu e reactoth n o i thame s r th t amount of 233U in Case-1 occupy 23% of total fissile in the core at around 0 day 90 s showa s n with broke nn e Case-2 figurei curvth t n Bu i e,. fis- sionable Pu is always dominant in the core, as illustrated with solid curve. Heavy metal inventorie e reactoth n i sr syste e listear m n Tabli d. IV e In Case-1, FUJI-P n transmutca u d producan e u 226kP e f o g112k f gfissilo e d 15kan minof go a P rd an actinide U 0 day90 sn i soperatio e t 100MWth a n f o

rated power. Similarlye e reactoth , n transmutca r d producan e u 242kP e f go 121kg of fissile U and Pa and 17kg of minor actinides in the Case-2.

1000

300 600 900 Burnup time (days) Figur 7 e Time-change fissile th f o se inventories. Broken curves indi- cate time-changeth e n Case-1i s d solian , d curve e one n th Case-2si s - To . tal means the total mass of fissiles (=233U+ 235U+ 233Pu+241Pu).

3.4 Fuel conversion ratio

The fuel conversion ratio, CR, is defined as the ratio of the fissile nuclei generate e reactoth n i dr syste o thost m e consume e fissioth d y b dan n capture reaction a uni n ti s time interval n Case-i R C 1. shown with broken

191 TablV I e FEED INVENTORIED SAN HEAVF SO Y METALS

Case-1 operation Case-2 operation Initial 0 day90 sFeed 0 day90 s Feed

233pa 0. 0. 5.0 kg -10.g k 8 4.6 kg 233 u 0. 0. 106.4 kg -73.0 kg 32.g k 2 235 u 0. 0. g k 3 0. 0. 0.3 kg total U 0. 0. 115.g k 2 -85.g k 1 37.3 kg

239pu 20g 5k 282 kg 246 kg 408 kg g k 5 35 2«pu g k 9 4 68 kgg k 4 11 98 kg 144 kg totau P l g k 3 37 513 kgg k 0 66 741 kg g k 3 87

Th n 17.to 4 g k 3 13 17.4 ton 128 kg 17.4 ton Am+Cm 0. 0. 15 kg 0. 17 kg curv e beginninn Fig.i e th e reacto s about i 8th a f 0 t- o g r in rund an , crease o about s t 0.6-0.64 afte 0 days20 r , becaus e productioth e 233f n o ni U the reactor is delayed owing to the decay time of protactinium, Pa. CR in Case-2 indicated with solid curve decreases sharpl t ever a y0 days30 y- be , cause CR does not count the fissile uranium isolated out of the reactor system.

0 60 0 0 30 900 Burnup time (days) Figure 8 Time-changes of the fuel conversion ratio, CR, and the plutonium conversion ratio . Broke% , n curves indicat e time-changeth e n si Case-1 d solian , d curve onee Case-2n th si s .

192 e samth e n I way e definw , e plutonium conversio e ratinth ratioos a , F , of integrated amoun f (o t233 235 u+ U+Pa) produce o integratet d d fissionablu P e consumed durin e reactoth g r operation o thas ,f countt e fissilth s e uranium removed out of the reactor system. The changes of ? are depicted in Fig.8. Thoug e (233th h 235 U+ U+Pa) productio s enhancei n n Case-2i d e th , fissile Pu consumption is increased at the same time, resulting almost the same values of 0.46-0.5 in the two cases.

3.5 Production rates of uranium and plutonium

The production rates of 233U, fissile U and Pa, 239Pu and total Pu isotope n Case-i se broke e showth ar 1 y nb n curve produc e n Fig.9th i s d -,an tion rates of those in Case-2 by the solid curves. The reactor operation with yearly isolation of U isotopes increases neutron economy and mini- mizes 233U burning, resulting higher addition and consumption of Pu and higher production of fissile U and Pa, as shown in the figure. e molteTh n salt reacto s comparewa r d withe LWR th n vie i f so w Plutoniu froe mse m consumption n Tablca , FUJI-Pe V e on n transmuts A ca u. e 919kg of Pu in Case-1 and 980kg in Case-2 annually for lGWe of the electricity generation. These amounts are about 50% larger than the ones consume LWRa y b d, verifying higher performanc f FUJI-Pplutoniue o e th s a u m burner. Using uranium oxide fuel discharge 45GWd/tot a d e burn-upth f e o n th , boiling water reactor BWR(U) can produce 230kg of plutonium annually for lGW generation. If charged with MOX fuel in place of uranium oxide fuel,

the e reactor BWR(MOX) is able to burn 649kg of plutonium. But, LWRs charged

200 1 • i • j • i • i • i • i • i •

100

0

•100 co: 0-200 -oo total Pu £-300

-400 0 300 600 900 Burnup time (days)

Figure 9 Time-changes of the fissile production rates. Broken curves indicate the time-changes in Case-1, and solid curves the ones in Case-2.

193 TablV e PRODUCTION RATES OF PLUTONIUM, URANIUM AND MINOR ACTINIDES

BWR(U) BÏÏR(MOX) FUJu I-P FUJI-Pu Case-1 Case-2

Thermal output 3000 MWth 3000 Mïïth 0 MW25 th 250 MWth Electrical output 1000 Mïïe 1000 MWe 100 Mff, MW 0 10 Net generation y - W 1G r pe totau P l 23g k 0 -649 kg -919 kg g k -980 Am+Cm 25 kg g k 3 13 62 kg 71 kg

239pu+241pu g k 2 14 -617 kg -990 kg -1062 kg 233u+235u+pa 0. 0. 453 kg 489 kg total fissile g k -659 g k -639 -537 kg g k -573

BWR(U) denote boiling-watee th s r reactor operated with uranium-oxide fuel and BWR(MOX) denotes the reactor operated with mixed-oxide fuel. The generation rates of plutonium iso- topes were calculated from the results of reference [12]. The discharge burn-u f fueo p l element BWR(Uf o s d BWR(MOX)an s )i 45GWd/ton.

with MOX fuel seem to be ineffective for plutonium disposition, because about 70% of Pu remains unused in the discharged fuels, and multiple recy- cling is necessary. Moreover, Pu causes harder neutron spectrum and may spoil neutron economy. Therefore e diminutioth , f plutoniuo n y LWRb m s might be expensive and wasteful of the energy resource. FUJI-P n consumca u e 990k fissilf producd go an u P ee 453k f fissilgo U e in Case-1, and consume 1062kg of fissile Pu and produce 489kg of fissile U i n e fissil Case-2th t rat ne f o o ethaes e , th consumptiot s aboui n t 537kg/GW eCase-/n yi d 573kg/GW1an eCase-2n /i y , respectively t ratne e e Th . e fissiloth f e consumptio f LWRo n s were 659kg/GW r BWR(U)e/fo y d an , 639kg/GWe/y for BWR(MOX). It can be said that FUJI-Pu can transmute large amount of plutonium and consumes little fissile comparing with LWRs run with the MOX fuel. FUJI-Pu run in two operation modes produce 62kg and f mino71ko g r actinides annuall r lGWfo ye generation. Thiy causma s a e problee incineratioth n i mradioactive th f o n e wastes, although more amount of minor actinide e producear s y BWR(MOX)b d .

4. Conclusions

A theoretical consideration has been made of the nuclear characteris- e smalth tic lf o smolten-sal t power reactor, FUJI-Pu whicn i , h plutonium recovered from spent-fuel e LWRs supposeth wa s e f usedb e o sreac Th o t d. - tor adopted yearly isolatio f uraniuo n m isotope o increast s u consumptioP e n and 233U production followine th d An .g conclusions were made. 1) The reactor transmutes 980kg of Pu annually for lGWe generation,

194 whic larges i h r thaR operateLW n a tha y b dt witX fuelMO h . t fissilne e eTh uraniu) 2 m produce reactoe th y 489ks b di r g annually r lGWfo e generation. 3} The fuel conversion ratio of the reactor is about 0.6, and the plutonium conversion ratio is about 0.46. e reactoe operateb Th n ) 4 ca r d without graphite replacemen0 3 r fo t years of the reactor life.

e authorTh s wis o exprest h s sincere . TsuyoshthankDr o t s i Misawd an a Mr. Masahiko Osaka of Nagoya university on their help and advise in cal- culating the nuclear cross section data by using the code "RESENDD".

REFERENCES

[I] ROTHENTAL.M.W.,HAUBENREICH,P.N.,BRIGGS,R.B., "The Development Status of Molten Salt Breeder Reactors", ORNL-4812 (1970). ] FURUKAWA.K..LECOCQ.A.,KATO,Y[2 MITACHId an . "Thoriu, ,K. m molten salt nuclear energy synergetics" Nuclea. J , r Scienc d Technologyan e , Vol.27, No.12 (1990) 1157-1178. LECOCQ.A.] [3 ,

[4] FURUKAWA.K.,MITACHI,K..CHIGRINOV,S.E.,KATO,Y.,LECOCQ.A. and ERBAY.L.B., "Rational Pu-disposition foU-productior233 n by THORIMS-NES (Thorium molten-salt nuclear energy synergetics)", This meeting (1994). ] FURUKAWA.K.,MINAMI.K..OSAWA.T..OHTA.M.,NAKAMURA,N.,MITACHI,K[5 KATOd an . , Y., "Design Study of Small Molten-Salt Fission Power Station suitable for coupling with Accelerator Salt Breeder", Emerging Nuclear Energy Systems (VELARDE.G.,Ed.) World Science (1987) 235-239. ] MITACHI,K.[6 , FURUKAWA,K.,MINAMI,K KATO.Y.d an . , "Nuclea burn-ud an r p characteristic a smal f o sl molten salt power reactor" Nuclea. ,J r Scienc Technologd an e Japanese)n (i y , Vol.32, No.4 (1990) 377-384. ] MITACHI,K.,FURUKAWA,K.,MURAYAMA,M[7 d SUZUKI,T.an . , "Nuclear character- istic a smal f o sl molten salt power reactor fueled with plutonium", Emerging Nuclear Energy Systems (YASUDA.H.,Ed), World Scientific (1994) 326-331. [8] TSUCHIHASHI,K.,ISHIGURO,Y.,KANEKO,K. and IDO ,M.,"Revised SRAC code system", JAERI-1302 (1986). ] NAKAGAWA.T.[9 prograA " , r reconstructiofo m f resonanco n e cross sections from evaluated nuclear data in the ENDF/B4 format", JAERI-M 84-192 (1984). [10] OSAKA,M.,MISAWA,T MITACHI,K.d .an , "Incineratio minof o n r actinidy b e usin moltega n salt reacto n witru rh 233U-thorium fuel",Proceedingf o s Fall Meeting of the Atomic Energy Society in Japan (in Japanese) Q-ll (Sapporo 1994). [II] CROFF.A. user'A " , s manua r ORIGEN-fo l 2 computer code",ORNL/TM-7175 (1980). [12] HIDA.K..ANDO.Y d YOSHIOKA.R.,"Transmutatioan . minof o n r actiniden i s BWR", Proc ANP'92f o . , Vol.1 (1992) 3.1-1.

Next page(s) Lef5 t 19 blank POSSIBILITE STUDTH F YO USINF YO G MOLTEN SALTR SFO PLUTONIUM UTILIZATIO ACTINIDD NAN E TRANSMUTATION

V.S. NAUMOV, A.V. BYCHKOV, O.V. SCIBA, P.T. PORODNOV Federal Scientific Centre Research Institute of Atomic Reactors, Dimitrovgrad, Russian Federation

Abstract

Propose concepe th s di usinf o t g molten salt plutoniur sfo m utilizatio actinidd nan e transmutatioe th n ni electronuclear power complexes. Described is the arrangement of fuel cycle with account to fuel composition, content of plutonium and total volume of fuel salt. The technological flowsheet provides for electrochemical reprocessing methods with electrolitic actinide release into metallic alloy and with their further return to the fuel salt.

Traditionally RIAR carrie researct sou developmend han t wor plutoni-un ko m utilizatio closen i d fuel cycl f FBReo . For 25 years the following milestones have been : - development of pyroelectrochemical technology for FBR MOX fuel production; to date about fueX ltonn3 havMO f so e been produced; - development of vibropacked fuel pins and its technology; currently the beine yar g irradiâ BOR-60n i t , BN-35 BN-60d 0an 0 reactors; - development and construction of remotely controlled facility for granulated fuel, vibropacked fuel pins and fuel assemblies production; - full-scale demonstratio fueu P lf utilizatioo n FBRn ni ; BOR-6e th 0 reacto s beerha n successfully fueabouX r operate fo yearsl2 MO 1 t n ;do - 26 % burn-up was reached with MOX fuel in the BOR-60; - one experiment was carried out on the pyrochemical reprocessing of spent FBR MOX fuel. Now, studu utilizatioP f yo Pu-burned an n FBRni R ,PW r mai f reactoro e n experimentaon s si l trende RIAth n Rsi activity. Plutonium can be used in modern type reactors as oxide fuel but plutonium recycle can be effectively realized only on the basis of pyroelectrochemical reprocessing and vibropacking technology.

For future nuclear systems it would be possible to use new principles related to Pu and other fissile materials recycle. One of proposed ways of implementing this could be to use molten salt fuel. The concept of using liquid fuel based on halide molten salts in nuclear reactors has been known since the 50s. The efficiency of using molten salts is determined by some of their specific properties and features: - ability to dissolve many inorganic compounds both in gaseous and in solid state; - enough inertness to the structural materials; - intensit processee th f yo s taking plac thosn ei e environment hige th h o t masvaluee e th sdu f o s

197 TABLE 1.1. HYPOTHETIC COMPOSITIO MOLTEF NO N FLUORIDE SOLVENTS f"îl Mixture eute:

1. LiF - BeF2 350 0 5080 0- (47 - 53)

2. LiF - BeF2 458 0 5080 0- (66 - 34)

3. NaF - BeF2 360 ) (544 6-

4. LiF - NaF 625 (60 - 40)

5. NaF - MgFj 824 (77,4-22,6)

6. LiF - ZrF« 51O 600 - 900 (50 - 50)

7. LiF - NaF - ZrF< 530 (24 - 41 -. 35)

8. LiF - BeF2 - ZrF< 355 (48 - 50 - 2)

9. LiF - NaF - KF 492 (46,5 - 42- 11,5)

10. Li FF - MgF Na ä- 63O (47 - 1 0 - 43)

11. LiF - MgF2 - NaF 684 (59 - 29 - 12)

transfer factors, thermal conductivity and heat removal ( when the molten substance is in contact with the molten salt). Among molten salts the fluorides and chlorides of metals are characterized by the high chemical and therma greatese l stabilitth o t e t heaydu t formation. face spitn I th t f ethao t individual fluorides have high melting temperatures their mixtures witha particular component compositio characterizee nar lowey db r melting temperatures r exampleFo . ,

LiF-BeF2 mixtures has an eutetic point at 350°C and LiF content of 47 mol%. Hypothetic compositions of molten fluoride solvents for fuel corresponding to eutetic compositions are presented in Table 1.1 [1.2]. i f theso e e us salt e s componenta s Th f fueo s l compositio s conditioneni e followinth y b d g advantages: -high radiation resistance permitting work with fue t burnu a MWd/kgl0 10 f po ; -high thermal resistance permitting reprocessing fuel with specific heat release exceedin MWd/kgg1 ;

198 TABLE 1.2. PHYSICAL PROPERTIE FLUORINF SO E SALT

53,5NaF-40ZrF4- 540 3,27 5,7 0,005 0,24 -6,5UF4 71LiF-16BeF2-12ThF4- 500 3,25 2,52 7,1 0,00287 0,32 -1,OUF4 73,7LiF-16BeF2~ 500 3,25 2,01 2,29 0,34 -10ThF4-0,3UF4 67LiF-30,5BeF2- 464 2, 10 1 , 90 5,5 0,57 -2,5UF4 52LiF-48BeF2 350 2,16 4,98 0,67 76LiF~14BeF2-10UF4 500 3,32 5,93 0,39 .Water (at 20 c) i/o 2,0 1,0 0,00143 1,0 3-0

8OOCC 2-5

§ 2-0

f. 5 700'C H. O

60O°C

O-5

i t i t i i i IO 20 30 4O 50 6O BeF mole %

/7G. 7.7. Solubility of PuFs in UF-BeF2 melts.

-possibility to create high concentration, of transuranium elements in melts because salts which are use workins da g environment t neutrono e sar n moderators. The physical properties of fluoride salts mixtures as liquid nuclear fuel are given in Table 1.2 (3). These properties of fluoride salts suggest that one of the ways for utilization of plutonium excess and long-lived actinides can be electronuclear power facilities using accelerator and blanket with circulating fue r molteo l n salt reactors (MSR). The classic projects of molten salt reactors (MSBR,MSRE) developed at Oak-Ridge National Laboratory demonstrated'a high solubility of uranium and thorium compounds in fuel salt.

2O 3O 40 50 mol% e FIG. 1.2. Solubility of PuFs in UF-TJiF4 melts.

200 PuF3 Concentratio Saln i t

Irradiation Time (years) FIG. 1.3. concentrationThe of PuFs moltenin salt UF-BeF2-ZrF4.

Most dat composition ao usagd nan fuef eo l saltdate base e th f phas saar o n do e diagram sucf so h systems which provide an estimated solubility value for these salts.

Figs.l.1-1.2 present data on solubility of PuF3 in LiF-BeF2 and LiF-ThF4 melts [2,4-6]. Therefore, as evident from the data presented for the MSCR and MSBR reactors the Pu content in the molten salt at 0.5 mol.% is quite possible. The same concentration of Pu in fuel salt is ABC/ATe th expecte t a e b Wo dt facility (see Fig. 1.3.) presene Ath t t tim maie eth n scientific prerequisite r problesfo plutoniuf mo m utilizatio molten ni n salt fuel has been formulated. 1. The investigations were performed to study the behaviour of uranium and plutonium compounds wels a soms a l e propertie f minor-actinideso maie th nd fissiosan n product thesn i s e environments. 2. A great amount of information has been generated sufficient for substantiation of pyroelectrochemica d pyrochemicaan l l fuel reprocessin r developmenfo d gan f differeno t t model thif so s cycle dependin initiae th finad n glan o l fuel state. 3.The fundamental possibilit r handlinfo y d reprocessinan g f molteo g n salt s fuebeeha l n demonstrated. foundatione 4Th . closef so d fuel cycle using molten salts have been developed. Studying the processes of plutonium utilization and actinide transmutation at MS R and also at the ABC/ATW facilities it is necessary to consider and evaluate the alternatives related to the arrangement of their fuel cycle due to the fact that the use of plutonium and minor-actinides propose that the effective burning is posssible with its multiple recycling for irradiation. well-knowe Th n flowshee MSBe th f o tR fuel cycl developes ewa t ORNLda .

201 Heat Fuel composition Exchanger correction

ABC/ATW BLANKET

Noble Gases P.P. Removal ~ - Sal ~ t - Salt Bi with F.B Noble and Seminob/e removal Freezing

Waste Preparation Disposal _ _ _ Fuel Salt Without Actinides Pure Salt Salt With Actinides

FIG. 1.4 e consideratioTh f fueo n l cycle option r sucfo s h facilit s performewa y d with accoun o fuet t l composition, conten plutoniuf o t totad man l volum fuef eo l salt. contrasn I fueo t l cycle MSBf so MSRd Ran E projects which involve flowsheete dth s with complete fuel composition reprocessing and a great number of technological operations, the flowsheet of the actinide transmutation facility can be rather simplified (Fig. 1.4.). The gas removal system, for example, may be similar to that developed in the MSBR project. Moreover, the fuel salt flux should pass through "noblee th releasP "F e circuit thin I . s cas complete eth e reprocessin possibls gi carriee b t o et dou accordin flowsheee th o gt t allowin pyroelectrochemicae th r gfo l reprocessing. These methods sugges firse t th that n stago t e electrolytieth c depositio l actinideal f no s inte oth metallic liquid alloy is performed. After that, the salt from the first apparatus (electrolyzer) is delivered secone th o t d apparatus intende rarr dfo e earth elements removal secone th n dO . stag sale eth t comes firse bacth t o apparatukt s wher electrochemicae th e l extractio actinidef no s into fuel salt takes place. In this case the more electropositive elements (for this system such elements as Zr, Mo,Fe etc.) will be accumulated in the alloy. After reprocessing the fuel salt again returns into the actinide burning facility circuit. The metallic allo subjecs yi furtheo t t r reprocessin molten gi n sal havo t tfissioe eth n product compacn si t fors ma oxide e fordeposit th f metallim o n i r o s c alloy with further disposal. Such technology will make it possible to maintain a particular content of fission products and rare earths that will depend on the molten fuel volume subjected to reprocessing. expecterange e th Th f eo d separation factofrou P mf o r rare earth 20-5s salse i thad th t f 0froan o t m impurities is 100-500. From the technical standpoint the flowsheet of this fuel cycle option has the most similar efficiency with the one developed on the basis of the ORNL data. Its application can lead to sufficient fuel purification fro impuritiee mth s wit simple hth e equipment used. Therefore , e possibilitstudth f yo f usino y g molten salt r plutoniufo s m utilizatio d actinidan n e transmufation shows thafuee f molteth tl o cycle us n e e saltbase th fue n si n do l reprocessing can organically be integrated into the structure of nuclear power complex and will allow tackling the proble f higmo h toxic radioactiv wayw e wastene . a Nevertheless n si havo t , e these problems solved, further work in the field of fundamental research and development of separate methods for fuel handlin necessarye ar g . opinior Onou practicae nth l implementatio sucf no h program projectd san s will greatly depend on the efficiency of the international collaboration on this problem.

This wor performes kwa supportey db U.Se th .f do LAN L under contract 5058L0013-9Y.

REFERENCES

1. Cnpaso^KH o nuasKocTn K H COJICBHX CHCTC M/ Iloj a peÄ.H.K.BocKpeceHCKO Ä- M.-JI.: HS-BO AH CCCP, 1961.: .l-845c.;x.2-585c.

203 2. /{HarpaMMH roiaBKocxH COJICBBIX CHCTCM. TpoiteHe CHCTCMH /CnpaBomMK./ HOÄ B.H.IÏocHnaôKo, E.A.AneKceeBofit -M: XHMHJI, 1977, -328c. 3. BJIHHKHH BJL, HOBHKOB B.M. ^KnoKocoJieBHe sy^epHHe peaicropH.-M.: ATOMHaflax, 1978. -H2c. . ^[HarpaMM4 H ruiaBKocTH COJICBBTX CHCTBM:. MHoroKOMnoHeHTHtie CHCTCMH. /CnpaBOHHiocR /HÖ PCÄ- B.H.IIocbmafiKo, E.A.AJieKceeBoft -M.:XHMH«, 1977, -216c. 5. V.N.Vaidya, R.Prasad, V.Venugopal, P.N.Iyer e.a. Molten Salt Chemistry. Part IV. Solubility behaviou f PuFo r fluoridn 3i e salt interesf so molten i t n salt reactor technology.- BARC -676, (1973), Bombey, India. 6. C.J.Barton. Slubilit f plutoniuyo m- trifluorid fusen i e d alkali fluoride beryllium fluoride mixtures. . PhysJ - . Chem.,1960 ,, p.306-312 v.64N3 , .

204 Next page(s) Left blank SESSION 5: FAST REACTORS STUDY ON Pu BURNER FAST REACTOR CORES WITHOUT URANIUM

M. ISHIKAWA, A. SHONO, T. WAKABAYASHI O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Ibaraki, Japan

Abstract

In Japan, extensive stud f fasyo t reactor cores, whic efficientln hca y burn plutonium and/or minor actinides bees ha , n continue ordedn i demonstrato t r e eth flexibilit actinide th f yo e utilizatio fasn i t reactors. Recently, concerns ovee th r excessive plutonium have been significantly increased fro viewpoine mth f o t nuclear proliferation. Fast reactor cores without uranium are considered as one of the most effective solutions to reduce the plutonium stock. The present paper summarizes nuclear characteristic e fasth t f o sreacto r cores fueled with PuO2/ceramics, instead of PuO2/UO2, including their burnup performance of plutonium, reactivity coefficients and some problems from the safety viewpoint. As a result of these surveys, such unconventional cores without uranium are found to consum e plutoniuth e t mvera y large burnup rate e theoreticallclosth o t e y maximum valu f 110-12eo 0 kg/TWhe. Further, sodium void reactivit none th -n yi uranium cores decreases extremely, whic preferabls hi safeto et y consideration. otheOe nth r hand, some disadvantage substantiato st ereactoa r desig none th -f no uranium core woul large db e burnup reactivity lossmald san l Doppler effect.

. Introductio1 n

Recent global relaxation of the demand for primary energy sources leads to the serious dela f fasyo t breeder reactor developmen constructiond an t havw no e e W . expeco t larga t e amoun f overbalanco t excessivd ean e stoc f plutoniuko m from light water reactor (LWR) operation Japann I . , ther fire mear policies!e th r 1fo ) atomic energy usage: (a) to restrict Japanese nuclear research, development, and utilization exclusivel peacefue th r yfo l purpose recyclo t ) (b , e nuclear fuel, o t tha , is t reproces spene sth t fuel, recycl recoveree eth d plutoniu uraniumd e man us d an , the nucleas ma r fuel repeatedly accepo t ) IAEA'e (c , th t s safeguards agreement under the Non-Proliferation Treaty (NPT), and, as a consequence of above- mentioned fundamental lines, (d) not to reprocess and stockpile more plutonium than is needed to feed reactors as fuel, since we understand that there are some concerns over the proliferation of nuclear weapons with the dispersion of reactor grade plutonium. The emphasis of Japanese policy for the plutonium utilization is to treat plutonium as a valuable energy resource, not as useless nuclear waste. Due to their good neutronics characteristics, fast reactors intrinsically have good flexibility of plutonium consumption as well as breeding. In the course of these backgrounds, we are now investigating an efficient way to control the plutonium stock using fast reactor system. As an international cooperation related to the efficient plutonium

207 (Conventional MOX Core)

Elimination of Bfanket 'ide Gap between Pellet and Clad| 4- Increase of Pa Enrichment Larqe Annular Hol Pelleen i t

Spatious Fuel Pin Pilch

Decrease of Fuel Volume Ratio Scattered Emptn yPi

Enhancement of Neutron Leakage Reductio f Corno e Hiqht

Inclusion of Neutron Absorber Small Core Size

Introduction of B^C in Fuel

(Unconventional "Non-Uranium" Core)

Complete Exclusion of Pu Source from Core Reqion

Fig.1 Approach to Efficient Pu Burning in Fast Reacor Cores

burning CAPRe th , A project^ alss i ) o ongoin asseso gt feasibilite sth o t d yan develop the reality. The present paper describes recent result plutoniur ou f so m burning stud fasyn i t reactor cores.

2. Approach to Efficient Burning of Plutonium

There are general approaches to suppress the generation of Pu and to burn Pu efficiently in conventional MOX-type fast reactor cores, as shown in Fig.1. First of s naturai all t i , l thablankee th t t region whose objectiv breedinu P s ei g from depleted uranium, mus eliminatee b t d fro coree mth . corNexte th en i ,fue l region, the Pu enrichment of the core fuel should be increased as much as possible, in orde mako rt amoune eth f U23to 8 minimu tha mo conversioe s tth n cor e ratith e on i region decreases. The concrete methods to increase the Pu enrichment may be: (a) to obtain small volume rati f fue oo employiny b l g innovative fuel design suc wid s hbetweep a ega n fuel pellet and clad, large annular pellet hole, spacious fuel-pin pitch, scattered empty pin, etc., (b) to enhance neutron leakage from the core region by the adoptio shortef no r core height, smaller core sizeintroduco t ) , (c etc. d e,an neutron absorber material such as 840 to the fuel. straightforwara s Ii t d consideratio extensivele w f ni y evolve above-mentioned methods for the conventional MOX burner cores, that the ideal solution to burn Pu

208 most efficientl completelo t s yi y exclud generatine eth g sourc, i.e.Pu ,f eo uranium , fro coree m th thin I .s "non-uranium burninexpec"u n caseP ca e e th gtw , rate close to the theoretically limited consumption of 110-120 kg/TWhe, because almost all thermal energy generated in the core arises only from the fission reaction of Pu. As actual fuel material of non-uranium idea, however, we cannot apply pure PuOa fuel, sinc core eth e size smalo woulto e ld b and/o fuee configuration th rpi l n would be irrational from the viewpoint of power generating core. We adopted here a dilution method where PuO2 is dispersed with some non-fuel material such as ceramic construco st t reasonable configuratio coree th f n.o concerne Th f thiso s unconventional approac buro ) ht woulu (a nP : dbe performanc core poweth s ef a e o r generation plant, especially, lengt f continuouho s operating period ) controllabilit(b , e cord safetth an y ef o ywit h unfamiliar characteristics such as small Doppler effect, (c) soundness of fuel during irradiation, ) suitabilit(d d an fueo yt l manufacturing, reprocessin fuee th l cyclen gi . Thus, there are many physica industriad an l l aspect investigatee b o st non-uraniue th n di m core option. As the first step of feasibility study, we here concentrate on the neutronics characteristics of the core.

3. Characteristics of Reference "Non-Uranium" Core

In this chapter, nuclear characteristics of reference cores without uranium are described wit desige hth n parameter calculationae th d san l method. 3.1 Reference Core

We tak epowea r generatio electriW M n0 c cor 80 (208 thermalf eo W e 0M th s a ) reference casemaie Th n. design parameter cor e specifiee th ear f so Tabln di , e1 and the configuration and dimensions are shown in Fig.2.

Tabl e1 Design Parameter f Referencso e "Non-Uranium" Core

Item Parameters

Reactor Power SOOMWe (2,080MWt)

Operation Period 182.5 days/cycle

Number of Refueling 6 batches

Pu238/Pu239/Pu240/Pu241/Pu242/Am241 Pu Isotopic Composition 1 . 1 / 5 5. / 58. / 8 =1 1. 2. / 22.1 1 3 /

Core Concept Two-Region Homogeneous Core

Core Height 60 cm

Core Diameter (Outer / Inner) m c 8 5236 3/

Fuel / Coolant / Structure / Void Volume Ratio of Core Region = 23.4 / *57. 3 / 14.7 4. 6 /

* includes Ceramics Diluent

209 R Axial Shield R a d d a o, ' » 1 Inner Core ?|i Ml LÎi J 0 t1 < * t d Axial Shield d

3680 l 5230 1 - 5830 (Diameter (mm))

O Inner Core 324

) (f Outer Core 326

® 3 S Shield 306

© Primary Control Rod 29

® Buckup Control Rod 12

Total 997

Reference MW 0 80 e "Non-Uranium 2 g Fi " Core

The reference cortwo-regioa s ei n homogeneou highm c ,0 s whic6 typd ehan s i a kind of pancake shape to enhance neutron leakage from the core. On the other hand core th ,e diamete quits i r keeo t e , plargem half-a-yea2 5. , r continuous operation period which woul needee db d fro econome mth y viewpoin powea f o t r generation plante numbeTh .f refuelino r g batche s sixi s , hence each fuel subassembly is irradiated in the core for three years. The isotopic ratio of the Pu fuel, 69% of Pu fissile contents, is assumed as spent fuel of some typical LWR. Some kind f ceramicso s suc Al2Oss ha r CeC-, o BeO selectee O 2ar ,Mg diluens da t of PuOa fuee Th l .volum e rati 23%s oi , whic quits hi compare w elo d with conventional MOX-type breeder cores. Note tha increasee w t coolane dth t volume ratio correspondin decrease th o gt f fueeo l volum ereference ratith n oi e case, however, there would be some alternatives such as increase of the structure material or diluent.

3.2 Analytical Method

The nuclear data used ABBN-typn hera s ei e 70-group constan r fasfo tt se t reactors based on Japanese Evaluated Nuclear Data Library (JENDL) version 2(3).

Concerning with nuclear data dependency of the analysis, we also applied the latest version of nuclear data, JENDL-3.2(),4 and found no visible difference in the results burnue Th . p composition chang burnue th f fued eo p an l reactivity lose sar obtaine two-dimensionay db l diffusion calculation with collapsed seven energy groups. The reactivities of sodium void and Doppler are evaluated by 70-group calculatio conditioe th o nt n with zero sodium densit r isothermallyo y increased

210 temperatur core th en ei region , respectively componente Th . f thesso e reactivities are estimate diffusioe th y db n perturbation theory.

3 Characteristic3. f Referenco s e Core

The neutron spectrum of a reference non-uranium core with AlaOs ceramics is compared with bot conventionaha l fast breede MOX-typra cord ean burneu eP r core in Fig.3. The spectrum of the non-uranium core is greatly softened from these conventional MOX cores because of enhanced slowing down by lighter elements existing in non-fuel materials. The extreme change of the spectrum is expected to affect reactivity characteristics. Some important core characteristics of the non-uranium cores are summarized in Table 2, which also includes those of a MOX-type Pu burner core with 30% of Pu enrichment for comparison.

(1) The Pu consumption rates of the non-uranium cores are almost double of the MOX-Typ kg/TWh0 value 11 burneu th f eP ed o closs ei an r theoreticao et l maximum as expected. (2) Due to lack of Pu conversion, the burnup reactivity loss is extremely increased to 1.1-1.2 %dk/kk'/month. If a longer operating period than half-a-year is required, havy w consideeo ma et r some remedie controe th number d sfo ro l r otheo r r design parameters.

10 io'~ -

10 i o'

0 1 Flux (Arbitrary unit) 10 .

-7 10 L

10 .

10 .

10 . -n: 10 .

10 10" 10 10' 10" 10 10' Neutorn Energy (eV) Fig .3 Compariso f Neutrono n Spectrum

211 Table 2 Characteristics of Reference "Non-Uranium" Cores

Fuel Type MOX burner Pu02/BeO PuO2/ AI2O3 PuO2/ MgO PuO2/CeO2 Pu Content in Fuel 28.5/31.5 25.9/25 6 269/273 26.8/27 1 26 9/27.6 {in/out, (v/o)) Pu Consumption 55 110 111 110 110 (kg/TWhe) Puf Consumption 58 118 116 116 114 (kg/TWhe)

Conversion Ratio* 0.51 0.15 0 12 0.12 0.11

U ReactivitB. y Loss 0.49 1 20 1.13 1.13 1.10 (%Sk/kk'/month) Na Void Reactivity* 2.14 -0.66 -0.95 -0.98 -1.03 (%5k/kk')

KDoppler* -6.6 -4.7 -2.8 -2.9 -2.4 (XlO-3Tdk/dT) Reactor Power 2080MWth. Reactor Size=06mHX5 2mD Pu Vector (Pu238/Pu239/Pu240/Pu241/Pu242/Am24 , (w/o))1 = (180/5820/2230/11 10/550/) 110 Volume Fraction / 6 Structure)=2 (Fuea 4 1 N / l/ 3 7 5 / 34 * at Beginning o< Equilibrium Cycle

sodiue (3)Th m void reactivity shows marked decrease fro conventionae mth X MO l core and reaches even negative values. This is quite favorable from the safety poin viewf mechanise o t Th . greae th mf to decreas understooe b n eca d froe mth component analysis show Fig.4n ni dominane Th . t reductio f sodiuno m void reactivity was achieved by the decrease of scattering term, which came from the spectral softening as mentioned above. We also found the effect of neutron leakage rather insignificant for the sodium void reactivity change.

Fission 0 MOX H Pu02/BeO

D Pu02/AI203

Capture D Pu02/Ce02

Scattering

r Y//// ///////////77f Leakage

'///////////////A Total

-3-2-101 234 a VoiN d Reactivit k/kk'S % y( )

Fig. 4 Components of Sodium Void Reactivity

212 (4) One of the most concerns for the non-uranium cores would be the large decrease of Doppler reactivity absolute values, since the missing U238 is the dominant source Doppleth f o e r contributio n conventionai n coresX MO l . However, the Doppler coefficients of the non-uranium cores still keep negative. The contributio f individuao n l nuciide s clarifiei s n Fig.5i d y whicb , e w h understand the negative Doppler reactivity originates mainly from two nuciides, tha , Pu24is t verironneed e e b 0 yan W . do t carefu resultse th r fo l . Pu24e on 0s i fuee oth f l isotopes initiae th t l amounbu , tand/o dependR kine e LW rth f th d o n so burnup degre spene th f eo t fuel from originates u whic P isotopie e th hth d can , composition also largely changes in time during the reactor operation. On the other hand, iron is included in structure material, therefore, the iron contribution to the Doppler reactivity in actual reactor operation will be quite smaller than the isothermal conditio o bott e h ndu smalle r temperature chang thermad ean l response delay. We are now studying the effect of the small Doppler reactivity to the safety characteristics by transient analysis. (5) We should point out some other concerns for the non-uranium cores. One is the small value of effective delayed-neutron fraction, approximately two-thirds of conventional MOX cores, which comes from the difference of delayed-neutron yield between plutonium and uranium. Although we could manage this in the plant control system design phase, the verification by dynamic calculation wouid be needed othee Th .r concern relate e powee sar th o dt r distribution swinn gi the core region and the power mismatch among different burnup-experienced subassemblies by many refueling batches. In the conventional MOX cores,

Nuclide Na Cr Fe Ni Mo Pu239 Pu240 Pu241 Pu242

-0 -0.4 -0.3 -0.2 -0.1 0 0.1 Contributio o Dopplent r Reativity

Fig 5 . Contributio o Dopplent r Reactivity O Core I (Pu+A O) 223

213 Tabl e 3 IntroductioC 4 EffecB f o t Corn i e Region

(Fuel Type: PuO2/AI2O3)

B4C Volume Ratio (%)* 0 10 Pu Conten Fuen i t l 18.4/19.0 28.2/30.4 (in/ouU (v/o)) Pu Consumption 108 111 (kg/TWhe) Puf Consumption 114 111 (kg/TWhe)

Conversion Ratio** 0.14 0.09 B.U. Reactivity Loss 0.86 0.44 (%Sk/kk'/month) Na Void Reactivity** -0.76 1.48 (% S k/kk1) \f ** ^Doppler -3.4 -0.17 (XlO-3Tdk/dT) Volume Fraction / Structure(Fuea N l/ ) =41.6 / 37. 5 / 20. 9 Core configuration is slightly different from the reference core. * to Fuel Volume ** at End of Equilibrium Cycle these power change tim n si alleviatee e ar conversio e th y db . f U23no Pu o 8t We expect, however, this power drift could be handled in the optimizing effort of the design, because the Pu burner cores have low power density in nature so as to reduce the burnup reactivity loss and to keep a sufficient operating period.

Tabl e2 4 Introductio EffecUO f o t n Homogeneousl Corn yi e Region

Fuel Type PuO2/ BeO PuO2/CeO2 Pu Consumption 17% Decrease 12% Decrease (kg/TWhe) Puf Consumption 14% Decrease 11% Decrease (kg/TWhe)

Conversion Ratio* 0.15=>0.24 0.11 =»0.19

B.U. Reactivity Loss 9% Decrease 9% Decrease (%£k/kk'/month)

Na Void Reactivity* 0.6~0.7 9 0. — 8 0. (%5k/kk') Increase Increase

\s * ^Doppler 22—25% 11-30% (X10-3Tdk/dT) Increase Increase

f Fueo 2 o amounl v/ VolumUO 0 1 ; t e U/Pu=0.3 Volume Fraction (Fuel / Na / Structure) =23.4/57.3/14.6 * at Beginning of Equilibrium Cycle

214 . Improvemen4 f Reactivito t y Coefficients

Although the non-uranium cores possess ideal performance of Pu burning rate, have w e found some disadvantage take w e f si the powes ma r generation sources. Some option improvo st core eth e characteristic surveyee sar d here.

4.1 Introduction of Burnable Poison

In order to reduce the large burnup reactivity loss, 10% fuel volume of 640 was introduced homogeneousl core th en y i regio burnabla s na e poison effecte Th .n so core characteristic summarizee sar burnue Tabln dTh i . pe3 reactivity loss swa decreased largely, almos reference th hala tf o f e case. However sodiue th , m void reactivit increases ywa approximately db y 2%dk/kk' Dopplee th d an r, reactivity came close to zero. These changes are caused by the spectral hardening due to neutron absorptio borony nb t seemI . s reasonabl discare ew d this option froe mth safety aspect.

4.2 Re-introduction of UO2 Homogeneously in Core Region

Apparently disadvantagee th , reference th f so e cores arise fro lace mf th ko uranium itself. There migh possibilita e tb y tha smala t l recoveramoun2 UO f e to s th unfavorable characteristics of the non-uranium core without serious influence on Pu burning performance. This is a kind of trade-off problem. We first selected the introductio homogeneousl2 UO f no core th e n yi region . Tabl showe4 effece sth r fo t dilueno tw e tth cases, wherfuef o amounladdee 2 % volumeeth 10 UO ds f i s o t A . negative changes, the Pu consumption rate decreased by 12-17%, and sodium void reactivity increased 0.6-0.9%dk/kk'. On the other hand, the burnup reactivity loss improved by 9%, and the Doppler 11-30%. We judge the merit of this option is attractivo s t no e compared wit sacrifice hth f burneeo r performance.

3 Adoptio 4. 2 InternaUO f no l Blanket Axiall Corn i y e Region

In the conventional MOX-type FBR design, axially heterogeneous core is considered as one of the most promising concept to enhance the core performance. We considere similae dth r idea coul burneu adoptee P d b e rth casen di , because internae th l blankeexpectee b n ca t affecdo t reactivite th t y characteristics efficiently by the placement in the highest importance region, while Pu burning is the characteristics of the whole core summation. Tabl comparee5 core sth e characteristic internae th f so l blanket case with those of a reference non-uranium core and a MOX-type Pu burner core. Although the Pu consumptio ne interna th rat f eo l blanket cas s decreasei e fro% e mth 15 y db reference non-uranium core, the improvement of burnup reactivity is 24%, which is far larger than the sacrifice of burning performance. In addition, increase of the sodium void reactivit s onli y y 0.5%dk/kk', whic s muci h h smaller thae th n homogeneous UO2 case. A marked improvemen f thio t s option appear isothermae th n si l Doppler reactivity, which recovered by 87% and even close to that of the conventional MOX- type FBR cores. Although we have to consider the degradation of the Doppler

215 Tabl e 5 2 InternaEffecUO f o t ! Blanket Axiall Corn yi e Region Axial Internal Blanket*" Fuel Type MOX Burner* PuO2 / AI2O3** in Pu02/AI2O3** Pu Content in Fuel 28.5/31.5 15.7/15.6 18.5/17.5 (in/ouU (v/o)) Pu Consumption 55 111 94(-15%)# (kg/TWhe) Puf Consumption 58 119 102 (kg/TWhe)

Conversion Ratio**** 0.51 0.13 0.23

B.U. Reactivity Loss 0.49 1.13 0.86 (-24%)# (%£k/kk'/month) Na Void Reactivity**** 2.14 -0.09 0.44(+0.53%5k/kk')# ( %k/kk'S )

^Dopple!/• r ** * * -6.6 -3.8 -7.1 (+87%)* (X10-3Tdk/dT) Volume Fraction (Fuel/Na/Structure) = * 23.4/57.3/14.6, " 41.6/37.5/20.9 *** Thickness of Axial Internal Blanket ( Inner / Outer Core) = 7.4 / 3.8 cm *** t Beginnina * f Equilibriugo m Cycle * Difference from Reference No-Uranium Core

reactivit actuae th n yi l reactor operation becaus f temperatureo e distributione th , adoptio f internano u burnelP blankee rth coren i t s seems worth pursuing furthermore. Other advantag f thieo s concep appeay alleviatioe th ma t s powee a r th f no r swing in the core region, because we could optimize the thickness of the internal blankets between inner and outer core regions independently. Finally surveyee ,w validite dth thif yo s axially heterogeneous concep variour fo t s reactor sizes, core height, and fuel volume ratio. Figure 6 shows the results related to sodium void reactivity and Doppler reactivity. The internal blanket was found to play an important role for the improvement of reactivity characteristics in various types of non-uranium fueled core.

5. Concluding Remarks

presene Th t burninstudu P f y o unconventiona n gi l fast reactor cores without uranium can be summarized as follows: consumptiou P e Th - n non-uraniue rateth f so m core approximatele sar 0 y11 kg/TWhe, which is close to theoretical maximum. othee Th - r advantag non-uraniue th f eo m core extremels si sodiuw ylo m void reactivity. - There are some concerns with the large burnup reactivity loss and the small Doppler effect. Although, these disadvantages withiseee b range mo nt th e settle design-wore th n di k elaboration favorabls i t i , improvo et reactivite eth y characteristics by some creative idea. possibilite On - improvo yt core eth e characteristics adoptiowoule th e e db th f no

216 X MOX(30w/o ennch)800MWe,60H (Vf/Vc/Vs=0.23/0.57/0.0 I /A O Pu O 1 5)800MWe,60H

O Pu02/Al203(Vf/Vc/Vs=0.42/0.38/0.2 1 ))800MWe,60H

• Pu02/AI203+ln. Bk (ic.oc dif)(Vf/Vc/Vs=0.42/0.38/0.21 ))800MWe,60H • PuO /AI 0 +ln. Bk (ic,oc dif)(Vf/Vc/Vs=0.42/0.38/0.21))600MWe,75H

4 Pu02/AI203+In (ic.ok B . c dif)(Vf/Vc/Vs=0.42/0.38/0.21 ))300MWe,60H

T Pu02/AI203+ln. Bk (ic.oc dif)(Vf/Vc/Vs=0.42/0.38/0.21 ))300MWe,75H

2.S i .... i .... i .... i .... i ....

«0 2 X . j 1.5 X 1 _ Effect of U0 Internal Blanket '_ 0.5 H 2 o eCOu 0 -0.5 "O "o -1 i ... i ... .i ... i . .... i ..... - 2 8 -7 -6 -5 -4 -3 -2 t /N~J"T~_Jf / / _IT""| f \\ Doppler Fig. 6 Effect of Axial UO Internal Blanket (Fuel Type: PuO /AI O ) 2 223

internal blanket axiallnon-uraniue th n yi m coreadvantage Th .reactivit e th r efo y coefficient exceey s ma degradatio e dth burninu P f no g performance.

Finally, general R&D themes to substantiate the efficient Pu burner concept would be considered as: - Improvemen d optimizatioan t e corth ef o nconcep t relate o reactivitdt y coefficients, power swing, reactor core size and operation length per cycle, - Verification of safety characteristics including transient condition, - Accumulation of material property, especially, of fuel with diluent, - Compatibility with fuel cycle such as fuel fabrication and reprocessing, - and, investigatio f reactono r physic criticay sb l experimen analysisd an t .

REFERENCES

[1] Y. IHARA: "Future Nuclear Systems: Japanese Fuel Cycle Strategy," Proc. Int. Conf. on Future Nuclear Systems: Emerging Fuel Cycles & Waste Disposal Options (GLOBAL'93), Seattle (Sep. 1993) [2. ROUAULTJ ] t al.e , : "Physic f Plutoniuso m Burnin Fasgn i t Reactors: Impacn o t Burner Cores Design," Topical Mtg. of Advances in Reactor Physics, Knoxville (Apr. 1994) [3] T. NAKAGAWA: "Summary of JENDL-2 General Purpose File," JAERI-M 84-103, Japan Atomic Energy Institute (1984) [4] T. NAKAGAWA: "JENDL-3 Revision 2," JAERI-M 94-019, Proc. 1993 Symposium on Nuclear Data, p.68, Japan Atomic Energy Institute (Feb. 1994)

Next page(s) Lef7 t 21 blank Pu BURNIN FASGN I T REACTOR CORES USING UNCONVENTIONAL FUEL WITHOUT U-238

G.G. BAIBURIN ARSRHM, Moscow

A.P. IVANOV, LYu. KRIVTTSKI, V.I. MATVEEV, E.V. MATVEEVA SSC IPPE, Obninsk

Russian Federation

Abstract f faso te reactorUs r actinidfo s e burning could serv s rathea e r effective optio f commono n problem solutio f nucleano r power long-lived waste activity reduction. Analysi f faso st reactor possibilitie r sucfo s h objectives showe th s expediency of use of specialized cores ( or specialized fast reactors), fuel of which contains actinides withou U t replaceoo j? inery db t matrix. Concep sucf o t ha core conformabl BN-80e th o yt 0 reactor typ consideres ei papere th n di . Result f calculationo s f basio s c physics parameter o reactotw f o rs core options are given: for plutonium burning and for minor actinides burning. Means o improvt e power distribution flattenin e shownar g . Assessmen f reactivito t y effect e givenar s . Issue f safeto s y provisio r corenfo s under consideratioe nar discussed.

1. INTRODUCTION

Fast reactors can be used efficiently for burning long-lived actinides contained in the spent fuel of thermal reactors which are working within the open fuel cycle. Plutonium is one of the most dangerous elements in the spent fuel actinide group from ecological poin viewf o t . Utilizatio plutoniumf no , whics hha been accumulate s stili ld beindan g accumulate P (nuclea resula NP s df a o t r power plant) operation is one of the main tasks within the frame of general problem of activity decrease of the long-lived wastes of the nuclear power. Along with the NPP plutonium, the use of military plutonium can be considered, although it has of course some specific character. Another side of this problem is necessite th f burninyo g minor actinides: americium, neptunium, curium, which will determine nuclear power wastes activity long time afte e plutoniuth r s mha been extracted. Actinide burnin possibls gi coree th existin f sn o ei g LMFBR's (Liquid Metal Fast Breeder Reactors), although it would be advisable to make some changes in the core design (e.g., fertile blanket elimination, fuel enrichment increase, etc.) in order to improve the efficiency of the process. However, actinide burning can be carried out more perfectly in the fast reactor core using new fuel type in which U T"2Q is replaced by inert matrix. The expediency of U"?^ eliminatioA n is obvious

219 becaus s actiniai t ei e supplier under irradiation maie Th .n result f suco s h core physics study made conformabl BN-80e th o t y0 reactor desig e presentenar d below.

. 2 FUEL CORE CONCEP BASID TAN C DAT CALCULATIONSR AFO .

At present various fuel compositions without U 0*3Q on the basis of MgO,

otherd A1an considerede s2N ar O 3A1 , productioe Th . fuef no l pellet s- fue l pins- from this material bases si technologn do powdef yo r metallurgy. Her briefle ew y consider another approach to the obtaining of such compositions, based on utilizatio porouf no s materials, saturate fissily db e isotopes considen ca e W . r ZrC, C, Ti and others as such materials. At present the most studied material is composed material on the basis of zirconium carbide with porosity from 50 to 80 %, saturated by plutonium oxides. This material was developed by All-Russia Scientific Research Institute of Inorganic Materials named after Academician A.A.Bochvar (ARSRIIM). supposes i t I d that technolog introductiof yo f otheno r actinide oxides (Np, Am, Cm) will not bring principal difficulties. e expectedb n Ica t , that porou frameworC Zr s k weakly interacting with neutrons, provides fuel column integrity under high burn-up of fissile isotopes. e technologTh f actinidyo e oxides introductio bases ni n utilizatioo d e th f no following compositions: colloi- d solutio plutoniue th f no m dioxide; - metal-organic composition. The methods of obtaining of colloïde oxide solution are sufficiently develope used dZol-Gean n di l processe f granulateso (MixeX dMO d Oxide) fuel production. The technology of production of metal-organic composition, containing plutoniu developes mwa standed dan tese dth t under real manufacture conditions. Re-generation of irradiated materials on the basis of PuO2 -ZrC can be carried e traditionath n o t lou mod f nucleao e r fuel reprocessing whic s basei h n o d extraction technolog f radiochemicao y l industry without change f existino s g scheme. maie Th n characteristic fuef so l materia presentee lar Tabln di eL U238 replacement by inert matrix in fast reactor fuel will result in certain after-effects, the main of which are as follows.

increasn 1A . reactivitf eo y change rate wit fuee hth l burn-up, giving rise problee th maintenancmf o o t requiref eo d refueling intervals. 2.Considerable increase of the total nonuniformity of the core power profil resula s f powea eo t r rate difference between fresd irradiatean h d subassemblies. This also gives appreciable dilutio coolane th f no t outlet temperature.

220 3. An increase of neutron leakage portion in the general neutron balance, leading to the decrease of sodium void reactivity effect (SVRE) and to the corresponding reactor safety improvement. . Considerabl4 e reductio f Doppleno r connected negative componenf o t temperatur powed ean r reactivity effects which would apparently require some special measures in order to intensify negative reactivity feedback.

Table I THE MAIN CHARACTERISTICS OF FUEL MATERIAL

Characteristics Value Porosity, % 80 Mas inerf so t matri fuen xi l elementg ,k 0.0353 Effective densit inerf o y t matri fuen xi l element, kg/m 3 1340 Coefficien f lineao t r expansio inerf no C t° matrixI I " 10 , 6.73 Thermal conductivity, W/m °C 20.5 Melting pointC ° , 3500

Using the fuel not containing U^•5 0 to burn minor actinides makes it possible generan i provido t l e criticality wit isotopy han minof eo r actinide grou r witpo h their mixture. However, this core desig s hardlni y feasible becaus higo to hf eo positive SVRE and low effective delayed-neutron fraction values. The above mentioned unfavorable consequences of utilization of the fuel containing inert matrix instead of U238 should be neutralized or alleviated at the core design development stage. o problemTw s have been considere d solvecourse th an d f n studiesi do e , which result presentee sar d f thesbelowo e e problemOn . determines i s e th y db high nonuniformity of power rate in different subassemblies, the other one - by necessite th f takinyo g into consideration physical safety criteria when choosing fuel compositio minoe th r rnfo actinide burning. e firsTh t problee solveb y n meanb dmca f transpositioo s f fueo n l subassemblies from the periphery to the center of the core at the refuelling. This provides fissile isotopes concentration distribution along the core radius that makes power rate profile more flat. This flattening extent depend t fuese ln o s burn-up, and it can correspond more or less to the core power profile needed. Anyhow, this method of flattening is rather efficient and it provides acceptable average core power rate value. Along wite assemblth h y transpositioe th n necessary coolant flow rate distribution adjustmen s providei t accordancn di e wit subassemble hth y power rat definitn e(i e range) t shoulI . e notedb d thae th t subassembly transpositio carries n wa durint dou g normal refuelin BN-60e th f go 0 reactor core.

221 Table II THE MAIN CHARACTERISTICS OF THE REACTOR AND CORE UNDER CONSIDERATION

Characteristics Value

1 Reactor thermal powerW ,M 2100

2 Efficiency, % 38 3 Maximum linear power of fuel element, kW/m 500 4 Average core specific power rate, MW/m3 400 5 Maximum fuel burn-up, % h.a. 50 6 Core subassemblies life-time, eff. days 300-400 7 Refuelling interval, eff.days 60-80 8 Breeding ratio 0 9 Sodium void reactivity effect Ak/% , k 0

The second problem is connected to the necessity of meeting physical safety criteria when providing efficient minor actinide burning. Utilizatio f fueno l composition based on plutonium and inert matrix results in considerable SVRE reduction (dow approximatelo nt Ak/k% 2 y- ) This allows some amoun minof o t r actinide e addeb o t sd inte oth fue l composition providin r examplfo g e zero SVRcoree bese th En Th .ti resultobtainee b n sca d using fuel composition made on the base of U and minor actinides. This provides acceptable range of effective delayed-neutron fraction thin O s. bas acceptabln ea e fractio f minono r determinedactinidee b fuee n th ca l n si . Calculations were made for the cores containing usual fuel elements, subassemblies and control rods according to the BN-800 reactor design[l]. Table II shows the main characteristics of the reactor and core under consideration. Preliminary calculation results have shown that satisfactory flattening of the core power profile can be achieved at the core refueling ratio of 5. In accordance with this, 5 annular sub-zones are provided each containing equal number of subassemblies, which are placed in succession in all these sub-zones during refueling operation r subassemblo s y transpositions froe periphermth e th o t y center of the core. In order to improve power rate distribution flattening two-zone core design is considered with different plutonium content in the subassemblies of both periphera d centraan l l sub-zones (this cor s similai e o that r t wito fuetw hl enrichment zones mad achievo et e required power rate distribution).

222 Fuel subassembly transpositions simila thoso rt e mentioned abov carriee ear d out during refuelin eacn gi h sub-zone maie .Th n characteristic corf so e types under consideratio presentee nar Tabln di e III. sho2 calculationaFigsd D w2 an 1 . l cormodelo etw r typessfo .

3. CALCULATION RESULTS

3.1. Codes

Codes use calculationr dfo followss a e sar :

- RBR-80 [2] diffusioD :2 n code with burn-up calculation modulr fo e- isotope concentration, power profile, fuel balance and SVRE calculations; MMCFK[3]- codeD 3 : , using Monte-Carlo metho Boltzmar dfo n kinetics equation solution, for more accurate calculation of SVRE and control rod worth.

ABBN-78[4] Russian constant system together with ARAMACO-C1[5] nuclear data preparation system were used in the codes.

Table III COR BLANKED EAN T COMPOSOTION

Characteristics One zone core(type Two zone 1) core(type2)

1 . Fuel subassemblies number in 565 565 the core including - central 565 175 periphera- l 390 2. Subassemblies number in the 384 384 radial blanket including: stee- l subassemblies 186 186 - boron shield subassemblies 198 198 3. Control and safety rods, 30 30 including: shi- m rods (SHR) 16 16 - control rods (CR) 2 2 safet- y rods (SR) 9 9 passiv- e safety rods (PSR) 3 3 4. Core height, m 1.05 1.05 5. Axial steel blanket height, m 2*0.3 2*0.3

223 S 150 Steel blanke ovet+ r core structures T E numbe subassemblief ro s E 113 113 113 113 113 L 12 SR 12SHR 105 2CR B 4SHR L A N K 150 Steel blanket + gas plenum + subassembly nozzless E T 61.47 23.0 17.0 14.49 12.88 50.

Fig Two-dimensiona. .1 l (R-Z) mode typf o l core1 e

3.2. Neutron Spectrum

Estimated index values, characterizing neutron spectrum in the core with zirconium carbide base inert matrix and without U 238 are represented in Table IV compare mixeo dt d fuel core. Average microscopic cross-section f o svariou s isotopes have been chosen as these indexes. Represented data show some neutron spectrum featurecore th ef o whers e U 718 is replaced by Zirconium carbide base inert matrix, i.e. considerable spectrum softening at the lower part of energy range and some its hardening at the upper part - for the fuel containing minor actinides.

3.3. Critical Parameters

Specific content (density f fissilo ) e materia e fueth l n i lelement e b n ca s considered as critical parameter for the core having no U 0-50 . This parameter

Tabl V eI AVARAGE MICROSCOPIC CROSS-SECTION VERIOUF SO S ISOTOPES(BARN)

Index 238 240 241 239 241 10 fJfU afPu QfAm GfPu CfPu acB

Fuel having no 0.042"' 0.365 0.286 2.492 3.895 5.389 U238 Fuel with inert 0.047 '' 0.448 0.362 1.818 2.340 2.417 matrix Traditional 0.046 0.396 0.308 1.791 2.398 2.523 mixed fuel

PuO2 1 235 U O7 + 35 % minor actinide oxides

224 150 steel blanke ovet+ r core structures S numbe subassemblief ro s T 35 35 35 35 35 78 78 78 78 78 E E 6SR 6SR 12 L 2CR SR, 105 6 CR B SHR L A N K E 150 Stee1 blanket + gasp enum+ £>ubasserribly noz;dess T

33.8 15.7 9.03 7.81 9.23 13.2 11.5 10.4 9.4 8.8 50.

Fig.2. Two-dimensional (R-Z) model of type 2 core

calculation result e representesar e followinth r Tabln di fo eV g three typef o s considered core composition: plutoniu- m oxide; - plutonium oxide with 10 % minor actinides addition; - U23 5 oxide with 35 % minor actinides addition.

NPP grade plutonium composition Pu239/Pu240/Pu241/Pu242 - 60/25/10.9/4.1%% has been used in analysis. For minor actinide mixture the following composition has been adopted: 237 241 242m 243 242 244 NP /Am /Am /Am /Cm /Cm -59.45/27.59/0.021/9.64/ 0.018/3.28%%. Minor actinide weigh 35%d t an content ,% indicate10 f so d above, correspond e corth eo t SVRE zero value. Calculations wer ee stabl madth r e fo erefuelin g procedures.

TableV SPESIFIC CONTENT OF FISSILE MATERIAL IN THE FUEL ELEMENT

Core fuel Composition materia base th en zirconiuo l m carbide impregnated with actinide oxide ZJ PuO2 PuO2+10% U 'O2+35% minor actinides minor actinides Fissile material 1280 1490 2060 density (kg/m ) Fuel element material mass material (heavy atoms) to the fuel element volume ratio

225 3.4. Reactivity Effects

3.4.1. Sodium Void Reactivity Effect

Sodium void reactivity effect was calculated for "before refueling" condition whic mose th s hti unfavorabl thir efo s parameter. Calculations were mad botn ei h diffusion and transport approximations. SVRE value for plutonium oxide version s i -2.22 %Ak/k when calculated using RBR-80 diffusion prograd an m -1.73+0.1%Ak/k when calculate y Monte-Carldb o method using MMCFK. This significant SVRE decrease as compared to that''of traditional fuel core, can obviously result from the neutron leakage fraction increase in the core having no U . If the fuel having minor actinide addition is used, the SVRE value is equal bees zeroo chosens thit d ha nsan wa t i , indicate s a , d above, when minor actinide permissible fraction was determined. 3.4.2. Temperature Reactivity Effects, Reactivity Change RateFunctiona as of Fuel Burn-up

Doppler effect value in the core having no U TTfi is sufficiently lower (almost 10 times s compare)a thao d t f traditiona o t l mixed fuel corecoule On . d suppose that effects related to the core size and shape changes would increase considerably in this core because of large neutron leakage rate. However, according to the estimates which have been made, chang ^ sensitivitK f eo relation y i cor e th e o nt size chang proves insignificante ha b o dt . Besides fuel material base madth en eo of porous zirconium carbide matrix has sufficiently lower thermal expansion coefficien comparison i t n with tha f oxido t e sameth r fuele Fo reaso. n (U238 absence) reactivity change rate as function of the fuel burn-up is appreciably increased. Estimated valuemaie th nf s o reactivity plutoniu e effectth r fo s m oxide cor representee ar e Tabln di . VI e

Tabl I eESTIMATEV D VALUEMAIE TH N F REACTIVITSO Y EFFECTR SFO PLUTONIUM OXIDE CORE

Axial Fuel axial Doppler Reactivity expansion expansion constant change as sensivity effect T d(Ak/k)/dT function of coefficient Ak/k /°C fuel burn up Ak/k/m durine gon month Ak/% k Fuel without 1.57*10" 0.17*10~5 0.0009 o j.i j U238 with PuO2 Traditional 1.35*10" 0.26*10° 0.007 0.72 mixed fuel

226 Similar characteristic fuee th l r witsfo h minor actinides additio valuee nar f so the same order of magnitude, although some additional Doppler effect decrease is possibl thin ei s case. The reactivity change rate as a function of the fuel burn-up is also decreased considerabl r thifo ys core type, makin t possibli g o increast e e refuelinth e g intervals in comparison to those for Pu oxide core (in our case - up to 80 eff. days). Thus, reactivity coefficients providing negative temperatur d powean e r feedbac e sufficientlkar y decrease core th e n di U wit contento hn . Perhapsn i , T">O order to increase negative feedback special measures would be required, e.g. resonance absorbers insertion into the fuel. Increasing reactivity change rate as function of the fuel burn-up rate results, e giveath t n shim rods numbe appreciabln i r e reactor refueling interval decrease. For the core design under consideration this interval has been decreased down. 60 days when using Pu oxide, and - down to 80 days when using U 5 oxide with 35% minor actinides addition. It shoul notee db d thareactoe th t r operation under decrease refueling interval conditions can be realized, although this operation mode is not perfect from the poin vief to powewf o r production. Ther some ear e ways withi desige frame th nth f increaso neo t e fuel bum-up compensation system efficienc y increasinb y g boron carbide enrichment value (from 60% up to 80% or 90%) and by means of functions redistribution between separate rod groups. In generalj acceptable operation interval provision for the nuclear power reactor using fue l problema s withoui , e b , U to whict s /1Rha h solve furthee th n di r design studies.

3.4.3. Core Power Profile Uniformity

The problem of keeping core power profile non-uniformity within the permissible limits can be solved by means of fuel subassemblies transposition from the periphery of the core to the center of the core during refueling. In the cases under consideration maximum fuel element linear powee rth valuen i e sar rang 440-50f eo 0 kW/m. When moving subassemblie necessare sth y coolant flow redistribution should be also provided in accordance with the power rate.

3.4.4. Efficiency of Actiniae Burning

Actinia e mose th e fue burnincarriee th tn i b lefficienf i t n dou y gca wa t without U23 8 is used in the core. Analysis has shown that using such core in the BN-80 gradP NP e f 0plutoniuo typg k burne eb 0 reactorn 60 mt ca o durint p ,u g1 year with Pu oxide fuel and - up to 250 kg of minor actinides (Np,Am, Cm) with U and minor actinides mixture fuel.

227 CONCLUSION

Studies BN-80 e madth r efo 0 type reactor have shown principal feasibilitf yo e corth e wit fuee hth l containing inert matrix , althoug instea U f do h some TO Q parameters of this core are different from those adopted in the reactor design. Detailed analysi s needei s provido dt e reactor safety becaus f sufficieneo t weakenin temperature th f go powed ean r feedbacks thid san , will obviously lead to special measures development aiming at the reactor safety increase. Reactor refueling interval reduction is also the disadvantage of this core, although some margin reae existth l n i sBN-80 0 s desigincreasit r nfo e (absorbe d bororo r n carbide enrichment increase, functions redistribution between control and safety rod groups). e possibilitTh e mosth f t o yeffectiv e burnin f highlgo y active long-lived actinides, accumulated as a result of nuclear power engineering activity, is the simulating factor for the further studies and development of the core without UQ 5 " O content.

REFERENCES

[1] MATVEEV, V.,I., et.al., "Studies, Development and Justification of Core with Zero Sodium -Void Reactivity BN-80e Effecth f o t 0 Reactor", Sodium Cooled Fast Reactor Safety (Proc. Int. Top. Meeting, Obninsk, 1994), vol. l,pp 45-59. [2] ZININ, A.,L, etal., Mathematical Mode f Faso l t Reactor Core, VANT Series "Physic Technologd san Nucleaf yo r Rectors" ,33(19834 . 59 ) [3] MAIOROV, N.,V., Code Complex MMCFK for Rector Calculations using a Monte-Carlo Approximation , developed by A.R.Frank-Kamenetski, ibid. 8 21 (1981) 16. ] ABAGYAN[4 , L..P., et.al., Group Nuclear Datr Reactofo a d Shielan r d Calculation, Energoizdat, Moscow (1981). ] BAZAZYANZ[5 , N.,O., et.al., ARAMACO-2 Syste r Neutronifo m c Data Providin Reactof go Shield ran d Caslculations, Energoizdat, Moscow (1976).

228 MONONITRIDE MIXED FUEL FOR FAST REACTORS

B.D. ROGOZKIN, N.M. STEPENNOVA, Yu.E. FEDOROV, M.G. SHISHKOV, O.N. DUBROVIN, L.V. ARSEENKOV All-Russia Scientific and Research Institute of Inorganic Materials, Moscow, Russian Federation

Abstract As regards the property combination, methods of fabrication, reprocessing and irradiation behaviour mononitride fuel Is concidered to be most promising (like monocarbide amd carbonitride) for inherently safe fast reactors.

One of the most important properties of mononitride fuel is its creep. The studies carrieRussin i t dou a establishe dependence dth f creeeo compositionn po , stresses, temperature and fission number.

Method f mononitrido s e fuel fabricatio sourcs a e enus materials oxided an s metals, weapon's grade plutonium included methoe Th . f fuedo l preparation from metal is most simple in. its instrumentation and technology and results in more pure impureties a (ar sfa comcernede sar ) mononitrid. In-pile BR-10, SM-2, MIR and BOR-60 reactors tests of mononitride fuel revealed its adequate compatibility with structural materials, lower gas releases compared to oxide fuel caesiuo n , r iodinmo e induced corrosio feasibilitd nan f burn-uo y p more than 9% h.a. at the heat generation rate of 400-800 W/cm. Studies of hydrometallurgical (Purex-process) and electrochemical molten salt method f regeneratioso n using unirradiated mononitride fuel indicated their promising characte feasibilityd ran . 1. INTRODUCTION The mononitride fuel (as well as carbonitride one) possess unique set of valuable r fo reactor fuel properties—high concentratio f o fission n elements, thermoconductivity (that increases with temperature rising), hiqh melting point, good compatibility with structural material coolantd san s (sodium, sodium-potassium alloy, lead, lead-vismut alloy etc.) possibilite th , f differenyo t way f reprocessinso g etc. That y thiiswh s fue s considerei l s mosda t advance r fasfo te reactoron d d spacan s e reactors /1,2/ Russin I . problee ath f mdevelopmeno f mononitrido e us d an t e fuels si e solvebroa th d carefull n an do d y planned scale e investigatioTh . n programme include followine sth g main items: • investigation of physical and chemical properties of UN, UPuN, UN-UC, PuC, UPnC, UPuC pre-anNn i d in-reactor conditions; • the development of fuel synthesis methods from initial metals and oxides and fabricatio f fueo n l that possess specified density, porosity, impurity content, form,sizes and compositions; e investigatioth • f compatibilito n f mononitrido y e fuel with steeld an s coolant pren si in-reacto d -an r conditions. developmene th • exprese th f o t s method r elementfo s s containmen fuen i t l composition checking; • fuel pinsubassemblied san s design;

229 • irradiatio f MeCn o fue N l, Me pinsubassemblied san s inBR-10, CM-2, MIR, BOR-6 BN-35d 0an 0 reactors; • the electrochemical and hydrometailurgica! reprocessing methods investigation.

Putting intpractice omononitridth e th f eo e fue buildinaimes e i l e th th o d t f go commertially profitable closed fue inherentlle cyclth r efo y safe fast reactor t shoulI . e be noted that the mononitride fuel investigations are performed simultaneously with the investigations of carbonitride and carbide fuels. Some deficiency of the mononitride fuel connected wit catchine hth f neutronsigo n reaction ca (n,pN n14 C )14 be eliminated by using of 15N isotope. The investigations of mononitride fuel UPuN was proceeded by UN fuel investigation. The main results are presented below. . CREE2 P

The creep (e) investigation was conducted in complete rahge of uranium carbide-nitride composition temperaturr sfo stressed ean s changin ranggn i e 1200 — 1873 K and 10 — 80 M Pa, correspondently. There were finded two extreme points - minimum creep rate for composition UC N and maximum creep rate for composition UCojNo. e valueTh 3f o whoss e ratee 2-10ar sd 5-10 an " " correspondently for temperature 1580 K and a =20 M Pa /4/. Steady state creep rate of uranium mononitride for stresses heigher then 40 M Pa doe t depensno describee b grai n dy o nma sizequatiod y db e an n

£ = a3exp (-Q/RT);

where a— stress, [MPa];

Q32657= , 0J

The uranium mononitride Uo.8Puo.2N creep rate depends on oxiden and carbide content: r oxidea Fo )carbid d nan e conten range th f 0.2—0.3%n ei to : £=35a1>27exp (-Q/RT);

b) For oxiden and carbide content less than 0.15%:

e = 308fj 1'35exp (-Q/RT);

MPa 0 4000= 6 ;Q - wher0 kal=1 e.u A facility for irradiation creep investigation in IRT-2000 and BR-10 reactors have been created.

The results of irradiation creep investigation have shown that the creep rate was ten times heigher than that for preirradiating testing for temperature 1000-1800 K. Conventional temperature boundary, where thermal creep is observed, depends

remarkably on fission densitu in the specimen. So, as the fission density changes from

3 l3 3 1.0-1crrf 3.6-100o 13 t cnrfboundare th y lowers from 1500 1250KKo t .

230 E URANIUTH D . MIXEAN 3 M D URANIUM-PLUTONIUM MONONITRIDE FABRICATION PROCESS

3.1. Equipment

In Russia there was built a number of devices and facilities for investigation and fabrication of small amount of uranium carbide-nitride fuel and mixed uranium-plutonium carbido-nitride fuel. Some facilities make possible to fabricate uranium mononitride fuel for the whole core of fast BR-10 reactor loading or for fuli-scale experimental subassembl f BN-35yo 0 reactor. Those fscilitie numbea e sar r of connecte e linon e n i glovd e boxes with implemented experimentad an l technological devicis celle fillee .Th s ar d wit inery hdr t fabricatiogase Th . n procesf so mononitride fuel depend grean si t methoextene th n o td that have been chooser nfo irradiated fuel reprocessin type th f fina eo n o l d producgan e r repeateth fo t n i e dus closed fuel cycle. It is nessesary to take into account the existing supply of oxide raw materials and accumulated metallic plutonium too. Therefore facilitie r mononitridesfo s fabrication from initial oxide metallid san c uraniu plutoniud man m were created r uraniuFo . m mononitride fabrication from initial uranium the facility of continiuos operation type was built. This facility have productivity about 1.5 kg per hour and is a prototype one for fabrication of mixed nitrides from initial metals. Ther designee t intear pu o d operatiodan n facilitier fo s fabricatio f microsphericano l uranium mononitride fue plasmy b l a meltin f powdego r and by centrifugal-arc method. It is nessesary to note that when the new data concerning the physical and mechanical properties of mononitride fuels or modified design of fuel pins and subassemblies determinappean ca e requirementrw w ene qualityo st , dessit sizd eyan of mononitride columns /2-8/.

3.2. Fabrication of nitrides from oxides s statewa t dI thar carbothermicafo t l methoe uraniuth f d plutoniuo d man m mononitrides fabrication one can use initial oxides itself as well as jointly precipitated uraniu plutoniud man m oxides amoune Th . f carboo t n adde mixturn di e dependn o s oxigen coefficien experimentalln o d an t y choosen valu f techologicaeo l surplus thas i t usuall rangn yi f 0.5—2%eo . Initial powder mixe milln di r mixerso f differenso t typs ei subjecte o pressint d g with pressure valu e slowle ar equa d yo 100-30t an l a MP 0 heated up to 1400 C in vacuum or argon and then in chemically pure nitrogen atmosphere. The duration of treatment with temperature 1650 -1850 C can be from 2 to 10 hours depending on activity of initial substances. The cooling of mixture is conducted in the inverse order. For this process accomplishing the furnaces of permanent and periodical operation are designed (for UN). Those furnaces can operate with suspeded granulated mixture. The fabricated mononitrid is analysed and its phase composition is defined. The finally obtained porouse mass is grilled and pounded until! the powder size not exeeds 40^m. e mononitridTh e powde s mixei r d with binding substances (ethylene glycol, paraffin, polyvinil acetate, zin r sodiuco m stéarate etc) pressurised ,an fuen di l stack form with pressure value of 100-300 MPa. This stacks are sintered in nitrogen at 1550-1750 °C. The temperature and duration of the sintering depends on required density.

231 The described scheme of UN and UPuN fabrication ensures the oxigen and carbon content less than 0.25%. In Russia there achieved the level of about 500 kg r yeape r productio f uraniuno m mononitride fuel from oxides.

For production capacity examination of developed method there was manufactured mixed mononitride fuel with plutonium mononitride content equa0 3 o t l and 35%. It was stated that the tempperature level can be lowered by 50 — 60 °C. Those results are well correlated with that obtained by Indian specialists. Ther s performewa e d technological researc n synthesio h f solio s d solutions UN-ZrN and fabrication of fuel columns with different density, shape and size. The carbothermical reduction makes possible to product the solid solutions of full variety f compositiono synthesye Th . s temperatur f solieo d solution equas i 180o t l 0— 2100 °C. There is developed the mrthod of synthesis from oxides of solid solutions Puo.55 Zrn.4. 5C fabricatioe 3.3Th . mononitridf no e fuel from initial metals There was conducted in Russia the complex research and development work for synthesis of uranium and plutonium mononitrides from initial metals and following fabrication of fuel columns with different values of density. This method makes possible to organize the low temperature process of synthesis at the temperature 200—650°C. As initial substances one can use initial uranium or plutonium as well as their alloy. Initial metals or alloys are hydrogrnated by hydrogen at temperature 100—220°C and resulting powders are nitrided by nitrogen at the temperature 200—650 C. The process is stable and lead to uranium nitride and mixed uraniu plutoniud man m nitrides formation. The uranium nitride fabrication can be made on the permanent operation type facility.

The uranium and plutonium nitrides are powders with grain size 4—100 «m. The grain size depend fabrication so n duratio temperatured nan .

It was stated that nitride powders are suitable for uranium and plutonium mononitride fuel columns of different density fabrication in spite of rather high nitrogen content fuer Fo .l column fabrication cold pressing with pressure 100-500 MPa and following sintering at 1550—1750 °C in argon or in argon-nitrogen mixtur e usedear numbeA . f differeno r t sintering type e developedar s . Resulting monophase mononitrides fuel columns have a specified grain size and density. In any case the even distribution of plutonium througth the fuel volume is provided (the deviatio impurite s lesni ) conten Th O s . , tha %) y(C mononitridn i t1 n e dependn o s initial substances purit doed t exceeyan sno d 0.15% meany B . f thiso s method there were manufactured experimental banchs of fuel columns for irradiation in BR-10, SM-2, MIR, BOR-60 and BN-350 reactors.

analysie 3.4Th . s facilities fuee th l r analysiFo s ensuring ther developes ewa l stageal r f fabricatiodfo so n f analysio t se se th methods suc chemicals ha , neutron, rentgen, a-radiographycal, metallographica microrentged an l n analysis. There have been buil foundatioe th t a f no neutron method for oxigen and nitrogen content determination in fabricated mononitride fue phass l ga pins er analysiFo . chromatographe sth usede sar .

232 4. IRRADIATION TESTS OF MONONITRIDE FUEL IN RUSSIA

Irradiation test f mononitridso e fuel have been began fro e latmth e sixtien i s SM-2, MIR, BR-10 and BOR-60 reactors. The aim of irradiation tests was as following:

• the irradiation properties of mononitride fuel (swelling, gas release, fuel - claddin fued gcoolan- an l t compatibility etc) investigation; hige th h burn-u• p fuedesign pi fuel d nan l core density optimisation; • the investigation of fuel fabrication technology (impurity content) influence on the irradiation properties of fuel. The main irradiation tests results:

l uraniu1Al . mixed man d mononitride fuel pins that achieve burn-ue dth p levef o l

4.0—9.0% wit lineae hth r power 1045—600 intactW/cs wa m . Fast neutron fluence

2 2 2 2 (E>0.1 4.38-10s wa ) -4.54-102 e irradiatiocm"Th . n timranges ewa do t fro 5 m0. years5 4— . Neither iodin caesiur eno m corrosio detecteds nwa maximue .Th m fuen pi l cladding deformation did not exceed 1.4%. The fuel-sodium and fuel-lead interaction wat foundsno .

2. The mononitride fuel swelling depends on oxigen and carbon content, core center temperature and fuel density. For gas bonded fuel pins with fuel density equal o =92t % f theoreticao l value, claddin cented gan diametem r m temperatur 9 5. r e 1440—147 e oxigeth carbo d C 0nan n content increasing from 0.1 0.4o t 5 % leado st swelling rate growth from 1.2—1.5 burn-u% 1 r 1.9—2.5o %ppe t burn-up% 1 r %pe . Fofuee th r l with density 88—90 witd an h oxigeD %T carbod nan n content less than 0.15 core th %e center temperature growt o 1800—220t p hu causeC 0 swelline dth g rate increase 1.5-1.7o sburn-upt % 1 tendence r %pe Th . swellino yt g rate decreasf eo UN and U Pu N fuels was noted when the fuel density decreased from 94% to 85—89% TD. The swelling of mononitride fuel with density 92% TD in sodium-bouded fue t exeel no pin d ds di 1.0-1.1 burn-u% 1 r r cente%ppe fo r temperature 1000—110. 0°C

fissios Ga . n3 product (GFP) release depend oxigen so carbod nan n content, fuel density and core center temperature. The gas release from MeN fuel in the He-bonded pins was 2—4 times lower than that from oxide fuel and was equal to 25—50%releasP GF e e Th .fro m mononitride fuel increases 1.5—2 times whee nth oxige carbod nan n content increases from sam0.3—0.4e o t 0.15 th . er wt %fo . %wt irradiation conditions. GFP release from mononitride fuel increases from 23—25 whe% e 40 nth %o t core center temperature increases from 1300—145 1850—220o t C 0° . C 0

mononitride Th . 4 e fuel keep structurale sth dimensionad an l stability better than monocarbide one. «-radiography and microrentgen analysis results show that there is no visible plutonium redistribution alon axiae radiad gth an l l direction fuee th l n columnssi e Th . amount of released cesium not exceedes 5%. The fuel-cladding interactio t year4 exceeaftenm no zonf 2— 0 o srd di e11 d irradiation in BR-10 and BOR-60 reactors and has the local character.Mainly it was a zon f increaseo e d etching n thaI . t zone thers note wa e increasin dth f carboo g n content. Besides whe mononitride nth e fuel contained less than f oxige0.15o . %wt n

233 and carbon carboe th , n concentratio interaction i time3 n2— s zon s lesewa s thar nfo fuel with carbon content 0.3—0.4%. 5. The irradiation tests make possible to conclude that it is worth while to use fuel with density value equa o 85—89t maximad l an D %T l open porosite th r fo y He-bonded pins witd an ,h density 92—95 liquir fo D d%T metal bonded pins.

For the good performance of mononitride fuel pins with gelium gap it is nessessary to decrease oxigen and carbon content to 0.15% wt. and never exceed the value 800—850 W/cm for the linear power rate. 5. THE MONONiTRlDE FUEL REPROCESSING

There are two ways of the mononitride fuel reprocessing that are developing in Russia: a) Hydrometallurgical reprocessing (PUREX-process);

b) Electrochemical reprocessin moltegn i n halides.

The resulting produc f hydrometallurgicao t l reprocessin f e o oxide ar gd an s electrochemical reprocessin gmetals- .

5.1. Hydrometallurgical reprocessing of mononitride fuel There was finished the development of first stage of hydrometallurgical reprocessing of nonirradiated uranium and mixed mononitride fuel with density 85—95. %TD After a pin cladding removing the mononitride fuel is dissolved in nitric acid. It was found that the dissolving rate for mononitriedes is essentially higher than the for oxides and the dissolution process is more stable. The dissolution rate of uranium mononitride can be approximated in the following way:

2 5 2 2 V=0.4885+0.2-1O' [HNO3]+0.1T 9 10' [HNO3]-2.03[HNOs] 5

o dissolutioe wherth - eV n rate, g/min.-c; m temperaturee th - T ; K , [HNO - nitri a] c acid concentration, mole/l. This dependency can be used in the following range of nitric acid concentration and temperature:

4.0—12.4 mole/; K r T=363 fo I36 5—

6.0—12.4 mole/l for T=333 — 313 K. The further operations can be performed in the same way as for oxide fuel reprocessing.

5.2. Electrochemical reprocessin mononitridf go e fue molten i l n halides

The developmen f electrochemicao t l reprocessin f nonirraditego d mononitride fue molten i l n halide initiates wa s n Bochvai d r Institute /10/. This method make possible to keep in the produced fuel the minor actinides for their following burning in reactors. There was investigated the following items: electrolyte composition (40—45% KCI, 40—45% LiCI, 10—16% UCls) methoe th , f uraniudo m cioride intrusion electrolytee inth , anode dissolutio f uraniuno m mononitride.

234 Electrolys conductes ewa elektrolysen a n di r tigh containinx ai r n bo built a n i t g showeargons wa t I . n tha circoniuma t , molib.den, rutenium remai e sludgth n f ni eo anodic part of electrolyser. A numbe f experimentro r electrochemicasfo l dissolutio f uraniuno m mononitride was conducted thosn I . e experiments there were obtaine f uraniudo aboug 0 mn i 50 t the form of dendritic shaped crystals. The catode deposit (70% of uranium and 30% of electrolyte melteds )wa , electrolyt sens repeateeo dwa t d electrolysi uraniue th d msan to refining meltin furthed gan r fabricatio f mononitridno e fuel. The oxige carbod nan n conten fabricaten i t initiad dan l practicalls fuewa l e yth same. The investigatio n i Bochvan r Institute have showed tha n principli t e th e electrochemical reprocessin f mononitridgo e fuefollowind an l g buildin f closego d fuel cycl fasr efo t reactor basie th f mononitridn so s o e fuel fabricatio electrochemicad nan l reprocessing is feasible.

feasibilit5.3e utilisatioTh .N f yo n 15. Bot hydrometallurqicae th h electrochemicad an l l reprocessingN permi e th t 15» utilisation. The catching of N during the hydrometallurgical reprocessing can be performed on the stage of preliminary oxidation of the mononitride.

During the electrochemical reprocessing the nitrogen is emitted as a gas on the stage of anodical dissolution. This gas can be easily bind into uranium or another element nitrides that desintegrate after heating and release the gaseous nitrogen. That nitrogen can be send to repeated nitriding. Fig.1 presents principle diagram of fuel cycle using hydrometallurgic and electrochemical reprocessing.

CONCLUSION Manufacture of monotride Manufacture of monotride fuel from oxides fuel from metals 1. UC-UN, UPuN have been Fuel equipping and investigate t a 1200d — 1873d an K assembly of FA stresses 10—80 M P A.

BREST-300 fast reactor with , UCN methodUN e , 2,Th . UC f so lead coolant UPuC, UPuN, UZrC, PuZrC fabrication have been developed from the oxides, metals, including weapons-grade plutonium.

3. Radiation tests of UN, UPuN, have been conducte e reactorth n di s Hydrometallurgie SM-2, MIR, BOR-60. BR-1 buro t p n0u reprocessing. Purex-process up ~9% h.a. at power rating 400—1000 W/cm, which Wastes demonstrated high operational Fig 1 Genera. l scheme fueth lf o ecycl e reliability with MeN-containing fuel elements.

235 4. It has been found that hydrometallurgic (PUREX) process of non-irradiated UC, , UPuCUN , UPu Ne operatinth reprocessin t e a b carrie t g n ou dca g apparatus-technological line of the PO "Mayak" RT-plant.

5. The experiments held at the VNIINM demonstrated a potentiability of non-irradiated MeN reprocessing by electrochemical technique in molten salts.

floA . w6 shee f closeo t d fuel cycl bees eha n offerred.

REFERENCES

1. Oriov V.V, Avrorin E.N., Adamov E.O., Solonin M.I. et al. Unconventional Concept f NPPso s with Inherent Safety. Atomnaya Ehergia, V.72, N.4, p.317, 1992. 2. H.Bernard, P.Bardelle, D.Warin. Mixed nitride fuels fabricatio conventionan i l oxide line. IAE ATECDO- C 466- , Vienna, 1988.

3. Reshetnikov F.G, Rogozkin B.D et al. Investigation of methods of manufacturing fast reactor fuei element pins from uranium monocarbide, mononitride or carbonitride. Atomnaya Energia, V.35, N.6, pp.377-386, 1977.

4. Evstjuhin N.A. et al Creep uranium carbonitride. VANT, series 1(12), p.19-22, Atominform, Moscow, 1983.

5. Uchida M., Ichikawa M. Compressive creep of sintered UC-UN solid solutions. f nuclo . J . mater., V.49, p.91-97, 1973/74.

. Dmitrie6 v V.D.Mosee . Capacital t e r wor , yfo vLI k mononitride fue reacton i l r BR-10 with burning untill 8,2% h.a. Third Conf. of Reactor Material, MAE RF, Dimitrovgrad, Russia, 1993.

. Rogozki7 n B.D, Koteiniko . Mononitridal v t e R.B. e fue r inherenfo l t safetd yan economical reactor Branc4 . h Industry Conf. Space Nuclear Industr y. Material d san Fuels. Abstracts . KI(RSC)RF E , MA Podolsk. , Russia, 1993.

. C.Prunier8 , P.Bardelle, Y.P.Pages, «.Richer, R.W.Stratton, G.Lederberger. Europen colloboratio n mixeo n d nitride fuels. Intern.Conf n Faso . t Reactord an s Related Fuel Cycles. V.11,p.15.9.1.-15.9.12 Kyoto, Japan, October 28 - November 1, 1991.

9. H.Blank, H.Bokelung. Problems expected in future fuel cycle development of dense fuels for LMFBRs. Advanced Fuel Technology and Perfomance, p189-213, IAEA - TECDOC - 352, Vienna 1985.

10. Dubrovin O.N, Rogozkin B.D. Electrochemical metho f mononitriddo e fuel regeneration in molten salts. Technical Conference of the Nuclear Society, NE-93, June28-Jule2, Nizhni-Novgorod, Russia, 1993

236 PLUTONIUM DISPOSA BURNIND LAN LEAGN I D COOLED FAST REACTOR

V.V. ORLOV Research and Development Institute of Power Engineering, Moscow, Russian Federation

Abstract Utilizatio f accumulateno d plutonium will take decadeeth thid san s problem should be considered in connection with the long-term prospects of nuclear power. papee Th r addresses.the requirement nucleae th o st r technologe th y r whicfo n hca large-scale development, reasons and results of the conceptual design of lead cooled Fast Reactor.

. INTRODUCTIO1 N

Irradiated fuel from NPPs contains about 1,000 t of the fission Pu; this amount wilincreasine b l 50-10y gb 0 t/y, mostlspene th n tyi fuel pools Francen d I . an K U , Russia large radiochemical reprocessing plantoperation i e sar n whic t fulha l capacity can separate about 20 t of R-Pu per year from the NPP fuel.

Reduction of nuclear weapons will add about 200 t of W-Pu which should be considered together with R-piutonium (for example, it can be mixed with R-Pu). Apart shoule on fro, dm Pu bea n mini r d abou froU e nucleatm HE th 2,00 f o 0t r weapon dilutee whicb n reachdo ca t h X<20% enrichment. Thivera s si y valucable fuel clos hean ei t binvalu0 f high-qualit1 o .t o et y coal. Accumulation of high-toxic Pu, increasing the risk of proliferation, brings about the fas desirs possibles a a tt i f manf eo o d y. ri peopl t ge eo t However, this cannot be done fast enough, because preparation and implementation of any option of Pu disposal or utilization will take decades, incurring considerable expenses. Firs f allnecessars o i t ,t i construco yt t facilitie r long-tersfo m safe Pu storage. The construction costs will constitute the major part of the storage discount costs, hence reducin e incentiveth g r fasfo s t emptyin e storagth f o ge facilities.

Finally, utilization of Pu and HEU will not be a cost-effective option in the near decades, with sufficient resources of U, excess capacities for its production and enrichmen markeU w io t d pricesan t . Accordin somo gt e estimates, depletio f ricno h ores and higher U prices may be expected by the second quarter of the next century, and only then serious economic incentived san wilU l HE appea d r usinan fo r u P g closin fuee gth l cycle. (Larg fueP le NP reprocessin g plants were constructe somn di e countries in the 1970-1980s awaiting rapid growth of the world's nuclear power and U prices, constructio f fasno t reactors causes.n Howeverma i theso t al , e edu , fell short expectations.e th f o ) Hence, there are sound arguments for considering utilization of Pu (and,

probably, HEU) within the framework of long-term development of nuclear power.

241 Loss of part of PuPu (241 ) in long-term storage and accumulation Aofm would hardly be treated as a decisive argument in favor of rapid utilization.

237 Today, the predominant opinion deals with long-term nuclear power evolution on a moderate scale using conventional reactors, largely, LWR. Report "Energy in the Tomorrow's World" presented at the XYWEC Congress (Madrid, 1992) forecasts doublin f nucleago r capacitie 202y sb 0 which wil primere l stilth f o l yaccoun% 6 r fo t energy produce e worldth n thin I i d. s case o extrn , a fuel wile needeb l d soon.

However e situatioth , develoy nma anothen i p r waye above-mentioneTh . d report indicates inadequate power supply for the major and rapidly increasing part of the Earth population whic s alsi h o confronted wite probleth h f environmenmo t a t degradation. Even now, the nonuniform distribution of conventional fuels is one of the contributor consideree th e internationa f th o o st d den periode lth tensiony b , d an , s resourcega d depletioan s l wiloi f l o nals o manifest itself. These problems will become more serious with more and more countries getting involved into industrial development. Seeing no real radical changes in the structure of the world's energy production, the author reporthe sof t urg governmenethe supporto t developmenthe t new of t energy technologies right now, not waiting for 20 or 30 years, to make them viable commercially and technologically. The fission nuclear power undoubtedly has the right potential for radical solution of the problems posed, and is the most prepared, among the new energy technologies, fora large-scale evolution.

This evolution may become the content of the next stage — "a second nuclear era", afte . WeinberA r g /1/ .t i depend solutioe th numbea n o sf no f problemso r , including fission fuel resources. The accumulated amounts of Pu and HEU can ensure th enexe starth f t o tstag shald consideree an b l d from this viewpoint.

2. LARGE-SCALE NUCLEAR POWER

Nuclear power can stabilize consumption of the conventional fuels, if by doublin capacities g it nexe th f t o d s mi everyear0 2 e orde n t i sth risey~ a y y b rsb century, increasing its share in the world's primary energy production to as least about 1/3 as compared to the current 5% level. However, such a rise could hardly comresula eas technica of t l evolutio conventionanof l nuclear technologieswe If . remain withi e framewornth f theso k e technologies cannoe w , t even convincingly justify such prospects. LWRs and other thermal reactors consume too much uranium while fast reactors capabl f solvineo problee gexpensiveo th to mose e mb th tunet o tt important Bu . dou t thing is that the nuclear technology of the first stage matured 40 years ago on the experience basith f so perseptiond ean s existe t thada t tim neithes eha r experimental nor theoretical ground r sucsfo long-terha m forecas safetys it f o t .

O The nuclear power experience constitutes 6-10 reactor-years, while the large-scale nuclear power would reach about 10 reactor/years in the next century.

The proof for elimination of NPP nuclear catastrophes would require consideratio f eventno s wit lowerprobabilitiee d hth an , whic0 1 usuallf e so hw o yd not know.the same as possible correlations between the events, unique events, especially associated with human activities, whic e difficul estimatehe ar b o t t n i d terms of probabilities.

238 Recognitio f nucleao n r powea conventiona s a r l technolog r large-scalfo y e applicatio n i botn h developin d developean g d countries will need nuclear technologies satisfyin requirementsw gne technologw ne A . y takes muche timew t bu , can speak about its development and demonstration in 15-20 years, since we are aware of its basic components in the existing civil and military nuclear engineering and in recent developments. The discussion below refers essentially to a new evolutionary technology culminatin benefite gfirse th th tf o sstage . w Hencene e th , stage in nuclear power may be launched as early as in the first decades of the next century. . REQUIREMENT3 NUCLEAW NE O ST R TECHNOLOGY

Above ail, this concerns a convincingly proved prevention of NPP nuclear catastrophes. Certainly, we are not speaking in terms of an absolute safety. We can consider accident as prevented if its probability has been reliably an estimated as much lower tha 0 n1 (fo r example, collision wit asterioda)n ha . Large-scale nuclear power is hardly possible without political measures adequate safeguarding against nuclear attacks on NPPs. However, one cannot ignore potential non-nuclear actions (acts of terrorism, missile attacks). These actions may be estimated at the maximum leve considered an l initiatins da g event r maximusfo m acciden desigtP callinNP nr gfo to prevent a catastrophic radioactive release. As a limiting value, 10 Ci in J equivalent can be adopted which is 1,000 times lower tha Chernobye nth l releas required ean extraordinaro sn y measure proteco st t population and the areas. If one is successful in proving that the release will remain within the above limits in case of an external effect resulting in maximum damage to NPP structure systemd an s s damage t wouli , d stil more b l eaccidente th vali r dfo s cause internay db l effects. With the catastrophes prevented, the probem of NPP safety becomes just a more ecological-economical problem. The probabilities of severe accidents of 10 —10 become permissible, as defined by the insurance against loss of NPP with limited radiation consequencies. Such estimate e considereb y ma s sufficientls da y reliable.

In a most general form, the requirements to the reactor consist the following: in case of a maximum accident , the fuel, coolant and reactor vault shall retain radioactivity releases withi specifiee nth d limits /2/.

The deterministic safety requirements to radwaste management technology lie in observanc f naturaeo l radiation equilibriu t mtheia r disposa lattaine e whicb n hca y db transmutatio storaged nan .

The proble f nonproliferationmo n viei ,f possiblo w enrichmentU e o n s ha , deterministic solution within the nuclear technology and demands political measures. The new nuclear technology shall contribute to implementation of such measures, makin t i gmor e difficul produco t t recoved ean fro u e fueP rm th l cycl r furthefo e r weapon applicatio facilitatind nan inspectione gth . To keep within known relatively cheap resources of uranium and current capacities of its production, its specific consumption should be reduced by an order as compare curreno dt t LWRs which correspond r higher o clos R 1 B o et o s.t Modern perceptions of power development do not require short Pu doubling time, which simplifies requirement breedingo t s , fuel power density, external fues l it cycl d ean duration.

239 Pb leve t steas l m Û11SOO + 14,75

1-pump ?—separating shell 2-vessel 8—fuel storage 3-thermal shielding 9—steam generator 4-CPS 10—reactor vault 5-core 11—rottviy plug o-suppoit posts

FIG. 1. BREST-300 layout.

Besides, generatio f nucleano r power shall remain economically profitable both for developed and developing countries. For reference, we may adopt the current LWR generation cost, thoug s reductioit ho simplifiet e du n d design technology, requirement licensind san g procedures woul f advantageo e db , The requirements seem to be contradicting and difficult to be fulfilled as a whole. However, safet d econome harmonizedan b y n ca ye inheren th f i , t safety .principles are applied as early as at the conceptual design stage. The nuclear philosophy, the key words of which "inherent safety" have been introduce . VeinbergA y b widn di alreads e eha us , y influenced moder designsP nNP . However potentiae th , l hazard f fasso t runaway, los f coolantso , fireexplosiond san s occurred at TMI and Chernobyl are inherent in modern reactors and cannot be eliminated completely. Engineered safety measures reducing the probability of such accidents have becom e majoeth r caus f nucleao e r plant sophisticatio highed nan r costs. To provide maximum safety together with design simplificatio lowed nan r cosf o t nucleaw IMPP ne e concep th se r th technolog f o t y shal e choseb l n base n thio d s philosophy.

240 . 4 LEAD COOLED FAST REACTOR

That wer reasone startee eth w y dswh developing LCFR afte Chernobyle rth e Th . conceptual stagreacto e closw th f no ecompletiono o e t s ri . Besides calculationad an l preliminarily design effort 100reactorsd e r 300an 0sMW fo 0 criticae ,60 th , l assembly experiments, steel corrosion tests on Pb-loops and some other tests were carried out e characteristicth o 73,4,5/t e Du . s inheren n LMFi t R (reactivity margi /3< K enff , feedback), lead coolant (chemical passivity, high densit d boilinan y g temperature, natural circulation)fueN lPu (density- N U , , thermal conductivity, thermal-radiation stability) catastrophic cours f severeo e accident avoidede b n sca .

Fig 1,2 present a design sheme of BREST-300 reactor in metallic vessel, Fig. 3 present desigsa n shem f BREST-30eo 0 reacto reinforcen i r d concrete pool. Fig4 . show e reactoth s r core structure. Tabl show1 e s parameter f performanco s f o e lead-cooied fast reactor with loading R-Pu. W-Pu loading in BREST-300 reactor needs to rise reactor core height to 1400 mm. W-Pu loading in BREST-600 need to rise the density of mononitride mixed fuel untill 0,95 TD with equal geometry of core. Table 2 shows reactivity effect d coefficientan s r BREST-30fo s 0 reactor (R-Pu) r W-Pfo , u loading this effects are like.

A-A (Fig 1)

1 charging vaJve 2 steam generator 3 FA 4 pump 5 pipes of heat lemoval air system

FIG.2. Cross-section over header branches.

241 1-1$

I — pump 7-separatms uhell 2-h«ot-i-esistant concrete 8-i»ii--cooled ohoiitid with f 3-low heat conductive concrete products 4-conti-ol «lements a-fuel storege , a-core 10-sleam generator 6-support posts 1)-rotary plug FIG. 3. Reactor BREST-300 pond type.

-automatic controller FA (d(r=9.6mm) -scram rod FA (d,r=104mm) -shid mro lead reflector unit -shim rod FIG. 4. Cross-section of BREST core.

242 Tabl e1 Performance lead-coolee th f so d reactors

Performance BREST-600 BREST-300

Thermal output, MWt 1400 700 Electric output, MWt 600 300 FA numbe corn ri e 357 185 Reactor core diameter, mm 3190 2300 Reactor core heightm m , 1100 1100 Fuel element diameterm m , 9.0; 9.6; 10.4 ; 10.9.06 . 9 ;4 Fuel element pitch, mm 13.6 13.6 Reactor core fuel UN+PuN UN+PuN Fuel loading (U+Pu)N, t 28 16 Loading Pu/239Pu+241Pu, t 3.73/2.72 2.2/1.6 Fuel lifetime, years 6 — 5 5 Time between loadings, years 1—2 ~1 Core breeding ratio ~1 ~1 Lead inlet/outlet temperature, ° C 420/540 420/540 Peak cladding temperatureC ,° 650 650 Maximum lead velocity, m/s 1.8 1.8 Power effect AK/% , K 0.17 0.16 Total effect, % AK/K 0.33 0.32

Delayed neutrons share, ßQu, % AK/K 0.35 0.35 SG outlet temperature, ° C 520 520 SG outlet steam pressure, MPa 24.5 24.5 t efficiencyNe % , ~44 ~44

In accidents with damage of the reactor vessel and building, reactivity insertion, trip of pumps, loss of cooling from by the secondary circuit, failure of steam generator, ingres f vapoo s r bubbles inte coreth o , releas f gaseouo e s fission products, there is no runaway at prompt neutrons, excess temperature rise, fuel failure and catastrophic radioactive releases /6/. The integral void coefficient is deeply negative and the dangerous local positive effect is not manifested. The parameter desigd an s n feature e primarth secondard f o san y y circuits prevenb P t freezin emergencn gi blockind ycirculatioan e th f go n lines asseso T . s safety margins, besides realistic, hypothetical accidents are also considered. In case of clad failure no fast fuel collapse will take place, with formatio secondara f no y critical masd an s fast runaway. Insertion of reactivity worth tenths of $/s does not lead to steam explosion with large release of mechanical energy.

243 Table 2 Reactivity effects and coefficients

The inherent safety enables simplifying Name Value e reactoth r design, control systed an m constructions, eliminating intermediate circuits Effects, AK/K : e maiemergencd ith n an n y cooling systems, Temperature -0.2- 10~3 safeguard vessel (a pool-type reactor without metal vessel is also considered), a spent Neptunium -1.0- 10~3 assembly washing system. Economical Power -1.6- 10~3 estimations allow one to expect the NPP cost to be equal or lower to that of NPP with LWR. Fuel burnup -0.5- 10~3 Cladding corrosion 0.5- 10~3 P NP closeThe d fuel cycle, electromechanical fuel reprocessind an / /? g 3 Margin for control 0.4- 10" its recycling together with other actiniden i s Summary margin 3.2- 10"3 the reactor, transmutation of I and Tc,

long-terd an 3, Cs m utilizatiod storagan r S f eno Lead solidification 5.0- 10~ e remaininth of g fission producte ar s Coefficients, (AK/K)/degrees: envisaged, enabling radwaste disposal with their radiation hazard equivalent to that of Lead density change 1.9- 10"6 mined uranium with its decay products. Radial core expansion -6.7 -10'6 e Closinfueth lf o gcycl t a NPPe , 6 .Axial core expansion -1.1 -10" combined recycle of all actinoides, elimination Doppler -5.6- 10"6 of uranium reactivitw blankelo d an ty margin contribut o t non-proliferatioe d controan n l Doppler constant -5.2- 10~3 measures. (AK/T)-T Delayed neutrons The concep bases i tcomprehensiv a n do e 3.6- 10~3 fraction, ^eff experience wite LMF th hd PbB Ran i cooled t expecno y nava o an d t l e reactorw d an s Neutron lifetimes , 5.3- 10"7 principle technical problem developinn si d gan demonstratin beginnine th nexy e b th t tf gi go century.

possible Onth f eo e way develoo st concepe pth f naturao t l safet subcriticaa s yi l fast reactor with an external neutron source based on a proton accelerator. Such an approach makes an obvious safety case for the reactor in all the possible and conceivable reactivity accidents. An urgent task numbe a bot r r Russifo hfo f othed o r aan r countrie utilizo t s si e within the next 2-3 decades the plutonium accumulated in storages which comes from NPPs, extracted from their fuel, and from dismantles nuclear weapons, converted into spent fuel utilizatiou P . reactorn ni s which will mee e higth t h safet economd an y y requirements discussed abov wild ean l operat open a n i e fue l e firscyclth tt a e prolonged stage, allows reducin time r clearingth efo meetind storagee gth an u P g f so the Pu management recommendations made by C1SAC of the US National Academy of Sciences. For example, amount W-Pu BREST-300 reactor core is fixed 2240 kg, annual amount R-Pu formed in spent fuel may be about 757 kg. Such an approach to Pu utilization can be especially attractive for the next stage of nuclear power industry advance, since it dispenses with the necessity for the

244 economically inefficient effort to create special "burners" or to burn plutonium in the operating reactors. . CONCLUSIO5 N The considered example (concept) demonstrates tha a vert y d safan e cost-effective nuclear technology can be developed within the framework of the current experience and tradition, which can become the base of a large-scale nuclear f a naturautilizinpowero s i y e t accumulateI wa .th lg , requirinPu d o specian g l "burners".

REFERENCES

1. Weinberg Alvin "Revisitin e seconth g d nuclear era: probabilitied an s practicalities" G Nucl. Eng. Int. Nov. 1992, p. 36.

. Adamov0 . 2 E . .Orlo V . V ,v "Requirement nucleaw ne sa r technologe th r yfo large-scale power industry", Proceedin f ARS'9go 4 International Topical Meetinn go Advanced Reactors Safety, v.2, p 636-642, Pittsburg, PA, USA, April 17-21, 1994 year.

3. Victo . Orlol "PhysicaV ra t e v l characteristic f leao s d cooled fast reactor", Proceedings 1994 Topical Meeting on Advances in Re'actor Physics, v. 1, p 348-356, Knoxville, TN April 11-15, 1994 year.

4. E. 0. Adamov, V. V. Orlov et al "Conceptual design of BREST-300 lead-cooled fast reactor", Proceeding of ARS'94 International Topical Meeting on Advanced Reactors Safety , v.1, p 509-515, Pittsburg, PA, USA, April 17-21, 1994 year. 5. V. V. Orlov et al "Lead-cooled reactor core, its characteristics and features", Proceeding of ARS'94 International Topical Meeting on Advanced Reactors Safety, v.1, p516-523, Pittsburg , USAPA , , April 17-21, 1994 year.

6. V. V. Orlov et al "Study of ultimate accidents for lead-cooled fast reactor", Proceeding of ARS'94 International Topical Meeting on Advanced Reactors Safety, v.1, p 538-543, Pittsburg, PA, USA, April 17-21, 1994 year. . RogozkiD . 7 B .l "Propertiea t ne synthesid san reprocessind san g technologf yo mononitride fuer inherentlfo l y safe reactors",Proceedin f ARS'9o g 4 International Topical Meetin Advancen go d Reactors Safety, v.1 382-389p , , Pittsburg , USAPA , , April 17-21, 1994 year.

Next page(s5 ) Lef24 t blank PHYSIC FASF SO T REACTOR-CONVERTE U-23O T 3u RP

V.G. ILYNIN, M.S. KOLESNIKOVA, A.V. MAGAEV, D.N. ZIABLETSEV Institute of Physics and Power Engineering, Obninsk, Russian Federation

Abstract

Some physical characteristics of BN-800 type reactor in uranium-plutonium and thorium-plutonium fuel cycle are considered in the report. It is shown that due to use of thorium in LMR some safety parameters are improved, such as the void effect, burn-up reactivity swing. Besides, BN-800 type reactor using plutoniu s fissionablma e materia providn ca l e effective burnin yeara u f morP go g , k g aboue k 0 tha0 66 t n50 U-233 being built-up therewith for light water reactors. At the same time, according to the calculations, reactor loaded with thorium has somewhat lower Doppler reactivity coefficient.

1. INTRODUCTION

In assesment of evolution routes for nuclear power engineering (NPE) at present time, solutio f traditionano l problems shoul e envisagedb d , namely, fuel supplier fo s NPE and saving fuel resourses, meeting the modern requirements for safety, etc., as e problemth wel s a l f increasino s g actualit - yutilizatio f weapon-grado n d civian e l plutonium, burnin r minimizatiogo f build-uno minof po r actinides. s well-knowi t I n that minor actinidebuilt-ue ar ) considerabln pi Cm s , (NpAm , e lighn i ) t amountkg wate 0 3 r - reactor( s s loaded with enriched uraniumn i fueld an , even larger amount reactorn i - s s with MOX-fuelsequence th o t e f reactioneo du , f so - decay ô uraniun d si an subsequend - man y , n- t isotopes. This makes treatmenf o t radioactive waste lonn si g terms more complicated. thin I s connection f U-23o e thoriud 3us an , s fuema l materials could servn a s ea alternative to the uranium-plutonium fuel cycle. In [1,2] the scheme of fuel cycle is proposed s followsa , : thoriu materiaw ra mloades e i LMR o th t U-23ld s dan a , 3 built up therewit e use b therman i dn hca l reactors thin I . s case, both weapon-gradd an e civil plutoniu usee b fissionabl s da n mca e material. A a resuls f suco t h organizatio f fueo n l cycle e followinth , g objectivee b n ca s achieved: 1) Effective usage of fissionable materials: plutonium is the most efficient in reactors with hard neutron spectrum, i.e., in LMR uranium-233 is efficient fuel for thermal-neutron reactor s i well-know t I . n that coefficien f conversioo t e th r fo n thermal-neutrons reactors with U-233 can approach unity. Thus, saving of fuel resource provideds si . 2) In LMR with thorium, no secondary plutonium is bred, and, evidently, such reactor can serve as an efficient Pu burner.

247 Z , am

30.0 ss

35.0 AB

50.0 INN OUT RB SS

R , sm

93D 30.6 40.0 30.0

Fig.l BN-800 type reactor layout

Notes INN — inner core RB - radial blanket OUT - outer core AB — axial blanket S S- stee l shiellding

3) When thorium is used in thermal and LMR the long-lived minor actinides are in fact bredno t . 4) Assessment f uranium-23o e us s e d shothoriumth an 3 o wt n e i tha, du t principle, safety parameters can be improved in both LMR and thermal reactors. Some result f calculationo s f physicao s l parameter r BN-80fo s - 0 R typLM e plutonium burner, U-233 breede presentee rar thin di s work.

2. CALCULATION SCHEME.

e calculationTh s were n i two-dimensionacarrie t ou d l geometry. Diffusion approximation was considered using 26 energy groups. The calculation scheme with dimensions of zones is presented in Fig.l. It is based on the layout of BN-800 type reactor [3] with uranium-plutonium oxide fuel. The following variants were considered: with oxid e cored breedin th fue ean n i l g zones with replacmenw ra f o t materia Th-232y b l varianto s weltw a ,s a ls with metallic thoriu e blanketh d mn i an t fissionable materia s oxidea l , respectively 2 dissolve, UO PuO d n metallii dan 2 c thorium matrix. It was not our task to optimize the reactor core, and, accordingly, the fractions of the material by volume were not changed in all the variants. The results of calculations as compared to the breeder variant of BN-800 type reactor are given in Table 1.

248 . RESULT3 CALCULATIONSF SO . CONCLUSIONS.

Changes in safety characteristics and those of breeding in BN-800 type reactor were investigated in different compositions of fuel of the reactor core and breeding zones. The following conclusions can be made:

3.1. Safety of the reactor.

1) Replacement of U-238 raw material by Th-232 results in decrease of sodium void reactivity coefficient, in the case of oxide fuel - by the value - J>ff. In the

variant with PuO2 fue thoriun i l m matri blanked xan t made from metalli , becauscTh e e of harder neutron spectrum, decrease of sodium void reactivity effect (SVRE) is less considerable r analogouFo . s variant with fissionable material U-23 e voi3th d effect becomes negative ,reactoe botth n hi r core. 2) In the variants with plutonium and thorium the value of burn-up reactivity swin s i decreaseg d considerably, whic s i explaineh d e facmainlth t y b thay t effectivnes f newlo s y bred uranium-233 exceeds tha plutoniuf o t m being burnedn a s A . explanation consides u t le , r change valuee th f so :

r threfo e^ - main fissionable isotope- U - s 233, U-235, Pu-239. Evidently plutoniur fo , m -23 wil" valuequae e . w 9b 1 lth f o eo t l As this takes place, analogous values of U)1" for uranium-235 and uranium-233 will be

Tabl el Physical characteristic »5-80f so 0 type reactor with different fuel composition i 2 b 4 niât compoation type core UBOS-PuOe ThOM>uD2 Th-PuD2 Th-U302 Mnnlrnt VOS ThOS Th Th AK.% 23 17 UB8 35 SVR core% ai 2fl 3D -O3 SVR. % 23 ZA 23 -LO K D-Tdk/d. T 43B-03 3ÛE-O3 17E-03 2JE-03 fc£T OÛ0352 QÛOS9B OJÛ0303 QU004 BRC QJ99 Qfff 0.71 054 BR MR 115 L27 102 G floue, kg 2020 . 2240 2230 ?gj30 Gha,kg 13300 13300 15000 15000 Rp kg/yea, u r 1S6 -520 -515 - RuS .kg/year — 807 880 12 M rare (INN/OUTm ^pp — 900/700 1230/1080 330/170 V rh, ppm — 30 55 40 •> aJ», ppm — 80 110 90

Notes Rpu - year amoun builpe ru P p u f t o t Ru3 - amount of U-233 built up per year AK - burn-up reactivity swing (147 days) - content of (Pa232+U232) in uranium unloaded

249 0.6d 1.1an 89 respectively. Therefore, requirement e reactivitth o t s y control system become les e varians th strict r tFo . with fissionable material U-23 thoriumd 3an e th , value of burn-up reactivity swing is somewhat increased, which must be taken into consideration in choosing reactivity control system for the reactor. 3) Doppler constant KQ is decreased in all variants, especially in the case of metallic thorium, which is caused by harder spectrum of neutrons. e varianth n 4I )s with thorium-plutoniu s mapproximateli j f valufuee e th l$ e y less% 15 , compare e referenceth o t d d reactor, which, evidently s explaini , a y b d smaller fission cross-section of thorium, as compared to uranium-238. However, in the variant with uranium-23 d thoriuan 3 m s 2increasedi > f f o greatet e du ,r yeilf o d

delayed neutron U-23n si compares 3a . Pu o dt e

3.2. Fuel build-up.

1) BN-800 type reactor with uranium-plutonium oxide fuel breedsg k abou 0 20 t Pu a year. In case the raw material is replaced by Th-232, such reactor is able of effective breedin f U-23o g - ( 600-663 0 kg/year) burning. morPu g k e tha0 50 n hande Therefore on problee e th , th plutoniuf n mo o , m utilizatioe th solveds n ni o d an , other, effective fissionable materia bres i llighr dfo t water reactors. 2) U-233, "purest U-23y "b axia e radiad bres 2th i an ln di l blanke ppm)0 5 - ( t. Conten f U-23reactoe o t th n 2i r core with plutonium amount - 100 o t s0 e ppmth n I . reactor with fissionable material U-233, conten f U-23o t n uraniu2i m being bres di tree times less.

REFERENCES

] MUROGO[1 V V.M., TROYANOV M.F., SCHMELYO f thoriuVo A.Ne n mUs i . nuclear reactors. Energoatomizdat. (1983). ] MUROGO[2 V V.M., POPLAVSKY V.M e rol f fasTh eo . t reactor solvinn i s e gth problems of power, fuel and ecology in the future, Sodium Cooled Fast Reactor Safety (Proc. Int. Topical Meeting, Obninsk, October 3-7, 1994). [3] BAGDASAROV Yu.E., et al. Reactor BN-800 - a new stage in fast reactor development. IAEA-SM 284/41, (1986).

250 Next page(s) Left blank SESSIO : WEAPONN6 S GRADE PLUTONIUM PLUTONIUM ROCK-LIKE FUEL INTEGRATED R&D (PROFIT)

T. MUROMURA, N. NTTANI, H. AKIE, H. TAKANO Japan Atomic Energy Research Institute, Ibaraki, Japan

Abstract

A process has been proposed for the transmutation of plutonium. New fuels of multi-phase material e fabricatear s y makinf conventionab o d e us g fueX l MO lfacilities . Thee ar y irradiated in LWR to generate electricity, and nearly 98% of Pu-239 can be transmuted. The spent fuels, thaassemblage ar t f mineral-likeo e phase f geologicallso y stable, would become withouW HL t further processing procese higa Th .s hsha priorit proliferation yi n resistance, environmental safet economyd yan .

1. INTRODUCTION f plutoniumo e us e Oth n, Japan Atomic Energy Research Institute(JAERI s beeha ) n promotin gbasia c tharesearcR t almosLW n ho t completely burn plutoniump u s , makint gi possibl disposo et t directly[l]i f eo procese Th . s consist f productioso chemicallf no y stable (rock-like) fuels in conventional fuel facilities, burning the fuels in LWR and disposal of chemically stable spent fuels without further processing researce th r developmend Fo han . t of the process, a series of works has been undenaken that consist of studies on the fuels and their burning. It also includes evaluation and analysis studies of proliferation resistance, environmental safety, technolog economd yprocess[2-4]an e th f yo . Recently optione th , f o s plutonium disposition have been discusse world[5-12]e th n di . procese Th s rock-like baseth n do e fuels would appea applicable b o dispositiore t th r efo n of the plutonium. Figure 1 shows a schematic diagram of the plutonium burning process (PROFIT process) e fuel fabricatee Th ar s. conventionan di fueX l MO lfacilitie almosy b s t ordinary procedures. The fuels are then irradiated in LWR to generate electricity, and the plutonium is almost completely burned. The spent fuels thus obtained will be disposable as high-level radioactive wastes (HLW) after 30-50 years cooling without further processing. Accordingly, this proces s expectedi s mee 0 objectivee 1 th t f denaturino s plutoniume gth , formin generatind gan stablW geHL electricity[2-4]. In the present paper, the outline of R&D program of the process in JAERI will be described together wit consideratioe hth chemicalle th f no y stable fueltheid an sr burnupn i s LWR.

Fuels Spent Fuels • Ul — Fuel Production LWR "u Facility :! fl " -gï^ P.pnÎQtPrQ ' ' -fi x—-v L7 ."* ^ d Repository n /L [iTi i i t ij -4 — \ \ __

Figur . e1 Schemati c Diagra PROFIe th mf o T Process

253 2. CONSIDERATION

2.1 Rock-like fuels

PROFIe Inth T process desirabls i t ,i e thachemicalle th t y stable fuels shoul designee db d to satisfy the requirements in TABLE I. ) Proliferatio(1 n resistance It is favorable that the fuels are chemically so stable that they cannot easily dissolved into nitric acid solution. Accordingly, the expectee yar inherentle b o dt y proliferation resistant From chemical properties of ceramic materials, some oxides appear to meet the above requirements[ 13]. ) Econom(2 y Use of conventional technologies and facilities is economical and reliable. The additives to dilute plutonium should be abundant and/or be readily obtained. In the fields of ceramics and nuclear fuels, much technological backgrounds have been accumulated in the fabrication and processing of oxide materials. (3) Environmental safety In the process proposed, the spent fuels are to be the high level radioactive waste(HLW) without further processing the d expectee an y,ar geologicalle b o dt y stable well-knows i t I . n that most of the minerals survived in nature are oxide compounds.

TABL . ERequkementI Candidatd san e System Rock-like th f so e Fuels

REQUIREMENTS MEANS POSSIBLE CANDIDATE CHEMICAL SYSTEMS FORMS ) Proliferatio(1 n Stable ceramics, Chemical stability Oxide Resistance Oxide Technological Oxide, Carbide, Stable ceramics, ) Econom(2 y background of fuel Nitride, Alloys Oxide production (3) Environmental Stable minerals, Chemical stability Oxide Safety Oxide

From the comparison of chemical properties and crystal structures of ceramic materials and minerals in TABLE II, two oxide systems have been chosen for the stable fuels that will have rock-like structures and compositions: PuO2-ThO2-Al2O3-MgO and PuO2-stabilized ZrO2-Al2O3~MgO systems. They will consist of fluorite, corundum and spinel type phases. e spenTh t fuels obtained will becom n assemblaga e f compoundo e s suc s fluoritea h , corundum, spinel, hibonite and alloys. The fluorite, ThO2 and stabilized ZrO2 will become the host phase of the actinide and the lanthanide in fission products. With the presence of excess corundum(Al2O3) e alkali th ,d alkalin an - e earth e solidifieb element o t y e b dar s formation of hibonite type phase(SrO-6AKO3). The spinel MgO»Al2O3 will also be a host phase of them. The noble metals, Ru, Rh and Pd will make alloys with Mo and Tc under low oxygen potentials. These phase analogoue sar stablo st e minerals. Estimated distributionf so FPs among them are summarized in TABLE lu. e rock-likTh e fuels propose e presenth n di t wor e multi-componenkar t oxides thas i t chemically stabl acin ei d solutions. The burnee yar conventionan di sucR s PWRha LW l ,

254 BWR and VVER-1000, as to be described in section 2.2. The spent fuels become an assemblage of multi-phase compounds \vith crystal structures similar to those of stable minerals, tha disposabls ti HLWs ea . They woul stable db e under weathering over remarkably long times. Accordingly, the fuel system appears to constitute an inherent chemical barrier, that would contribute non-proliferatio f plutoniumno .

TABL . CandidateEII Fuele th f sso from Compariso Ceramicf no Stabld san e Minerals STABLE CERAMICS, GEOLOGICALLY CANDIDATE FUEL STABLE MINERALS SYSTEMS (Constituent Minerals)

(1) Alumina(Al2O3) (1) (A) PuO2-stabilized ZrO2- Sapphire(Al2O3) Al2O3-MgO, (Zirkelite-Corundum- (2) Magnesia(MgO) (2) Spinel(MgO«Al2O3) Spinel) (3) Zirconia(ZrO ) 2 (3) Baddeleyite(ZrO2), (B) PuO2-ThO2-Al2O3- Zircon(SiZrO4), MgO, Zirkelite (Thorianite-Corundum- (Ca,Fe,Th)2(Zr,Ti,Nb)207 Spinel)

(4) Thoria(ThO2) (4) Thorianite(ThO2)

(5) Titania(TiO2) ) Rutile(TiO(5 2), Perovskite(SrTiO3)

2.2 Plutonium burning reacton I r burnup study, survey calculations have been performe somn do e simplified cell model f LWRso , fast reactor(FR higd an )h temperature gas-cooled reactor(HTGR)[2-4]e Th .

PWR. Pu-239 HTGR, Pu-239 FR, Pu-239 -—«. — . -_»_. ——A — - PWR, total Pu HTGR, total Pu FR, total Pu

0.0 0.5 1.0 1.5 2.0 Burnup (1000 days) Figur . e2 Compariso burnine th f no g plutonium amounts estimate cely db l calculatioe th n i PuO2-ThO2-Al2O3 fueled PWR, HTGR and FR

255 results are summarized in Figure 2. The conventional LWRs appears to be promising for the plutonium burning. Such LWRs can transmute more than 98% of Pu-239 and 85% of total plutonium plutoniue th d an , m quality becomes completel yspene pooth n ti r fuels FRsn I . , about 50% of total plutonium is transmuted but the quality of plutonium in the discharged fuel was still high. This show indeee ar s s thad FR tsuitabl plutoniur efo m recycling e initiaTh . l plutonium inventor HTGn yi s Rrelativeli y small thid an s , e smalleleadth o t s r plutonium transmutation rate. The void reactivity and the effective delayed neutron fraction ßeff have been estimated for thoria (PuO2-ThO2-Al 2zirconiOd 3an ) a (PuO2-stabilized ZrO2-Al2O3) fuel LWRsn i s s a , show e voi n TABLi nTh d . reactivitEIV s calculatewa y e e closb th o tha o t ef t d o t conventional UO2 -PWthorie th Rn ai fuelclost zirconie bu zer,o th et n oi a r fuelbotFo h. fuels, ßeff was small. It is therefore necessary to estimate the reactivity coefficients accurately by performing core burnup calculation and to carry out the safety analysis study to observe the reactor kinetics behavior.

TABLE lu. Estimated Distribution of Pu and FPs in Fuels and Spent Fuels

PF1ASE Thorianite Spinel Corundum Hibonite Alloys Zirkelite (MgOAl203) (A1203) (SrO«6Al2O3) (Ru-Alloys) ATOMS (cub.) (cub.) (hex.) (hex.) (hex.) Fuels Pu t femeao . • Xv.-x * •• 'f ~" •. j .. Spent fuels

Pu,Am,Cm ©

Ce,Nd © Cs,Rb O © Sr,Ba 0 © Ru,Rh,Pd,Mo,Tc © © : main distribution, O : secondary distribution

TABLE IV. Void Reactivity and Effective Delayed Neutron Fraction (ßeff) of PuO2-ThO2-Al PuO2d Oan 3 2-Stabilized ZrO2-Al2O3 Fueled PWRs Evaluated bv Cell Calculations

PuOo-ThO9-Al9O3 PuO7-ZrO?(Y,Gd)-Al7O3 Void reactivity (Ak/kk') 0%-*95% void -0.69 -0.16

ßeff W 0.30 0.29

256 . SCHEDUL3 D R& F EO

3.1 Long term schedule

The schedule of R&D consists of three phases, phase I of fundamental study, phase II of engineering study and phase DI of commercial operation, as shown in TABLE V. In Phase I, fundamentals of the fuels are being studied to substantiate formation of fuel materials. They include phase relations, crystal structures, chemical and physical properties fuee th l f materialso irradiatioe Th . n test f fueso l material carriee alsb e sar o ot usint dou g JRR-3M and post irradiation test facilities. In the study of reactor core design, neutronics characteristics and safety analysis are examined. The fuel behavior under reactivity-initiated accident conditions is studied by the Nuclear Safety Research Reactor (NSRR). Economic assessment will als carriee ob witt dou h technological evaluation. In Phas fuea , lÏÏ e productio fundamentae nth proceso t e sdu l data e obtaineb o t s di examine conventionadn i facilitieX lMO confir o st advancd man engineerine eth g parameters. Advancemen f fueo t l composition examinee s alsb si o ot d through irradiation test f fueso l assemblages. In the study of nuclear core design, the most favorable core performance is designe reactoe th d rd an safet investigates yi areae th accidentaf sn o di l reactivit thermad yan l hydraulics. Phasn I , commerciaeHI l operations wil startee b l d fueusinX l productiogMO n facilities LWRd an e spenTh . t fuels obtaine coolee dar 30-5r dfo 0 years, then seale canistern di d san dispose repositoriesn di .

TABLE V. Schedule of R&D

Phase Phase I Phasen PhasI eH years3 ( ) (5 years) R&D Items Fundamental R&D Engineering R&D Operation •Formation of fuels •Irradiation exp. •Disposition of Pu and spent fuels, •Engineering test, (site specific) •Reactor burnup •Optimization, calculation, •Demonstration, •Safety analysis •Advancement

3.2 Present status (1) Fuel study In the preliminary examination of the rock-like fuels, the phase relations of fuels and spent fuels have been studie reactioe th y db n between inert additives, PuO simulated 2an d FPse Th . phases produced were identifie X-ray db y diffraction analysis(XRD) distributioe Th . s FP f no is determined by scanning electron micro-analysis(SEM) and electron probe micro-analysis (EPMA). In the phase relation study of PuO2-ThO9-A12O3 and PuO2-ZrO2(Gd)-Al2O3, it has been shown that the fluorite phase, the solid solutions of PuO2-ThO2 and PuO2-ZrO2(Gd), are in equilibrium with corundum(A!2O3). In the sample prepared from the inert additives and simulated FPs, formation of hibonite type phase has been identified, that is in equilibrium with the fluorit corundum[2-4]d an e .

In the phase relation study of PuO2-ThO2-Al2O3-MgO and PuO2-ZrO2(Gd,Y)-Al2O3-MgO systems s beeha nt i , shown thafuee th t l compoun three-phasn i s di e equilibriu f mfluorito e

257 30 35 40

Figure 3. Results of X-ray diffraction of a simulated fuel in ThO2-Al2O3-MgO system prepared at 1,400°C under low oxygen potentials. (A) fueA : l compoun f 5%PuOdo 2-25%ThO2-60%Al2O3-10%MgO consistinf go fluorite(F)+corundum(C)-*-spinel(S) (B) simulateA : d spent fue f 5%FPs-25%ThOo l 2-60%Al2O3-10%MgO consisting of fluorite(F)-K;orundum(C)+spinel(S)+hibonite(H)+alloys(A)

+ spinel(MgOAl2O3) + corundum, and the spent fuel in five-phase equilibrium of fluorite + spinel + corundum + hibonite(SrO«6Al-,O3) + alloys, respectively, as shown in Figure 3. The XRD and EPMA results obtained show that the actinide and lanthanide are mainly dissolve sucn di h fluorite type phase s thorianita s e ThO zirkelitd an 2 e (Zr,An)Oe th 2d _xan , noble metals and molybdenum in Ru-base alloys, respectively. The alkaline-earth elements forms hibonite type phase SrO»6Al2O3 with corundum A12O3. Tentative phase relations obtained are shown in Figure 4. Region (1) in ThO2-Al2O3-MgO system in Figure 4~(A) shows a three-phase equilibrium of ThO2+MgOAl2O3-i-Al2O3 and region (2) a two-phase equilibrium of ThO2+Al2O3, respectively. Plutonium oxide PuO2 is dissolved in ThO2 by formation of solid solutions. Thus, the composition regions of (1) and (2) would be acceptabl fuee th l r useefo . Fro formatioe mth f hibonitno e SrO6Al simulatee 2th O n 3i d spent fuels quaternare th , y system of SrO-ThO2-Al2O3-MgO system in Figure 4-(B) is to be valid, that contains a four- phase region of ThO2+MgO»Al2O3+SrO-6Al2O3+Al2O3 and a three-phase region of ThO2 +SrO6Al2O A13+ 2O3. Measurement d evaluationan s f physico-chemicao s l properties have been started. Irradiation studies of the fuel compounds are going to start using JRR-3M with a thermal neutron flu f 2X10xo 1 n/sec-cm t laboratorie2ho , fued san l examination facility.

258 TABLE VI. Void and Doppler Reactivities and the Effective Delayed Neutron Fraction(ßeff)Estimated for Pu02-ZrO2(Y,Gd)-Al2O3 Fueled PWR Based on 2-Dimensional Core Burnup Calculation 'BOEC EOEC Void reactivity(%Ak/k) moderator density 100% 0.0 0.0 60% -0.322 -9.74 30% -3.42 -28.6 5% -17.7 -92.6 Doppler reactivuy(%Ak/k) fuel temperature 1200K -0.0982 -0.201 900K 0.0 0.0 600K 0.107 0.209 300K 0.253 0.467 3 3 ßeff 2.87xlO' 3.50xlO' BOEC : Beginning of the Equilibrium Cycle, EOEC : End of the Equilibrium Cycle

MgO (A)

SrO

ThO igO-AI203

rO-6AI2O3

A!2O3 Figur . e4 Tentativ e phase diagram f fueso l systems (A) ThO2-Al2O3-MgO syste mt arouna d 1400°C (1): three-phase regio f ThOno 2+MgO«Al2O3+Al2O3 (2): two-phase region of ThO2+Al2O3 PuO dissolves 2i ThOn di formatioy 2b solif no d solution. (B) SrO-ThO2-Al2O3-MgO syste mt arouna d 1400°C (1): four-phase region of ThO2+MgO-Al2O3+Al2O3+SrO-6Al2O3 (2): three-phase region of ThO2+Al2O3-fSrO«6Al2O3

259 (2) Reactor bumup study

In the reactor burnup and safety analyses, the core bumup calculation was carried out to accurately estimat e Doppleth e d voian re reactivity coefficient f zirconio s a fuel (PuO2- stabilized ZrC^-A^Oß) PWR. Figure 5 shows the plutonium transmutation characteristics obtained by the 2-dimensional core calculation. As much as 83% of total plutonium and 98% of Pu-239 is transmuted, and the quality of plutonium becomes very poor in the spent fuels. A total of 0.9 tonne of plutonium would be denatured every year on the assumption of a 1 GWe PWR operating at 80 % availability. As show TABL voin nDopplei e d deh an , EVI r reactivitie r PuOsfo 2-ZrO2(YJGd)-Al2O3 fuel are negative both at the beginning and at the end of burnup cycle. The reactivities and ßeff are, however, very small at the beginning of cycle. The safety analysis study of zirconia progresn i w no sfuele R wildlPW confir effece m th f the o t reactivitn mo y accident event. To avoid the negative effects of small reactivity coefficients of the zirconia fuel, the examination has been started for two modified fuel systems of thoria-zirconia fuels PuO2- (Th,Zr)O2-Al2O3-Mg d zirconiOan a fuels PuO2-ZrO2(Yb)-Al2O3-MgO-(W metal)A . heterogeneous core will als examinee ob d where these fuel partialle ar s y chargede th d an , reactor safet operatiod yan n contro studiede ar l .

. SUMMAR4 Y

The fundamentals of the rock-like fuels and their burnup characteristics have been studied procese th schedules a f indicatee o sar e benefitus e e ddTh th belown si .

(1) Economica technologicad an l l advantages procese :Th s fue consistX l productioMO f so n repositoryW HL d facilitiesan . R The conventionae LW ,y ar l technologies necessart no s i t I .y to built new facilities. Accordingly, the process will be accomplished by use of present technology and existing infrastructures with minimum investment in facilities. It will provide for a favorable net profit in the generation of electrical power.

Pu-238 Pu-239 Pu-240 Pu-241 . ... -Q... m Pu-24C + m 2 A tota u P l --e-

4 1. 2 1. 0 1. 8 0.0. 0 6 0.0. 2 4 0 Burnup (1000 days)

Figure 5. Burnup dependence of plutonium composition obtained by 2-dimensional core calculatio PuOn ni 2-Zr(Y,Gd)-Al2O3 fueleR PW d

260 (2) Environmental safety: The rock-like fuels are multi-phase fuels, that are chemically and geologically stable. The spent fuels also consists of stable minerals. They are not soluble in nitric acid by conventional way and would be stable under weathering for geological period.

(3) Proliferation resistance: The chemical and geological stability of the fuels will provide a barrier for Pu proliferation (chemical barrier). Bumup calculations shows that nearly 98% of Pu-239 could be transmuted in LWR. A total of 0.9 tonne of plutonium would be denatured every year by the operation of a 1 GWe PWR at 80 % availability.

(4) R&D time: By taking advantage of conventional technologies, the time required for the research and development for the process would be relatively short. After engineering R&D in Phase II, the disposition operation could be started.

ACKNOWLEDGEMENT

e authorTh s would . likMatsuura S o than t e. Mr ke directo th , r generaTokae th f io l Research Establishment of JAERI, for his valuable discussions on the R&D. They are indebted to Mr. K. Ouchi for the experiments.

REFERENCES

[1] Atomic Energy Commission, Japan, "Long-Term Program for Research, Development and Utilization of Nuclear Energy", Chapter 3, 8(2), Tokyo (1994)68-70 [2] Akie, H., Muromura, T., Takano H. and Matsuura, S., "A New Concept of Once- through Burning of Nuclear Warheads Plutonium", Seventh International Conference Emerging Nuclear Systems (ICENES '93), (Ed.) H. Yasuda, World Scientific, Singapore (1994)298-302 [3] Muromura, T., Nitani, N., Akie, H., Takano, H. and Matsuura, S., "Once-through Type Fuels for Plutonium Burning from Nuclear Warheads and their Burnup Characteristics", Proceedings in "Second International Seminar on Long-lived HLW Transmutatio Weapon-gradd nan e Plutonium Utilization Base Proton do n Type Fuel Plutoniusfor m Burning from Nuclear Warhead theisand r Burnup Characteristics", Proceeding "Seconn si d International Semina Long-liven ro W dHL Transmutatio Weapon-gradd nan e Plutonium Utilization Base Proton do n Accelerators", May, 1994, ITEP, Moscow, Russia(to be published) [4] Aide, H., Muromura, T., Takano, H. and Matsuura, S., "A New Fuel Material for Once-Through Weapons Plutonium Burning", Nucl. TechnoL, 107(1994)182-192 ] [5 Albright , BerkhoutD. , Walkerd an . , "WorlF , W. , d Inventor f Plutoniuyo d man Highly Enriched Uranium, 1992", SIPRI, Oxford University Press, 1993 [6] USDOE, "Openness Press Conference, Fact Sheets", (1993)21 [7] USDOE, "U.S.Department of Energy Plutonium Disposition Study", Technical Review Committee Report, July, 1993 ] [8 Willett , "DispositioW. . L , f Excesno s Plutonium", Trans . NuclAm . . Soc.9 6 , (1993)88

261 [9] von Hippel, F. and Feiverson, H., "Disposition Options for Separated Plutonium", Trans . NuclAm . . Soc., 69(1993)89 [10] Walter, C.E. and Omberg, R.P., Proc. Int. Conf. and technology Exhibition Future Nuclear Systems:Emerging Fuel Cycles and Waste Disposal Options(GLOBAL '93), Seattle, Washington, Sept. 12-17, 1993, Am. Nucl. Soc.( 1993)846-858 [11] Alberstein Hamiltond an Baxter „ . , I "ThD , M . . eC , A , MHTGR-Maximum Pluto- nium Destruction Without Recycle". Trans. Am. Nucl. Soc. 69(1993)94-95 [12] Biswas Garkisch d , Rathbun an D. , ,"Characteristic . D. W . . H , R , 100a f so % MOX- fueleDesigR Weaponr dPW nfo s Plutonium Disposition", Trans . NuclAm . . Soc9 6 . (1993)95-96 [13] HandbooBockA " , R. , f Decompositioko n Method Analytican si l Chemistry", International Textbook Co., London. (1979)102-116

262 COMBINING AN ACCELERATOR AND A GAS TURBINE MODULAR HELIUM REACTOR FOR NEAR TOTAL DESTRUCTIO WEAPONF NO S GRADE PLUTONIUM

A.M. BAXTER . ALBERSTEID , N General Atomics Diegon ,Sa , CaliforniaA US ,

Abstract

Fissioning the excess inventory of weapons-grade plutonium (WG-Pu) in a reactor is an effective mean renderinf so g this stockpile non-weapons useable enormoue Th . s energy content of the plutonium can be released by the fission process to produce valuable electric power fissioo N . n optio bees nha n identified destrothan ca t y more than about 72% of the WG-Pu without repeated reprocessing and recycling, presenting additional opportunitie diversionr sfo . However turbins ga e e th modula, r helium-cooled reactor (GT- MHR) combines high plutonium destruction and electrical production efficiency with favorable economics in an inherently safe system. Accelerator driven sub-critical assemblies offer almost complete WG-Pu destruction t achievbu , e this goal onl usiny yb g circulating molten salt solutions of plutonium, with difficult engineering problems and safety implications combininy B . GT-MHe gth R wit accelerator-driven ha n sub-criticaR MH l assembly, the best features of both systems can be merged to achieve the near total destruction of WG-Pu in an inherently safe, diversion-proof system. In this concept WG-Pu in the form of ceramic coated plutonium-oxide particles in graphite fuel blocks are irradiated to about 65% total plutonium bumup in GT-MHR modules. The discharged fuel element thee sar n placed directl sub-critican yi l GT-MHR modules wit acceleratorn ha - driven neutron source located in the central reflector of each sub-critical module. Depending on the plutonium destruction desired, more than 99.9% Pu-239 and more than 90% total plutonium destruction can be achieved. The discharged graphite fuel elements cae disposeb n f directldo standarn i y d multi-purpose canisters without additional processing. The modular concept minimizes the size of each unit so that both the GT- MHR and the accelerator would be straightforward extensions of current technology. No molten salt solutions of plutonium are required, and the fuel is always in a solid encapsulated form.

1.0 THE WEAPONS PLUTONIUM ISSUE

Plutonium can be used as a basic material in a fission weapon or as a trigger for a fusion (thermonuclear) weapon. The threat posed by the diversion of plutonium for such purpose receives sha d increased attentio strategiresule a s th n f a to c arms reduction treaty (START) agreements which will make surplus ten f metriso c ton f weapons-gradso e plutoniu mbotn i Unite e hth d State Russiad ssafean e Th ., permanent dispositio f thino s material presents a difficult challenge.

Fissioning the excess inventory of weapons plutonium in a reactor appears to be the most effective mean rendeso t stockpile rth e non-weapons useable. Materia consumee b n ca l d and the isotopic composition of the remaining material degraded to make it a less attractive weapons material. Also fissioe ,th n process lead dilutioso t materia e th f no l with highly radioactive by-products. Further enormoue ,th s energy content withi plutoniue nth m is released by the fission process and can be captured and converted to produce valuable electric power. However degree th , f degradatioeo n induced during irradiation a s ni important consideration.

263 No fission option has been identified that can accomplish the destruction of more than plutoniue th f o l mal f withouo abou % repeatee t72 tth d reprocessin recycld e gan th f eo irradiated fuel fro reactore mth . Eve f nreprocessini recycld gan allowedes i time th ,e require accomplisdo t h total destruction level 95%> s( vers )i y long somn i , e scenarios exceedin years0 g10 .

fissioe th Ol nfal options Gas-Turbine ,th e Modular Helium Reactor (GT-MHR providn ca ) e the maximum plutonium destruction (up to 72%) without reprocessing, and can generate electricit highese th t ya t efficiency inherentln a n i , y safe manner.

The discharged fuel elements from the GT-MHR modules, containing the remaining plutonium, can be placed in a similar, subcritica! core and further irradiated using a neutron source provided by an accelerator proton beam striking a target in the center of subcriticae th i core. Almost total destructio plutoniue th f 99.99no o t p m(u % Pu-23d 9an achievee 90b n % ca tota d) withouPu l t recycl thien i s mannerdischargee th d ,an d graphite fuel element disposee b n sca f directl do standarn yi d spent nuclear fuel multipurpose canisters without further processing (Ref 1 ).

GT-MHR/ACCELERATOE TH 0 2. R COMBINATION

destructioe Th weaponf no s plutonium usin GT-MHR/acceleratoe gth r combination would accomplishee b followine th dn i g manner: • Firs weapone tth s plutonium woul Irradiatee db higdo t h burnu GT-MHpn i R modules. 90% of the original Pu-239 and 65% of the total plutonium would be destroyed in this "deep bum" step. • Nex dischargee th t d fuel elements woul placee db subcriticaa n di i GT-MHR core which is driven by an accelerator through a target located in the central annulus (AD-MHR). The fuel elements would be used "as is", and no reprocessing of the fuel would be required before irradiation in the AD-MHR. These discharged fuel elements would be irradiated to obtain the desired final plutonium destruction using the accelerator/target as a neutron source. Over 99% Pu-239 and 83% total

Table I THE GT-MHR/ACCELERATOR COMBINATION FOR PLUTONIUM DESTRUCTION

Basic Unit 3 GT-MHRs + 1 AD-MHR

GT-MHR Power Level 600 MW(t) GT-MHR Fuel Cycle 1/3 cor r GT-MHR/yeaepe r

AD-MHR Power Level MW(t 0 MW(t7 60 20 )o t t ) a constant beam power AD-MHR Fuel Cycle Full cor AD-MHr epe R year tonne f WG-Pso u destroyed 0.8 per year

Net Electrical Power 975 MW(e) per year per MT

264 DESTROYS/DENATURES 32MTOFWPulN40YEARS

PU CONVERSION

SECURE STORAGE FUEL FABRICATION

DISASSEMBLY STANDARD INTERIM STORAGE 6T-MHR-S Jn2S5MW(t|NET SURPLUS MATERAIS

SUBCRITCAL 6T-MHR t95MW|i)NET(AVG) STORAGE, PACKAGING

ACCELERATOR GEOLOGIC REQUIRES-72MWI«» REPOSITORY

to uO»S Figur : AD-MHe1 R PLUTONIUM DISPOSITION SYSTEM LAYOUT plutonium destruction can be achieved in less than one year of irradiation in the AD-MHR; and over 99.99% Pu-239 and 90% total plutonium destruction during a three year bumup.

• Finally, the fuel elements discharged from the AD-MHR core could be disposed of directl geologia n yi c repositor spens ya t nuclear fuel without further processing.

This system achieves near total destruction of weapons plutonium without reprocessin witd ghan minimum potentia diversior fo l materiale th f no .

The current GT-MHR desig "deer nfo p bum f plutoniuo " m utilize thresa e year fuel cycl whicen i h one-thir core th discharge es i f d o d each year. Sinc AD-MHe eth Rs i batch loaded, combining the annual discharge from three GT-MHR modules into a single AD-MHR core permit destructioe sth Pu-23e th totaf a 4 n o f 9n i oveo l % 99 r years. A schematic layout is shown in Figure 1, and basic parameters are given in Table I.

The small modular design of the overall system means that the beam power requirement acceleratoe th r sfo relativele ar r y smal thao s lt represent i t sa straightforward extensio f currenno t technology combinee Th . d syste GT-MHR3 f mo s an t electrica AD-MHd1 ne a s Rl ha outpu almosf o t t 1000 MW(e) acceleratoe Th . r requires only 72 MW(e), and with the very high efficiency of the gas turbine, the AD- MHR averages 195 MW(e) annually.

Both reactor types use the plutonium oxide coated fuel particle which has been successfully tested to about 747,000 MWD/Mt, and up to 1400°C, in irradiations Peace th carrie n i h t Bottodou m reacto Unitee th n ri d States.an DRAGOe th dn i d Nan Studsvik reactors in Europe (Ref 2-4).

The thermal neutro cornR spectruhares i MH de enougth mn i successfullho t e th e yus large, , Er-160.4eV 6 7 neutron absorption resonance. Thus this material fore mth n i , of coated particles of natural erbium oxide, is used as a burnable poison to control excess reactivity and ensure a negative core temperature coefficient under all conditions fertilo N . e material required additionao n d an , l plutoniu mcreates i e th dn i irradiation process, thu systee sth m truly maximizes plutonium destructio doed nan s not suppor plutoniua t m recycle economy.

The MHR, being graphite moderated and helium cooled, is capable of operating at sufficiently high temperatures that the helium coolant can be used in a direct Brayton cycle to drive a gas turbine which converts thermal energy to electricity at almost 50% efficiency. This results in excellent economics for the system when the electricity produced from the system is sold to the grid, and produces minimum thermal discharge to the environment.

3.0 DESCRIPTIO SYSTEF NO M COMPONENTS

3.1 GT-MHR Plant Design

The GT-MHR module arrangemen show s i tbasie th Figurcn d i parameteran , e2 f so the core are summarized in Table II. The components for each module are contained within a steel vessel system which includes a reactor vessel and a power conversion vessel. These side-by-sidvessela n i e sar e configuration connecte crosa y db s vessel vessee Th . l syste msites i d undergroun concreta dn i e silo, which serven a s sa

266 Figure GT-MHR2: MODULE ARRANGEMENT

267 independent high pressure containment structure reactoe Th . r vesse mads i l f higeo h strength modified 9Cr-1Mo-V alloy stee! and is approximately 8.5 m diameter and about 23.8 m high. It contains the annular reactor core, the core supports, control rod drives, refueling access penetrations shutdowa d an , n cooling system reactoe Th . r vessel is surrounded by a reactor cavity cooling system which provides totally passive safety-related decay heat removal shutdowe Th . n cooling system locatee th t da botto reactoe th f mo r vessel provides decay heat remova r refuelinfo l d gan maintenance.

Table II BASIC PARAMETER GT-MHE TH F SRO

Core Power Level MW(t0 60 ) Core Power Density 6.58 w/cc Core Coolan Pressur& t e Helium @ 7.07 MPa Core Inlet Temperature 490°C Core Outlet Temperature 850°C Core Shape Cylindrical Annulus Core Size: Height 7.93 m (10 blocks) O.D. 4.86 m I.D. 3.0 m Number of Fuel Elements in Core 1020 Fuel Cycle Length 3 years

The reactor core consists of an assembly of hexagonal prismatic graphite blocks. The fuel blocks are arranged in three annular rings The center and outer portions of the core comprise unfueled removable graphite reflector blocks core eTh . assembls yi surrounded by a steel core barrel and contained inside the uninsulated reactor vessel. When operating as a plutonium burner (called a PC-MHR), the fuel cycle is a once- through three-year cycle with one-thir core th e f d o refuele d approximately ever1 y1 months.

The power conversion vessel is also made of 9Cr alloy steel and is approximately 7.6 m in diamete aboud ran t 22. high9m . This vessel house turbomachinerye sth plate-fia , n recuperator, and a helical coif water-cooled intercooier and precooler. The turbomachinery include generatorsa compresso o turbinea , tw d an , r sections submerged in helium, all mounted on a single shaft supported by magnetic bearings.

A simplified flow diagram of the GT-MHR is shown in Figure 3. The helium coolant exits the reactor core at 850°C and 7.02 MPa, flows through the center hot duct within the cross vessel, and is expanded through the turbine. The turbine directly drives the electrical generator and the high and low pressure compressors. The helium exits the turbine at 510°C and 2.56 MPa and flows through the plate-fin recuperator. This compact and highly efficient device keeps as much energy as possible within the cycle,

268 3 <90°C (91S°F) 7.07MPa(102Spti) MODULAR REACTOR CORE

850°C (156ZT) rjQMPa (1017p*() RECUPERATOR

131«C (269>F) Z62MP»(380pSi}

FROM HEAT HIGH SINK PRESSURE COMPRESSOR 112"C (233-F) 7.24MPa(10«p*l)

FROM HEAT StJK

33-C<91«F) Z60MP»(377p*l) MTERCOOLEV R 7 LOW PRESSURE COMPRESSOR

Figure GT-MHR3: FLOW SCHEMATIC

contributin hige th h go plant t efficiency heliue Th .m then passes throug precoolere hth , and pressur w enterlo e sth e compresso t 33°Cra , wher compresses i et i 4.3do t 5 MPa. To maintain high pumping efficiency the gas is sent through a second cooler before going throug hige hth h pressure compressor. From ther et passei s througe hth recuperator where it picks up heat from the helium exiting the turbine, flows through the outer annulu crose th f sso vessel pas p reactoe u , th t r vessel walls finalld an , y flows down through the core, entering at 490°C and 7.07 MPa, to complete the loop.

3.2 The AD-MHR and Target

As show Figurn AD-MHi e th , e4 R cor veres i y simila GT-MHe th o rt R coree Th . primary differences are the provision for an accelerator target for neutron production by

269 spallation surroundina d an , g pressure vessel, both locate centrae th dn i l graphite reflecto fac e show s figuree th ra t th d tha systee n i an ; th t m would alwaye sb subcritical. The central graphite reflector serves to moderate the spallation neutrons and produc thermaea l flux spectru fuee th l mannulusn i . Tabl I provideeII summarsa y of the basic parameters. The reactor core components, fuel, and all the power conversion system components are identical to those in the GT-MHR and are described previoue inth s section.

Table ill BASIC PARAMETER AD-MHE TH F ACCELERATOSRO & R

Core Power Level: Maximum 600 MW(t) Average 410 MW(t) Minimum 210 MW(t) Fuel Cycle Length 300 days Accelerator. Proton Beam 800 A Mevm 3 3 , (avg) Beam Power 28 MW Power Requirements 72 MW(e) Neutron Source Strength 10x 6 s 1 8n/ ~

REPLACABLE CENTRAL 3€-K-0PER*f IN&l NOT & SIDE REFLECTORS NEEDED

TARGET & PRESSURE BORATED PINS (TYP) VESSEL

REFUELING PENETRATIONS

ACTIVE CORE CORE BARREL COLUMN2 10 S BLOCK0 1 S HIGH .12 X START-UP CONTROL RODS

PERMANENT 10 X RGCC \ NOT SIDE -eHAfîNEfcSJ NEEDED REFLECTOR

Figure AD-MHR4: CORE LAYOUT

270 110

Legend Total Pu Bumup

X Preliminar- y Design Point

50 400 600 700 800 900 TIME (days)

Figure 5: AD-MHR PLUTONIUM BURNUP VS. TIME

Several options are available for the accelerator target in the AD-MHR. However, based on studies by Los Alamos National Laboratory, the best design appears to be one using molten lead flowing through the target, cooled by helium. The lead has a verparasitiw ylo c absorption cross sectio r neutronsnfo handln ca , energe eth y deposition from the accelerator without damage, and does not have stress or other problems arising from the spallation process (Ref 5).

The target woul t intdcentrafi e oth l graphite column spacAD-MHRe th en beai e Th m. window separatin protoe gth n beam vacuum tube fro leae mth d target woul locatee db d at the top of the vessel, well above the core. The beam would then pass through a low pressure helium tube before strikin leae gth d target highese Th . t neutron production targete th f intensito , p wit productioe to hth e y th woul t a ne d b rat e decreasing linearly with depth rangprotoe e th Th f .e o n beam woul aboue dmeteb targete e th on t n i r .

271 Calculations show that this neutron production distribution leads to a satisfactory power distributio sub-criticae th nn i l core assembly protoe th .f o nAbou bea% 60 tm energs yi deposited as heat in the target, and the lead would be liquid at operating temperatures; it would be cooled by helium. The flow rate of the lead through the system would be about 0.5 meters/second. The 800 MeV protons from the accelerator would produce 20 to 25 high energy neutrons ( ~ 15 MeV) per proton incident on the lead target.

3.3 Accelerator

The accelerator design requirements are significantly less than for other applications being considered, and, as shown in Table IV, represent reasonable extensions of current technology.

TableV I ACCELERATOR DESIGN REQUIREMENTS COMPARED TO CURRENT TECHNOLOGY

COMPONENT STATUS REQUIREMENT

Injector > 100 mA demonstrated <125 mA RFQ V Me 6 0. , mA 0 7 <125 mA demonstrated Drift Tube Linac pulseA m 0 d 10 250 mA peak, pulsed configuration operated at operation LANL 250 mA at CERN Coupled Cavity Linac operation i A m 7 t n1 a 30 to 250 mA LAMPF RF Generators klystronW M 1 s available 1 MW Tubes from industry

AD-MHe th r Fo R concept proton producee sar accelerated dan energn a do t f abouyo 0 t2 separate peako MeA tw m Vn i ,5 curren12 , t beams before being combine funnea dn i r lfo acceleratio drife th t ntubn i e linac thin I .s stage the acceleratee yar energn 0 a do 80 t f yo reacd an peaha befor V A km Me curren e0 strikin25 f t o lea e gth d efficienc target% 42 A . y in converting AC power input to beam power is achieved by pulsed operation. Each proton pulse is 1.11 ms long, and the pulse frequency is 120 pps. At this pulse rate, calculations show tha normala t , steady-state neutron flux leve achieves i l AD-MHe th dn i R core, becaus lone th g f eaverago e neutron lifetime.

4.0 RESULTS

Figur eshow5 plutoniue sth m destruction capabilit AD-MHe th f yo R using plutonium fuel elements discharged from the GT-MHR after a 3 year bumup. As can be seen from the figure, less tha threna e year exposur AD-MHe th en i R core (900 days) tota result% l90 sn i plutonium destructio 99.99d nan % Pu-239 consumption. Detailed dimensional core burnup calculations show acceptable power distribution systee th n si m throughout this exposure time.

272 Even a single year bumup (300 days) results in > 99% Pu-239 destruction and 83% overall plutonium consumption. Figur egraphicall6 y compare WG-Pe sth u destruction achieved PC-MHe inth AD-MHd Ran compares Ra othedo t r options being considere dealinr dfo g with this material. As discussed in the next section, these levels of destruction are more than adequat preveneo t furthey an t r plutonium weapone th f o e sus .

100

WEAPONS VITRIFICATION SPENT FUEL DEEP BURN RAPD PLUTONIUM OPTION OPTION OPTION OPTION (LWR-MOX) CMHR] (MHR-ACCEL) Figure 6: COMPARISON OF DISPOSITION OPTIONS TO CONSUME AND DEGRADE PLUTONIUM

5.0 MEASURES OF PROLIFERATION RESISTANCE

Table V summarizes the diversion resistance of the spent plutonium fuel from both the PC- MHR and AD-MHR, including the isotopic composition of the fuel relative to the original seee b WG-Pu nn fro ca tablee s m A th AD-MH. e ,th R fuel cycle offer extremeln sa y high leve resistancf o l diversioeo t proliferationd nan plutoniue Th . m conten f eaco t h spent fuel elemen ver(abous i tw ylo t one4ent hkilogram)a plutoniue th d ,an mmucs i h more difficult to recover from spent PC-MHR fuel than from spent MOX fuel. Since oveAD-MH0 15 r R fuel element requiree sar criticaa r dfo l mass woult i , vere db y difficul divero t sufficiena t t quantif thif yo s fue obtaio t l weapons-usefuna l quantit plutoniumf yo technologe th ; r yfo separating plutonium from MHR spent fuel has not been developed; and the isotopic mixture of the spent fuel makes it unattractive for weapons applications.

The radiation level in the AD-MHR discharged plutonium is extremely high, even after it has been separated fro fissioe mth n products internae Th . l heat generatione ratth en i plutonium spontaneoue th d ,an s neutron emission ratbote ear h large enoug precludho t e f thio se materiaus weapoa e s th a l n evecountra y nb y with sophisticated weapons technology.

6.0 CONCLUSIONS

The gas-turbine modular helium reacto providn ca r verea y high leve f weapons-grado l e plutonium destruction, without the need for recycle, in an inherently safe system which provides maximum efficienc r electricayfo l energy production minimud an , m environmental impact. This system is also very cost effective because of its high thermal efficiency.

273 Table V DIVERSION RESISTANC AD-MHF EO R DISCHARGED PLUTONIUM

WG-Pu LWR PC-MHR AD-MHR Discharge Discharge Discharge

Compositionf o ,% WG-Pu

Pu-238 0% 0.1% 0.1% 0.3%

Pu-239 94% 47.9% 9.8% 0.7%

Pu-240 6% 18.7% 10.7% 5.0%

Pu-241 0% 10.6% 11.6% 4.3%

Pu-242 0% 4.0% 3.3% 6.8% Reflected Sphere:

Critical Masss Kg , 8.0 9.6 15.2 20

Elements req'd N/A 1 55 150

Volume, litre N/A 170 4,977 13,575 Weapons extreme useful very low not useful usefulness

By combining the GT-MHR with an accelerator driven subcritical assembly, this high efficiency, minimum environmental impact, and inherent safety can be retained while providing near total destruction leve f weapono l s plutonium without recycle.

REFERENCES

[1] A.L LOTTS, et al., "Options for Treating High Temperature Gas-Cooled Reactor Fuel for Repository Disposal", ORNL/TM-12627, Oak Ridge National Laboratory, February, 1992.

[2] BARR , POLLITTP. , , REDDINGN. , , G.B., "High-Temperature Irradiation Experiments Plutonium-Bearinn o g Coating Particle Fuel" Plutoniun i , Reactoa ms a r Fuel. IAEA, 1967, SM-88/38 . 391-417pp , .

] FORMANN[3 , etal.E. , , 'The Irradiation Behavio f Plutoniuro m Coated Particlee Fueth n i l Dragon Reactor Experiment", Dragon Project Report 526, 1967.

[4] SCHEFFEL, W.J., "Design and Operational Evaluation for the Plutonium Test Element (FTE-13)", Gulf-GA-B12271, Gulf General Atomic, August 1972

[5] BEARD, C.A., et al, "Flowing Lead Spallation Target Design for Use in an ADTT Experimental Facility Locate LAMPF"s da Alamos Lo , s report LA-UR-94-2495

274 PHYSICA TECHNICAL-ECONOMID LAN C ASPECTF SO WEAPON-GRADE PLUTONIUM UTILIZATION IN HTGRs

A.I. KIRYUSHIN, N.G. KUZAVKOV, N.G. KODOCHIGOV, A.S. KUDRYASHOV, Yu.P. SUKHAREV OKBM, Nizhny Novgorod

N.N. PONOMOREV-STEPNOI, N.B. KUKHARKIN, E.S. GLUSHKOV, V.N. GREBENNIK RR , MoscoCKI w

Russian Federation

Abstract The curren te prograth work f n reductioo o mse th n i n strategic nuclear armaments will leao releast d f o elarg e stock-piles of weapon-grade plutonium (WPu). Storage of these quantities of weapon-grade plutonium requires complicated equipmen largd tan e economic expenses. The proble f plutoniuo m m effective utilization requires urgent solution. Ther e somear e waytha r gooHTGRd a sfo tan e d sar base for plutonium burning. In this paper the estimations of physical and technical-economic characteristic HTGf so RW poweM e wit0 ar rh20 presented a fue e weapon-grads .th A l e plutoniu r plutoniuo m m together with different dilutions (U-238, Th-232 usede )ar . To guarantee of reactor safety requirement properties (to ensure a large negative temperature coefficient of reactivity at all temperatures) erbiu bura s nma able poiso useds ni . The difference between spectral characteristic reactoe th f so r with pebble bed core (VGM) and the MHTGR type reactor with prismatic fuel blocks leads to more deeper Pu-239 burning in VGM reactor at the more less average burn up. The comparison of weapon-grade Pu burning effectiveness in HTGR with its utilization in the other reactor- types shows the advantag HTGRf eo s usage. It is shown that HTGR allows to provide for a very deep,more than 80 %, Pu-239 burning and fuel element arrangement features facilitate the problem of irradiated fuel storage.

INTRODUCTIO. 1 N The current works on the program of reduction in the strategic nuclear armaments will leao releast d f o elarg e stock-pile weapon-gradf so e plutoniud an A mUS primaril e th n i y Russia (abou ton0 t20 s accordin estimates)e th o gt . Storag f theso e e quantitie f o sweapon-grad e plutonium requires complicated equipmen largd tan e economic expensed an s can not be considered as a final stage of the program on nuclear armament reduction. The present paper consider problee sth reductiof mo fulr (o nl destruction) of the stock-piles of weapon-grade plutonium and its efficient application. Thernumbea s ei f wayro thif so s problemf o solving e on d ,an the possible e solutionth f o f thio se us s probleo t s i m weapon-grade plutonium as a fuel for nuclear power reactors for electricit head yan t production.

275 _I

1. reactor . fue7 l circulation system . pressur2 e vessels unit 8. small absorber balls system 3. intermediate heat exchanger . heliu9 m purification system . stea4 m generator . relie10 f valve ••"* circulatos ga . 5 r . steam-turbin11 e plant . surfas6 e cooling system

FIG. 1. VGM reactor plant

276 In the present paper the comparison of the effectiveness of weapon-grade plutonium burnin n varioui g s typf reactoro e s i s given. The investigation carried out in the USA and Russia showed that the high-temperature helium-cooled reactors are most suitable for burning the weapon-grade plutonium, where more than 80 % Pu-239 maburne b y. tup Using of coated particles with Pyc-Sic coatings as a fuel elements in HTGR facilitates the problem of irradiated fuel storage. The comparison of weapon-grade plutonium utilization in VGM reactor (Russia) and MHTGR reactor (USA) is given as well.

2. THE MAIN TECHNICAL CHARACTERISTICS OF VGM AND MHTGR. The high-temperature gas-cooled reactors (HTGR) are the advanced tren n nucleai d r power ensurin e higth g h safety characteristics and unique possibilities of using both for electricity generation and for direct use of thermal power. The main HTGR characteristic advantaged san e determinear s d by the following factors: - thermal efficiency, small heat releases to the environment, low consumption of cooling water; - high nuclear safety ensured by the inherent characteristics: chemically inert one-phase helium coolant, high heat capacit graphite th f yo e core, high stabilit fuef yo l element base fuen do l particles with multilayer coatings, high negative temperature coefficien reactivityf to ; - low specific activity of the circuit and low level of radioactivity releas environmente th o et ; - minimum excess reactivity margin (due to application of the principle of continuous refueling with spherical fuel elements or usin burf go n able poison). In Russia concepts and designs of reactor plants with HTGR have been developing for many years within a wide range of power and for various purposes, extensive RoCD efforts have been thimadr fo se direction. 2.1. VGM reactor plant with HTGR module concept HTGR. The reactor plant (fig consist) .1 : sof - the reactor with a pebble bed core including control and shutdown systems; - the vessel block consisting of the reactor vessel,the vessel for heat exchanging equipment and the cross vessel; - high temperature heat exchanger; stea- m generator; blowers ga - ; syste- m providing fuel circulation, loadin reloadind gan f go the fuel elements; - surface cooling system. Technical solutions for reactor systems provide reactor o o operation at the temperatures of 750 C or 950 C. At the operating o temperature of 750 C all the heat from the reactor is rejected in the steam generator and steam is used in the steam turbine o e operatinunitth t .A g temperatur theC 0 y 95 star f o et using intermediate heat exchanger intermediate (IHXth d )an e circuit, wher ee place b ther n dca e steam generator s turbinga , r o e

277 experimenta le demonstratiodeviceth r fo s f reactoo n r heat application. VGM reacto rfollowine systeth s mha g parameter n bracket(i s s o the value e give r ar soutlefo n t ) temperatur(seC 0 e75 f o e table.!)/!/:

Tabl. e1

Reactor thermal power 200(215) MW SG thermal power 124(217) MWt IHX thermal power t 77MW Primary circuit helium parameters: o o core inlet/outlet temperature 300 C/up to 950(750 C) pressure 5,0 MPa Intermediate circuit parameters: o o IHX outlet inlet temperature 290 C/up to 900 C pressure 5,a 2MP Steam parameters: o temperature 540 C pressure 17,2 MPa Core: height/diameter 9,4/3,0m 3 average power density 3 MW/m fuel UO 2 enrichment for U-235 8 % average burp nu 76000 MWdays/tU average fuel residence time 950 days number of reshuffling cycles of fuel 12 elements Surface cooling system: numbe independenf ro t channels thermal power at emergency cooling 1.15 MW Reactor power change rate 100-50-100 % Lifetime 40 years Fuel elements contain the fuel particles of spherical form with surrounde m diametem 5 e 0, multilaye th f ro y d b r coatinf o g pyro-carbo d siliconan n carbide. Coated particle e evenlar s y distributed within the graphite sphere of 50 mm in diameter. Graphite sphere containg fuel matrix surrounded with the graphite thicknesn i m m laye5 sf r o (see. fig. .2)

2.2. Main characteristics and features of MHTGR (USA). The technical solutions of the design were mainly based on the domestic experience of designing, construction and operation Peace oth f h Botto Ford man t St.Vrain reactors. The main characteristics of the 450 MWt MHTGR design are listed in Table 2. Standard refractory of coated particle fuel, as shown in MHTGRe figur th uses i n .di , e 3 Plutonium applicatiow utilizatione HTGR.a e t th no r . s nfo ni Highly enriched, (88 % fissile) coated, plutonium oxide fuel particles were designed, fabricated and irradiated in the Peach

278 Fuel element (graphiie binding by pyrocarbon)

Cooled fuel particle Graphit t shell Layers densityto» PyC Fuel compact Fuel kernel high density PyC

PyC+SiC

Coated fuel particle PyC+S>C

Fuel element (pressurised carbonization)

Coated fuel particle Layer* Graph lie stteli densitylov PyC Fuel kernel Fvel compact high density PyC

Coated fuel particle

FIG.2.

Bottom HTGR projec late th e n earl196td i an 0 ys e tes197Th t. 0s fueirradiates lwa excesn i bur o dt s 70000f so nup 0 MWd/t. Similar test n Europi sf o plutoniue m particles manufacturey b d Belgonucleair e Dragoth r n fo eProjec t also yielded similar successful results.

279 Tabl. e2

Power t VW 0 45 Temperature helium o inlet of the core 288 C O outlet of the core 705 C Helium pressure a MP 7,05 Fuel cycle LEU enrichment("2% 0 ) Numbe fuef ro l columns 84 3 Power density MW/6 m Reactor vessel diameter 23,7 Ft Electrical power 173 MW Efficiency 38,4 %

3. NEUTRO PHYSICAD NAN L ASPECTF O S WEAPONS-GRADE PLUTONIUM UTILIZATION Any fissile material (U-235, Pu-239, U-233) both in separate particle mixer so d with fertile materials (U-238, e Th-232b n ca ) fuea use HTGRn s li da HTGRe .Th vere sar y flexible with respect to the fuel cycle. The ratio of the number of graphite nuclei to tha fissilf to e material nuclei, determin neutroe th s n spectrum HTGRe inth . Amoun fertilf to e materia degree core th th e d f eo lan heterogeneit variee b n dyca withi widna e range changin quantite gfissilth e th f yo e (fertile) materia d fuean ll particle size, whic difficule s othehi th case rth f e o n ttypi e reactors. In estimation of the neutfonic characteristics of the VGM reactor with weapons-grade plutonium loaded inte fueoth l elements some of versions were considered: use of pure plutonium without dilution, plutonium diluted with Uranium-238 (initial enrichment with Pu - 21 % and 10 %), and Thorium-232 (initial enrichment with Pu - 10 %).In all versions the once-through fuel cycle (without chemical processing) was assumed. The analysis of the results obtained showed that the initial loading 0,5 g of Pu-239 per a fuel element is optimal taking into account assuranc e necessarth f eo ye controth wort f o hl rods located outside the core (in the side reflector). The parameter M fue VG f o l weapons-grad f o witse us h e plutoniu variour mfo s version s applicatioit f so n comparisoi n n with the design for the uranium fuel are presented in table 3. For comparison this table also list e similasth r datr Americafo a n reactor MHTGR obtaine followt G.Ay I db . s froe analysimth e th f so tabl e3 dat a thaM e effectivelreactob VG t n ca r y r usefo d destroying of weapons-grade plutonium simultaneously with electricity generation. At the VGM thermal power of 200 MW the annual output of one reactor is (»? s 0,8) 615 MW • yr, with about 260 kg of the weapons-grade plutonium destroyed within the same period. Therefor r destroyinfo e e g weapons-gradth 50ton f o s e plutonium ove periora year0 4 f s do reactorM abouVG 8 t1 s wile b l needed with the electricity output of 45 GWe-yrs. The HTGR reactor fuelled with plutonium-239 has the positive temperature effect, whic undesirabls hi e froe safetmth y poinf o t view assuro .T negative eth e temperature effec e reactotth r fuel elements must be added with poisons having resonances near the

280 Table3 Fuel parameterweapons-gradf o ReactoM e VG us r f so rfo e plutonium Valu Parametef eo r initial Option Option Optio Option! n option 1 2 3 4 MHTGR VGM (USA) 235 Pu 21% % P10 u Pu 10% U 8% Pu , . 238 238 238 232 /2/ U 92% U 79% U 90% TH 90% Charg eg/MWd, t 235 U 0,94 - - - - - 239 Pu - 1,27 1,0 1,10 1,0 1,68 240 Pu - 0,08 0,06 0,07 0,06 0,11 PUZ — 1,35 1,06 1,17 1,06 1,79 239 Pu enrichment of Puz % - 94 94 94 94 94 Discharge ,g/MWdt 235 U 0,128 - - - - - 233 U - - - - 0.100 - 239 Pu 0,056 0,090 0,077 0,178 0,050 0,174 240 . Pu 0,045 0,081 0,100 0,146 0,090 0,23 241 Pu 0,024 0,102 0,100 0,151 0,090 0,25 242 Pu 0,020 0,086 0,077 0,065 0,070 0,04 Pu t 0,145 0,359 0,354 0,540 0,300 0,70 239 Pu content of PuE % 38 25 22 33 17 25 Fuel burn up, GWD/t 78 730 210 90 100 560 loadinu P g kg/y =0,8^ r( ) — 75 60 65 60 215 Destructiof no 239 Pu , % - 93 92 84 95 90 Pu destruction total % 73-»- 67- -*• 54— — 2 7 -* — v 60 t thermal regio f neutrono n spectrum (erbiu r example)fo m e Th . calculations showed that in insertion of 0,1 g of Er into each VGM spherical fuel element the negative reactivity coefficient of -5 - 7.10 I/degre s attainee i operatine th t a d g temperature, reactivity margin in this case is about 7 % Ak/k at room temperature. Although the delayed neutron fraction for Pu-239 is essentially lower than thaU-23r ß'tfftfo ( 5 =0,0021 instead 0,0065) it is expected that the slow thermal response of the core, the single-phase coolant and the large margin between the normal o and limiting fuel temperature ("40 ensur) 0 reactoe eth r safety.

281 FUEL PARTICLES PRISMATIC FUEL ELEMENT

FUEL COMPACT

FISSILE (URANIUM <20|% ENRICHED)

FERTILE (NATURAL URANIUM) FIG. GT-MHR3. fuel

4. THE COMPARISON OF THE EFFICIENCY OF WEAPONS-GRADE PLUTONIUM DESTRUCTION IN POWER REACTORS OF THE DIFFERENT TYPES. Tht fractione e f Pu-23o n 9 destroye t burneda r NPR's depends on the following factors: - fuel burn up (i.e. the fraction of fission products relativ o fissilet e atom l heaval s (fifao yt nuclider o ) s (fima); e impacth f radiatio- o t n captur e e th fuelosse l n i s 9 99999 characterized by

282 contene Th u P f to in irradiated fues reducei l d from 9(fres4% h WPu: to ) VGM- % ; 5 2 - - 40 % - WER; . BN - % 5 8 -

Table 4. Comparison of WPu destruction efficiency in various power reactors NPR ' s type Character i sties (Discharge fuel) VGM WER-1000 BN-1600 238 U , kg/t - 945,6 775,6 239 kg/, t u P 66,7 8,20 104,60 240 Pu 60,0 6,55 17,33 241 PU 75,5 3,45 1,09 242 Pu 63,7 1,10 0,06 Pu kg/t 265,3 19,30 123,08 Fission Products, kg/t 734,0 33,80 99,85 239 Pu destructio% n, 93,0 79,5 12,8 Cgeneratioo- electricitf no burneru WP t ya s NPR's (for 80 % capacity factor) during 40 years

Power, GW(t), 18X0,2 2 X3,225 5 X4,3 Net efficiency, % 38,4 31 37,2 Power, GW(e) 1,382 2 ,o a,o Fuel burnup: fim% a 73 3,4 10,0 % fifa 77,6 77,3 83,3 Co-generation of electricity,GW(e)- year o n , u P t 0 5 r pe 45,0 65-,o ,©5 •25 recycling i

CONCLUSION 1. In case of the fuel cycle with the single-act irradiation of WPu (without recycling) the module VGM burner-reactor possesses the advantage: - in Pu-239 incineration efficiency per one cycle (the burn reacy upma h 730000 MWD/tPu),it remove irradiatee sth df o fue t lou sphere th probablf eo e military use; e higheefficiencs th i R witM - NP rVG h tha f o yn thar tfo some other types of reactors-incinérators due to VGM maximum electricity productio burninn i n givea f go n amoun weaponf to s- grade plutonium. e developeTh 2. d technolog f o multi-layey r coatings provides tightness of particles and the absence of chemical

283 interaction between the fission products and the structural materials. It facilitates of solving the problem of irradiated fuel storage.

REFERENCES

Cooles Ga e dStat. th 1 Reacto f eo r investigation n Russiasi . Safety Concepts and Basis Design Characteristics V.N.Grebennik, I.S.Mosevitskii, A.I.Kiryushin, Yu.P.Sukharev, report for 11 IWGM, Vienna, 19-22 april 1993, IAEA-TC-389.35. e Modula2Th . e Destructiorth HTGr fo Rf o Surplun s Weapons-Grade Plutonium, GA - communication 10 december 1992.

284 LIST OF PARTICIPANTS

BELGIUM

Bairiot, H. (Co-chairman) FEX Lijsterdreef 24 B-2400 Mol

CANADA

Meneley, D.A. AECL CANDU 2251 Speakman Drive Mississaug2 B K aL5

FRANCE

Gamier, J-C. CEA/CEN-Cadarache Directio Reacteurs nde s Nucléaires B.P. 211 13108 Saint-Paul-lez-Durance Cedex

Lecocq, A.B.J. EURiWA Rue des Fauvette 27 F-91400 Orsay

Rehbinder, S. NUSYS Rue Christope Colomb 9 F-75008 Paris

GERMANY

Krüger, W. BfS Berli n- Karlshors t KT3

ITALY

Landeyro, P.A. C.R.E. CASACCIA Via Anguillares1 e30 1-00100 Rome

Lombard!. C , Dept f Nucleao . r Engineering Politecnic i Milanod o Ponzio 84/3 1-20138 Milano

Mazzola. A , Dept f Nucleao . r Engineering Politecnico di Milano S. Donato (Mi) 20097

JAPAN

Furukawa. K , Tokai University Kitakaname 1117 Hiratsuka, Kanagawa 259-12

285 Ishikawa, M. Power Reacto Nuclea& r r Fuel Development Corporation Narita 4002 Oharai-machi 311-13

Mitachi. K , Toyoshashi Universit f Technologyo y Tempaku-cho 1-1 Toyohashi 441

Muromura, T. Japan Atomic Energy Reseach Institute Tokai-mura, Ibaraki-ken 319-11

Yamashita. K , Japan Atomic Energy Reseach Institute Naritachyo 3607 Ohairai-machi 311-13

RUSSIAN FEDERATfON

Baiburin, G.G. All-Russian Scientifi Researcd can h Institute of Inorganic Materials (ARSRIIM) Rogov 5a 1 23479 Moscow

Belov, S.B. Experimental Design Bureaf uo Machine Building (EDBMB) Burnakovsky proezd 15 6036 . Novgoro3N d

Deküsar, V.M. Institute of Physics & Power Engineering (IPPE) 249020 Bondarenk1 . osz Obninsk, Kaluga Region

Ivanov, E.A. Institute of Physics & Power Engineering (IPPE) 249020 Bondarenko sz. 1 Obninsk, Kaluga Region

Kagramanian, V.S. Institute of Physics & Power Engineering (IPPE) (Scientific Co-ordinator) 249020 Bondarenk1 . osz Obninsk, Kaluga Region

Kazaritsky, V.D. Institut Theoreticar efo Experimental& l Physics 25, B. Cheremushkinskaya 117259 Moscow

Krivitski, I.Y. Institut f Physiceo Powes& r Engineering (IPPE) 249020 Bondarenko sz. 1 Obninsk, Kaluga Region

Koshkin. I , Ministry of Atomic Energy of the Russian Federation (MINATOM) Staromonetny pereulok 26 109180 Moscow

286 Kovalevski Researc Developmend han t Institute f Poweo r Engineering (RDIPE) P.O. Box 788 Moscow 101000, Russia

Kudryavsev. E , Ministr f Atomiyo c Energe th f yo Russian Federation (MINATOM) Staromonetny pereulok 26 109180 Moscow

Murogov, V.M. (Chairman) Institut f Physiceo Powes& r Engineering (IPPE) 249020 Bondarenko sz. 1 Obninsk, Kaluga Region

Naumov, V.S. Research Institut f Atomieo c Reactors (RIAR) 433510 Dimitrovgrad, Ulyanovsk Region

Pshakin, G. Institute of Physics & Power Engineering (IPPE) (Technical and Local Co-ordinator) 249020 Bondarenko sz. 1 Obninsk, Kaluga Region

Rogozkin, B.D. All-Russian Scientifi Researcd can h Institute of Inorganic Materials (ARSRIIM) Rogova 5a 123479 Moscow

Smetanin, E.Ya. Institute of Physics & Power Engineering (IPPE) 249020 Bondarenk1 . osz Obninsk, Kaluga Region

Sukharev, Yu. P. Experimental Design Bureau of Machine Building (EDBMB) Burnakovsky proezd 15 6036 . Novgoro3N d

Zabudko, L.M. Institut f Physiceo Powes& r Engineering (IPPE) 249020 Bondarenko sz. 1 Obninsk, Kaluga Region

Ziabletzev, D.N. Institut f Physiceo Powes& r Engineering (IPPE) 249020 Bondarenk1 . osz Obninsk, Kaluga Region

SWITZERLAND

Chawla, R. Paul Scherrer Institute 5232-Villigen PSI

UNITED KINGDOM

Page, R.J. British Nuclear Fuels pic Risley Warrington, Cheshire WA3 6AS

287 USA

Alberstein, D. General Atomics 3550 General Atomics Court San Diego, CA 9212

ORGANIZATIONS

Magill, J. CEC/Joint Research Centre Institute for Transuranium Elements Postfach 2340 76125 Karlsruhe Germany

Sukhanov, G. International Atomic Energy Agency (Scientific Secretary) Wagramerstrasse5 0 10 x P.OBo . A-1140 Vienna Austria

Rinejski, A. International Atomic Energy Agency (Co-Scientific Secretary) Wagramerstrasse 5 P.O. Box 100 A-1 140 Vienna Austria

Takamatsu. M , International Atomic Energy Agency Wagramerstrasse5 0 10 x P.OBo . A-1140 Vienna Austria

Zaribas, N. OECD/NEA Boulevard des lies 12 F-92130 Issy-les-Moulineaux

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