Control Rods Designsand of Shutdown Systems Advancedfor Liquid Metal Fast Reactors

IAEA-TECDOC-884

Absorber materials, control rods designsand of shutdown systems advancedfor liquid metal fast reactors

Proceeding of a Technical Committee meeting held in Obninsk, Russian Federation, 3-7 July 1995

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Absorber materials, control rods and designs of shutdown systems for advanced liquid metal fast reactors

Proceeding Technicala of Committee meeting held in Obninsk, Russian Federation, 3-7 July 1995

^±iL^ n n\ INTERNATIONAL ATOMIC ENERGY AGENCY The originating Section of this publication in the IAEA was: Nuclear Power Technology Development Section International Atomic Energy Agency Wagramerstrasse 5 P.O. Box 100 A-1400 Vienna, Austria

ABSORBER MATERIALS, CONTROL ROD DESIGND SAN F SO SHUTDOWN SYSTEMS FOR ADVANCED LIQUID METAL FAST REACTORS IAEA, VIENNA, 1996 IAEA-TECDOC-884 ISSN 1011-4289 © IAEA, 1996 Printe IAEe th AustriAn y i d b a June 1996 PLEASE BE AWARE THAT MISSINE TH AL F LO G PAGE THIN SI S DOCUMENT WERE ORIGINALLY BLANK FOREWORD

safee Th , reliabl economid ean c operatio nucleaf no r reactors depends upo reliable nth e operation of the reactor shutdown and regulation systems. In liquid metal cooled fast reactors (LMFRs), these consis f controo t l assemblies, drive mechanisms, guide tube d othean s r structural components maie Th n. elemen controe th f o t l assembl controe th s yi l rod, which consists of a neutron absorbing material and is (usually) clad in stainless steel. Neutron absorbing material developmenn si t includ ceramie eth c boron carbid europiud ean m oxide. currene Th t control assembly designs have worked well recent bu , t developments have shown that their lifetime in the core and reliability could be increased by design improvements.

Fast reactor R&D programmes in some countries reflect the effort to enhance the flexibility to manage plutonium stockpiles. Studies performed recently have demonstrated that U using the natural absorber B4C in combination with moderator B4C could improve the safety of fast reactors in the plutonium burner mode of operation.

Owin improvementgo t reliabilite th n si operatiof yo reactivitf no y shutdow decad nan y heat removal systems, core melting accident highle sar y improbable firse tTh . ste preveno pt t core melting is to be able to shutdown the reactor under all probable circumstances. For advanced fast reactors there are two independent reactor shutdown systems: a primary conventional active system and a backup passive system. At present, there are many proposed passive system insertioe th r sfo f contro no l rods intcoree oth .

The IWGFR (International Working Group on Fast Reactors) proposed that the IAEA organize a Technical Committee Meeting on absorber materials, control rods and designs of backup reactivity shutdown systems for breakeven cores and burner cores for reducing plutonium stockpiles, to review operational experience with absorber elements and control rods and the development of passive backup shutdown systems for advanced LMFRs.

thire th Thi ds IAEswa A meetin subjecge helth n absorbef do o t r material controd san l rod r LMFRsfo s othee meetingo Th . tw r s were hel Dimitrovgran di d (1973 Obninskd an ) , (1983), in the Former Soviet Union. 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 States nominating ofthe or organizations. Throughout textthe names Memberof States retainedare theyas were when textthe was compiled. Theof use particular designations of countries territoriesor does imply judgementnot any by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions delimitationthe of or theirof boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does implyintentionnot any infringeto proprietary rights, should construednor be it an as endorsement recommendationor partthe IAEA. ofon the The authors are responsible for having obtained the necessary permission for the IAEA to reproduce, translate or use material from sources already protected by copyrights. CONTENTS

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

Revie technicaf wo l approache solutiond san LMFr sfo R control rods developmen1 1 . . t V.I Matveev, A.P. Ivanov, R.M. Voznesenski V.P. Evdokimov Developmen improvemend an t f controo t BN-35e l th rod r BN-60d sfo 0 an 0 reactor9 1 . s A.I. Efremov, V.IRyakhovskikh, V.B. Ponomarenko, V.M. Chernyshov Experience of the BN-600 reactor control rods development ...... 33 Yu.K. Alexandrov, B.A. Vasilyev, S.A. Iskhakov, O.B. Mishin, V.A. Rogov, A.S. Shabalin Control assembl usee b CEF n di o y t R ...... 7 4 . Xie Guangshan, Zhang Ruxian, Wang Yonglan, ShikunLi The control rod modelling code REGAIN ...... 53 TruffertJ. Experience with control rod drive mechanism of FBTR ...... 61 P.V. Ramalingam, M.A.K. lyear, Veerasamy,R. S.K. Gupta, Soosainathan,L. V. Raj an Bab u Developmen passivf o t e shut-down systemEuropeae th r sfo n Fast Reacto ...R 9 6 EF r. M. Edelmann, G. Kussmaul, W. Väth Design philosophy of PFBR shutdown systems ...... 81 V. Rajan Babu, Vijayashree,R. Govindarajan,S. Vaidyanathan,G. G. Muralikrishna, T.K. Shanmugam, S.C. Chetal, D. Raghavan, S.B. Bhoje Desig shutdowf no n syste PFBr mfo R ...... 9 8 . V. Rajan Babu, R. Vijayashree, R. Veerasamy, L. Soosainathan, S. Govindarajan, S.M. Lee, S.C. Chetal Developmen passivf o t e safety device sodium-cooler sfo d fast reactors ...... 7 9 . Yu.E. Bagdasarov, Yu.K. Buksha, R.M. Voznesnski, N.V. Vyunnikov, V.P. Kornilov, N.D. Krayev, L.I. Mamayev, A.G. Portyanoy, A.P. Sorokin Main feature BN-80e th f o s 0 passive shutdown rods ...... 7 10 . Yu.K. Alexandrov, V.A. Rogov, A.S. Shabalin The desig backua f no p reactor shutdown syste DFBf mo R ...... 3 11 . K. Okada, K. Tarutani Y. Shibata, M. Ueta, T. Inagaki Irradiation performance Superphenie th f o s x type absorber element ...... 7 12 . B. Kryger, D. Gösset, J.M. Escleine Operation experienc BN-60e th f eo 0 reactor control rods ...... 1 14 . Maltsev,V.V. V.F. Rosfyakov, Babenko,V. G. A.N. Ogorodov The experience of post irradiation investigations of the BN-600 control rods ...... 153 V.P. Tarasikov, R.M. Voznesenski, V.A. Rudenko Irradiation behavio borof o r n carbide neutron absorber ...... 161 T. Kaito, T. Maruyama, S. Onose, T. Horiuchi Third shutdown level for EFR project ...... 173 D. Favet, B. Carluec, S. Dechelette Improvement of the performances of the "poisoned" CAPRA core by use of "B4C . . 189 G. Gastaldo, G. Vambenepe, J.C. Gamier Experienc developmentn ei , operating materiad an , l investigatio BOR-6e th f no 0 reactor control and safety rods ...... 195 V.B. Ponomarenko, A.I. Efremov, G.I. Gadzhiev, V.D. Risovany, A.V. Zakharov, T.M. Guseva Productio gamma-sourcesf no , base europiun do m oxid fasn ei t reactors ...... 5 20 . V.D. Risovany, A.V. Zakharov, E.P. Klochkov, T.M. Guseva, V.B. Ponomarenko,V.M. Chernyshov Reprocessin irradiatee th f go d boron carbide enriche boron-1e th y db 0 isotope and its reuse in the control rods of the fast breeder reactors ...... 219 V.D. Risovany, A.V. Zakharov, E.P. Klochkov, A.G. Osipenko, N.S. Kosulin, G.I. Mikhailichenko Experienc production ei f articleno s from boron carbid r fasefo t reactor control 5 rod22 s I.A. Bairamashvili Diving-bell and double-vented B4C control rod pin ...... 231 Shikun,Li Changshan,Z. YingxianXu

Lis Participantf o t s ...... 5 24 . SUMMARY

. 1 INTRODUCTION

e IAETh A Technical Committee Meeting (TCM Absorben o ) r Materials, Control Rods and Designs of Backup Reactivity Shutdown Systems for Breakeven Cores and Burner Cores for Reducing Plutonium Stockpiles, was hosted by the Institute of Physics and Power Engineering (IPPE), Obninsk, Russian Federation, from 3 to 7 July 1995. This was the third IAEA meeting held on the subject of absorber materials and control rods for liquid metal fast reactors (LMFRs). The other two meetings were held in Dimitrovgrad (1973) and Obninsk (1983), in the Former Soviet Union.

Thirty-five specialists from France, Germany, India, Japan e Republith , f o c Kazakhsan e Russiath , n Federatio e Republith d nan f Georgi o c a (observer) attendee dth meeting.

The meeting had seven sessions. The sessions were chaired by B. Kryger (France), R. Babu (India) . EdelmaM , n (Germany) . OkadK , a (Japan) . KaitT , o (Japan) . FaveD , t (France), V. Risovany (Russian Federation), J.P. Truffert (France) and I. Bairamashvili (Republi f Georgia)co .

The main topics of discussions were:

• status of control rod designs for fast reactors and experience with operation, • properties and behaviour of absorber materials for control rods; results of post- irradiation examinatio f absorbeo n r materials d mechanisman , s affecting their properties and behaviour, • desig backua f no p reactivity shutdown system utilizing passive mechanisms: Curie point electromagnetic mechanism, enhancement of thermal expansion of absorber rod drive lines, hydraulically suspended control rods, s expansioga n module coree d th an n ,i s possibilite th optimizinf yo reactivite gth y coefficient efficience th u d P an sf yo burning by using absorber and moderator materials in the core.

A paper 3 tota2 f o l s were presented technicaa d an , IPPe l touth Ef ro also took place. 2. DISCUSSION AND CONCLUSIONS

2.1. Status of control rod designs for fast reactors and experience with operation

Significant experience, accumulate absorben do r material fasr sfo t reactor control rods, presentes wa meetinge th t da . Boron carbid higf eo h enrichment (45-90 vented an %10f Bd)o absorber pins (with sodium bonding) have been successfully Phénix e useth n di , BN-350- BN , Superphénid an 0 60 x (SPX) reactors controe Th . lifetim d BN-60e ro lth r efo 0 reacto0 50 s ri effective days (e.d.) lifetime th ; reactoX e predicteSP e re th teste (baseth r th n i sdfo n do Phénix reactor e.d0 . 64 s Boro)i n carbid favouree th s ei d absorber material. Europium oxide (Eu2O3) appeared to be an alternative to boron carbide of natural enrichment.

Investigation of the properties and behaviour of B4C have been carried out in the BN- 350, BN-600, BOP-60, Phénix and JOYO reactors for a broad range of conditions affecting endurances it . Data from these studies define behavioue dth irradiatef ro d absorbing materials (both pins and the whole rods) provide the basis for improved control rod designs and extended control rod lifetimes.

Swellin irradiatef go d boro maie th nnf o carbidfactor e on s sei limitin lifetime gth f eo the rods. The dependence of the boron carbide swelling rate (Ad/d%/B%) on the burn up was determined from post-radiation studie controf so l rods fro BN-60e mth 0 reactor. After about 500 e.d. the boron burn up was -18%. A minimum swelling rate of -0.1% for enriched boron carbid d 0.2-0.4ean r natura%fo l boron carbid s observeewa temperatura t da f o e 500°C. At a working temperature of about 800°C, the swelling rate is -1.0% per percent burnup for natural boron carbide and 0.6% per percent burnup of enriched boron. Some reserve remained since the gap between the absorber and cladding was not filled completely. Under these conditions mechanical interaction of the cladding was not observed, and the cladding material retained a sufficient reserve strength.

A series of tests were carried out in the Phénix reactor irradiating an enriched boron e.d0 24 . carbid r Thesfo d eero experiments showed tha in-pile tth e residencn pi ee timth f eo is limited by the mechanical interaction of boron carbide with the cladding, owing to absorber fragments enterinbetweep ga e absorbe e claddingge th nth th d an r .

Irradiated boron carbide data (swellin s releasgga ratburnupn d eo ean ) were also obtained in Japan in tests in the experimental fast reactor JOYO. Burn up of about 23% (230xl0 capt/m achieves )wa maximua t da m temperatur f 1400°Ceo dependence Th . f eo 26 2 1 pellet swellinreleass ga burnud n eo g an measureds pwa . Japanes Frencd ean h resulte th f so absorber material behaviour under irradiatio goon i e dn ar agreemen t diffe tbu som n i r e cases from Russian results t appearI . s thae differenceth t e explaineb n ca sdifferences y db n i methods of obtaining enriched boron carbide. However, the results obtained are not contradictory thed an y, improv understandine eth swelline releass th ga f go d eg an processes .

absorbinw Ne designn gpi materialrequiree w ar sne d achievo dt an s e controd ro l lifetimes of 1000 e.d. Some measures to increase the control rods lifetime for existing reactors

The abbreviations capt. or cap. stand for capture were proposed by the French. The first proposal was to reduce the capture rate in boron 10 carbide by lowering the B enrichment of the B4C pellets in the lower part of the control rods pins. The second proposal was to prevent absorber fragment relocation by providing a thin stainless steel shroud to enclose the pellet stack.

Considerable progres s beesha n Russiae madth n i e n Federatio e desigth d n ni nan manufactur rodf eo s containing absorbing (boron carbide moderatind )an g (zirconium) hybrid materials. Using this design with its lifetime of 450 e.d. would result in considerable saving of high cost enriched boron.

participante Th s agreed thamose th t t efficienutilizo t y ewa thig h enrichment boron separato t s i t froei m irradiated materials developmene Th . f technologo t reprocessinr yfo g irradiated boron carbide, which provides complete removal of radionuclides from irradiated 10 materials, was presented at the meeting. This permits repeated use of B enriched B4C in fast reactors.

It became clear after discussion thaFrence th t h contro calculatiod ro l n code REGAIN can analyse both steady state and transient pin behaviour and covers thermal, mechanical and some chemical characteristics r boroFo . n carbide material e coddean th s ca el with both sodium bonde heliud dan m bonded design concepts.

The meeting participants concluded that the problems of reliable and effective control rods for sodium cooled fast reactors are being solved.

2.2. Optimizing core safety parameters for a plutonium burner reactor using absorber and moderator materials

s noteIwa t d thaapproacheo tw t optimizo st fasea t reactor cor plutoniur efo r m(o minor actinide) burning have been considered in France: the "dilution approach" (about two thirds of the pins are fissile pins, the remaining one third are fuel free) and the "poisoning approach" (the absorber B 4s Cintroducei e diluenth n di t subassemblies) t presentA . e th , dilution approach is preferred since it allows better core safety parameters (mainly the Doppler constant). Data presente Frence th n di h paper indicated tha introductioe tth moderatoe th f no r fuee poisoneC th 4 l B n bundli e th d f coreo e strongly increase Dopplee sth r constant (+60%) and reduce sodiue sth m void reactivity worth (-20%) resula s poisonee A th t. d core option n with B similas 4Cha r plutonium burning characteristic betted san r safety parameters thae nth dilution option mais It . n advantages result fro large mth e diameter fuel pins which increase U fuel lifetime usin gstandara d f fueo B 4ls e fillinC a bundle us e g Th materia. l doet no s introduce additional technical problems. concludes wa t I d tha poisonee th t d reactor option deserves further investigation.

2.3. Advanced passive reactor shutdown systems

The meeting participants noted tha classin ti c core disruptions (unprotected los flof so w (ULOF) and unprotected transient overpower (UTOP), the reactor can be shut down only by inherent passive reactivity feedback mechanisms. Therefore, special measure provido t s e additional negative reactivit needede yar . The Russian Federation has developed a backup passive shutdown system based on hydraulically suspended absorber rods (HSAR) which drop intcore oth e whe coolane nth t flo we cor th rat en i efall s below 50%. Full scale HSARe BN-60th r BN-80d sfo 0an 0 reactors have been tested. This syste mcapabls i f shuttineo reactoe gth r dow ULOa n ni F transient.

The German syste passivr mfo e shutdow Europeae th f no n Fast Reactor (EFR) uses enhanced thermal expansio f controno drivd ro l e lines (CRDL provido t ) e automatic scram. This system forces the control rods into the core when pre-set coolant temperatures are exceeded. This device uses a hydraulic expansion module and two individual (coaxially arranged) drive lines sodiue largea Th . s mha r thermal expansion coefficient tha containee nth r material. This syste mcapabls i shuttinf eo reactoe gth r dow ULOn i UTOd Fan P transients.

The demonstration fast breeder reactor (DFBR) in Japan includes the following passive backup shutdown systems:

• gas expansion modules (GEMS), and self-actuated shutdown systems (SASS) using a Curie point magnetic alloy.

In additio GEMno t SASSd San feasibilit,a y enhancestude th f yo d thermal expansion of CRDL is in progress.

The French presentation described and evaluated a third shutdown level for the EFR project which has been developed, implementing the following four new device systems: systea m whic• h terminate poweabsorbee e sth th o t r r electromagnets afte losa r f so primary pump electric power supply, • a device for CRDL passively enhanced thermal expansion, devica e whic• h overcomes contro jammind ro l motorizey gb d insertio f absorbeno r rodsd an , mechanicaa • l stroke limitation device which passively terminate withdrawae sth a f o l faulted control rod.

The inclusio thesf no projecR e systemEF e t th lead significana n s i o st t improvement of the shutdown function.

The shutdown system of the Indian prototype fast breeder reactor (PFBR) incorporates divers redundand ean t feature included an s followine sth g systems:

• a system which terminates the power supply to the absorber electromagnets after a loss of pump power, • a gas expansion module, a Curie point magnetic switch

concludes Itwa d that wit introductioe hth thesf no e passive backup shutdown systems, future advanced LMFRs will hav verea y high degre safetyf eo .

10 REVIEW OF TECHNICAL APPROACHES AND SOLUTIONS FOR LMFR CONTROL RODS DEVELOPMENT

V.l. MATVEEV, A.P. IVANOV, R.M. VOZNESENSKI, V.P. EVDOKIMOV IPPE, Obninsk, Russian Federation

Abstract

This paper review gived san retrospectivsa e analysi technicaf so l approached san solutions used during control rods developmen r sodiufo t m cooled fast reactors. General principle fasf so t reactor contro desigd ro l n developmen absorbed an t r material selection are considered maie wels th a , ns a l results obtaine fiele f th advance do n di d control rods containing absorbe moderatod an r r materials.

INTRODUCTION

In the process of BN-350 and BN-600 fast power reactors development and operation a problem of creation of highly efficient and economical control rods was solved. This problem arose due to some peculiar features of sodium cooled fast reactor physics and technology which, as is known, reduce to as follows: - because of low neutron - absorbtion cross-sections at high energies the range of materials controsuitable th n i their le fo rod rus substantialls si y limited; - because of comparatively small fast reactor cores dimensions connected with high thermal loads there arise certain difficulties with accomodatio larga f no e numbef o r contro drivd lro e mechanisms; hig- h neutron e corfluxeth e n i s(~1 6 n/cm2sec) determine high rate f nucleao s r reactions in control rod materials resulting in high energy release and burn-up values, as well as high fluence upon structural materials. In the course of fast reactor control rods development and validation, primarily for the BN-350, BN-600 reactors and for the BN-800 reactor project, extensive studies in e fiel f th physycso d , technolog d materiaan y l research were carried out, many engineering solution approached an s s were considere f thio sverifiedm d ai dan e Th . paper is a brief presentation of these solutions and their retrospective analysis.

1. Main Principles of the Choice of Control Rods for BN - Type Reactors.

BN-type Th e reactors control syste numbea s mha f featureo r maie th sf whic no h are as follows: - general arrangement of the control system is performed on a basis of the separated- functions, principle owing to which the rods containing high- enrichment boron carbid outside ear core eth e during reactor on-power operation; - in all types of rods with boron carbide vented absorbtion elements are used. Besidesadvancee th r fo , d BN-80 d BN-1600an 0 reactor project lona r gfo s time there were developed fuel compensators for reactivity compensation due to fuel burn- . Howeverup , such typ f rodexcludes eo wa s d from subsequent developmente th f o s above reactors. The principl f controeo safetd an l y rods separation accordin theio gt r functional duty was used still at the design stage of the first power reactor BN-350 and is continued to be used in all further control and safety system developments for BN-type reactors t shoulI . notee db d that therewit functione hth somf o s combinede b rod n sca .

11 r exampleFo , ther e combinear e d safet d temperaturan y e compensation functions. However, the main aim of using this principle is fulfilled: the high enrichment boron rod undee sar r facilitated condition f operationso mose Th . t important problee th t ma development of the control and safety system rods is the choice of absorbtion material. Alread vere th y t ya earl y stag f theBN-35eo 0 reactor desig enrichee nth d boron carbide was chose r safetnfo y rodautomatid san c control rods. Late , howeveron r t varioua , s laboratories there were started a search for and studies of other absorbing materials for fasn i thei te reactorsrus maie Th .n reaso r thesnfo e studie laca f confidencs ko swa n ei boron carbide irradiation resistance prevailing at that time. As one of alternative absorbers ther choses evalidatewa d nan d CrB 2base- d materia CrB( l 2 alloye Ta)y db . This material was used in the BOR-60 reactor in 1972-1973 and showed good radiation stability. However, because of lower ( by about 20 % ) efficiency as compared to boron carbide and more complex technology it was withdrawn from production and replaced by boron carbide) absorber y , n ( . s seeme highle b o dt y attractiv fasn i r thei te efo us r reactor sThes. e materials still coul t competdno e with enriched boro theiy nb r efficiency but in those cases where - enrichied boron carbide was used they seemed to be a reasonable alternative f suco he materialOn . tantalums BN-35e wa s th r 0Fo . reactoa r tantalum developes controwa d fabricated lro dan d whic h, however t usede no s Th . wa , most serious alternative to boron carbide was europium oxide (£^03) . By its main characteristics this materia subsantialls i l y superio boroo t r n carbide: high stabilitf yo its efficienc durinn i y g reacto runa r , absenc releass ga swelling d f eo ean . This material s use burn-uwa n di p compensatio nBN-60e rodth f so 0 reactor which were successfully operated till 1988. Only one disadvantage , i. e. their high residual activity, substantially complicating spent rods handling, predetermine d passinen e gth oven i dboroo t r n carbide. In the course of the BN-type reactors design the development work on some other absorbtion materialbase f th EuB^,o e n o ( s ReBß, s carrieetc wa t which ) .ou d , howeverintroductio e t broughth no o s t p wa ,u t n stage. The general conclusion from the investigations conducted is that boron carbide despit a enumbe s disadvantageit f o r e moss th prove te b optimu o t d m absorbtion fasn i te reactorsmateriaus s it r lfo . mose Th t important elemen controe th desigf d o t ro l n primarily determinin life geth time of the rod is the absorber pin. In the control rods of the first BN-350 loading the sealed absorber rods with helium sub-layer were used disadvantagA . f thio e s type absorber rod theis si r short life time determine permissibla y db e pressur claddinn eo f go helium released from absorber during operation. The life time of such rods was 2-4 eff. months. The use of vented absorber pins with a sodium sub-layer has radically influenced upo improvemene nth f technicao t economicad an l l characteristic controe th d f so an l safety system rods. A vented absorber pin is made in form of a cladding with two end plugse uppeth n I r. plu iten ga f unsealinm o mads gi forn systea ei f mo f holes mo . Through them fillinf absorbeo p gu r phis with sodiu s releasreactoe ga th d mn i e an r int circuie oth t take place thin I . s type absorber pin probleo s n pressur s e ga th mf o n eo cladding occurs a considerabl o T . e degree a proble, f absorbemo r pressure upon claddin withdraws gi presence th s sodiuna a f eo m sublayer allow chango st size f eth e o the absorbe claddin r- constan a t a p gga t absorber temperature. Therewith an established level of the absorber temperature determines, in accordance with its temperature dependence, the swelling rates, the minimum swelling rate bein time6 g5- s less tha sealea n i d absorber pin. With a sodium sub-layer the absorber is more resistant to fragmentation; in spite of presence th f crackeo t remainsi f intacti s a falso f smal n o ,l l fragmentp s ga inte oth between absorbe claddind an r g occurs above Th . e fact allowef o abandoo d t e us e nth small-diameter absorber pins used with the aim to exclude radial cracking of pellets due

12 to temperature stresses . Passing over to " thick absorber pins" has allowed by the present tim e absorbeo unif t eth rode th y ydesignn b s pi r o improvt , e economical characteristics indexes. It shoul e notedb d thaintroductioe th t ventee th f no d absorbe design s pi r nwa precede testiny db f variougo s type items f unsealino g( hydrauli c valve, discs witha discontinuous weld, compacted metallic wire, etc.para s f experimentaa )o t l absorber BR-5e pinrodd th sf an ,s o BOR-60, BN-350 vente e reactorsth w d absorbeNo . r pine sar used in all types of control and safety boron carbide - based rods. No remark upon their operation were made. A maximum life time of rods with vented absorber pins was achieve e th f BN-60 o n effd i 0 . at .50 day~ % 0 burn-u a s 9 sreacto( 1 wa ~ d f po an r boron). Specific design of the rods is determined by their functional duty. The control and safety rods are made in form of a cylindrical multimembered structure connected by hinges. For a free passage of rods inthe guide tubes a sufficient gap between the rods and the tube, as well as an optimum length of members should be assured. The working part of the rod is made in form of an absorber pins bundle. The load-bearing element in the working member of the rods of the BN-350 first loading was the central absorber pin with a thicker cladding. Its ends were fitted with lattices in which, as in a separator, the absorber pins were assembled secone th n dI - loadin. g rod wrappee sth r tubs ewa used as a load-bearing element that allowed to avoid rods seizing in the guide tube becaus f peripherao e l absorber pins bendino adjust e conditiond th tan g f heao s t removal from absorber pins as well. At present the number of working members in experimental rods is reduced to one (at a length of ~ 1 m) that allowed to increase rods efficiency due to elimination of hinges ( in the safety rod of the BN-350 core first loading the working section consisted of three members). rode Th s efficienc alss yi o markedly affecte choice th y structuraf deb o l materialsn I . control and safety rods of the BN-600 and BN-350 reactors the following structural materials are used: absorber fo - r pin sX16H1- 5 (3H-847) type austenitic steel; - for the rest of rod elements - X18H10T type austenitic steel. At present the life time of the BN-600 reactor standart control and safety rods is set eff5 alimites 36 .ti day d structuray san db l material (X18H10T) performancee th n I . BN-600 experimental contro d safetan l y rods more radiation-resistant materiale ar s used: for absorber pins - Xl6Hl5+Ti c.d. austenitic steel; for the rest of rod elements IXI3M2BOP-type ferrite steel e abovTh . e rods were operateeff0 50 . r dfo days up to a dose of 70 dpa. At present these materials have been operated in fuel a thasubassembliedp t 0 validate10 o t p s u sthei r performanc s par a ef contro o t d an l safety rods durin eff0 g.70 days t shoulI . notee db d tha contron i t safetd an l y rode sth same structural material fueusen e i sar s l d a subassemblies . Tabln I mai1 e n characteristic e BN-350th f o s , BN-60 d BN-800an 0 controd an l safety rod presentede sar .

. Desig2 n Developmen Controf o t l Rods Containing Absorbe Moderatod ran r Materials (rod-traps). As it was mentioned above, the experience of fast reactor development shows that expensive absorber usereactomateriale e b t th leas d a r o t fo tre controsar l rods. High absorbee th cos f o t r hig e baseth h n enrichmendo t boron resultin hige th h n gi cos f o t the control rod, stimulate searce d th technica r hfo l solutions that would allow replacing high enrichment boron by other materials or decreasing its amount in the control rods. The most attractive way of this problem solution is using of so called rod-traps, containing absorbe moderatod ran r materials moderatoe .Th r softens neutron spectrum in the absorber section of the rod, resulting in the increase of absorber neutron capture

13 Table 1 Main Characteristics of BN-type reactors control and safety rods.

Reactor BN-350 BN-600 BN-800 BN-600 .(test rods)

Rods SR TC RR CR SR SR-L RR CR SR SR-L RR CR SR SR-L RR CR Total 12 27 3'0 4 numbef ro rods The number 3 1 2 6 5 1 2 19 9 3 2 16 1 1 1 1 of various- type rods Absorber B4C B4C B4C Fuel B4C B4C EU2Û3 B4C B4C B4C B4C B4C B4C B4C B4C B4C material 60-80% 60% 60% 80% nat. nat. 92% nat. nat. nat. 92% nat. nat. nat. (B-10) (B-'°) (B-10) (B-10) (B-10) (B-10) Absorber 2.545 2.100 0.225 1.853 0.455 6.2 0.595 3.06 3.06 0.56 2.0 2.38 0.32 0.17 0.32 quantity (B-10 and EuaOj), kg Average 1.10 1.00 0.20 0.45 0.20 0.28 0.27 0.47 0.47 0.21 0.4 0.6 0.2 0.17 0.38 efficiencf yo a singld ero (%Ak/k Absorber V V V - V V s V V V V V V V V V pin type Specified 390 390 240 365 365 365 365 280 280 280 280 550 550 550 550 life-time (eff. days) Notation used in the table: SR - safety rods TC - temperature compensators SR-L - saefty rods (loop-type) CR - fuel burn-up compensation rods RR - regulating rods (v) - vented absorber pins; (s) - sealed absorber pins rate, i. e. its efficiency. The calculations show that the efficiency of natural absorber material considerable b n sca y increase thin di s type rods. Analytical and experimental studies have proved that the highest rod efficiency can be achieved by combining absorbers on the base of boron carbide or europium oxide with mmaterial hydride-based moderators. First technical proposal on this approach to the fast reactor control rod design was developed in the USSR in 1969. This proposal e developmen th wae bas r th s e BN-35fo th f o t0 contro d designro le workinTh . g section of the rod contained the annular design absorber element made of europium oxide wit stainlese hth s steel cladding. Inside this element, seven moderator pins were located, made of zirconium hydride with the stainless steel cladding. This rod was in operation in the BN-350 reactor during about 300 effective days in 1980-1981. According to the results of in-reactor measurements, the efficiency of this rod-trap somewhae b turneo t t dou t lower tha regulane thath desigf d o tro r n wit% h 0 B8 4f Co enrichment. Enriched boron carbide was used for the further design development of the rod- trap. Accordin calculatioe th o gt n results thin i , s case B4C amoun controe th d n i tro l can be several times reduced as compared to that of regular rod design. This can result d efficiencyaboun i % ro increas 0 e 1 th t f .eo Various designe rod-trath f o s p with enriched boron on the periphery and zirconium hydride in the central part were tested at the BFS critical facility and in the BN-350 and BN-600 reactors. Several rod-trap designs with the outer absorber in the form of an annular element f Bo 4t Cse pinfore a th f ms o n wit i d h an boron-10 isotopd ean enrichmen0 8 , 60 f o t 9wer2% e tested withi systee nth f scramo m BN-35e rodth n BN-60d si 0an 0 reactors. These rod-traps wer operation eeffi 0 claimo .40 n days o t d reactorn 0 ni san , 15 r sfo were brought from the reactor personnel on their performance. Main data on experimental rod-traps tests in the BN-350 and BN-600 reactors are presented in Table 2.

Table2 Main data on experimental rod-traps tests in the BN-350 and BN-600 reactors

JVs Reactor typd ro ,e Absorber Operation life Time of tests time, eff. days 1 BN-350, fuel compen- Eu2O3 300 1980-81 sator, annular type 2 Bn-350, safety rod, B4 C(9) 2% 230 1987-88 typn pi e 3 Bn-350, safety rod, B4C(92%) 400 1989-91 pin type 4 Bn-350, safety rod, B4 C(92 %) 450 in operation pin type since 1994 5 Bn-600, safety rod, B4C(60%)- 311 1984-85 pin type lower member B4 C(8- ) 0% upper member 6 Bn-600, safety rod, B4C(80%) 311 1989-90 pin type 7 Bn-600, safety rod, B4C(80%) 348 1990-91 pin type 8 Bn-600, safety rod, B4 C(8) 0% 320 1991-92 pin type

15 Physical processes in the rod-trap have rather a complicated pattern for the descriotion of wich sufficiently precise calculation methods should be used. At choosing a design of such a rod the optimization of the absorber and moderator materials ratio is required. As an example, some results of pin-type rod-traps calculations for the BN-600 reactor are presented. Europium oxide was used as absorber, zirconium hydrige-as moderator. calculatione Inth s various version f trapo s s were used: direct (moderator inside the rod), reverse (moderator on the outside), homogeneous ones-placed in core th e centre e calculatioTh . ncylindea corrode e s modeth th e wa sf f o o rl height and with a radius of =5.36 cm. For direct and reverse versions the cylinder was two-zoned ( the zones differing by their material). In the calculation variant dimensione th s innee d werth ro f r e o se zonevariedth f o s , with material density in them being retained. In a homogeneous version the absorber-moderator ratiaccordancn i s oi e wit direce hth t versio wicn ni h this ratio changed at the expense of the inner zone radius variation. Besides, for the direct versio alsoutsids e noth wa variedd ero radiue th . f so calculatione Th s wer multigroue eth carriey b t dou p diffusion codese Th . results are presented in Figl. As it follows from these data, the direct trap is most efficient: its efficiency is 30 % higher than that of arod consisting of the

0,4

0,2

direc1- t trap, 2 - homogeneous trap, 3 - revers trap 4moderator-fre- e Fig. 1. Various-type trap-rods efficiency as a function of abcorber fraction in the rod (s =

16 moderato e rsam th alon et (a evolum e fractions) e rod-traTh . p with homogeneous composition loses in efficiency by about 5 %. The reverse trap "loses" in efficiency by about 25 %. At an increase of the outside radius of the rod the gain in efficiency as compared with a rod consisting of moderator alone increases radiud ro a st :a increas abouy eb . time2 t% gai0 e 5 s th s ni f investigationo t se A validatioe th s rod-trae carrien o th t f no dou p designs give good prospects for their use instead of expensive rods with enriched boron carbide.

Developmene Th . 3 Fuef o t l Compensators.

At the first stage of the BN-800 and BN-1600 reactors design it was expected fuee us l compensatoro t r burn-usfo p compensation. Such solutio s causenwa d by an aim to decrease the control and safety system effect upon the reactor breeding characteristics t usin A .absorbine th g g compensatorse th n i , g. . e , BN-600 reactor, total decreas breedine th f o e g rati e s i ob 0.05-0.0 n ca t 6(i roughly considered that compensation of 1 % Ak/k leads to aloss in the breeding f fueo le us compensatorratiy 0.02) b e o Th . s would reduce these losses practically to zero. At the same time it is known that the main disadvantage of traditional-type fuel compensator BN-350n i , g. s . e (asconnectes i ), d with their w efficiency.Thereforelo , various proposals aime t a increasind e fueth gl compensator efficiency were consideredproposale th f o e connectes si On . d with using uranium metal, having densit timey2 s higher than tha f uraniuo t m oxide, in the absorbing section of the compensators. The development and implementation of this proposal will allow to increase compensators efficiency by about 30 %. Another proposal is related with the use of somewhat unusual composition, - uranium-238-free material, - in the fuel section. It is known that the positive reactivity of a fuel composition is the difference between the positive reactivity of fissionable material and the negative reactivity of raw material. Henc t followi e s o materiaremovw t tha ra f i te th el (uranium-238) thee th n efficiency of the remaining fissionable part increases substantially. Let us note, thaimpossibls i t i t increaso et e efficiency simpl expense th t ya f increasineo e gth amount of fissionable material, because of the limitation from the viewpoint of heat removal. At the same time, uranium-238-free fuel material in the process of irradiation loses its efficiency much faster than conventional fuel material, as there is no compensation due to fuel production from raw material. Therefore, the comparisonof these two type compensators efficiency should be carried out with account of their burn-up. Fuel compensators (or burn-up compensators) are two-member constructions consisting of the absorbing and fuel sections their lengths being approximately equal to the core height. The fuel element of the upper absorbing section contain materiaw sra l (U-238). Fuel element bottoe th f so m fuel section are similar to those of the core. In Russian fast reactors both absorbing control rods and fuel compensators are of a cylindrical form. This determines the round form of hole in the hexagonal tube, wich is inserted into the cell to be used for contro safetd an l y rods. Thus core th ,e cell wher fuee eth l compensato locates i r d is characterize additionay db l "dilution" with steesodiud an l m comparee th o dt conventional fuel subassembly that results in some increase of breeding ratio (BR), thoug larga t ehno one. Actually fuee th l, compensator t reducno e o sd eth BR, in contrast to the absorbing rods. Compared to these rods, the fuel compensator (at equal efficiency) renders a smaller effect on the energy release fields. This is caused by oppositely directed perturbing effects of the absorbing fued an l sections course th n f burn-uI .eo p compensation, fuel compensatore sar

17 inserted int core oth e fro bottoe mth m upwards; this kin f motiodo s beenha n practiced fro e viewpoinmth f safet a compensatoo casn ti f o - ey r dror pfo example, at reloading, negative reactivity will be inserted. Table 3 presents the calculated values of efficiency of various types burn-up compensator BN-35e th r sfo 0 reactor. Table 3 Efficiency of various types fuel compensators positioned in the centre of the BN-350 fast reactor. typd eRo Efficienc f fueyo l Efficiency of Efficiency of the rod sectio k/kA ) n(% absorbing section (%A k/k) (%A k/k) 1 0.137 0.094 0.231 2 0.137 0.145 0.282 3 0.237 0.145 0.382 4 0.181 0.145 0.326

The rod types presented in the table have the following compositions: fuee Typth l- sectioe1 n contains PuCh+UCh absorbine th d an , g one-UC>2 (depleted uranium); Typ PuO2+UC- e2 fuee th l n sectionhi depleted an , d metallic uraniumn i the absorbing one; Type 3 - the fuel section contains the low enrichment zone plutonium without uranium-238, wit n inera h t e absorbindiluentth d an , g section is made of depleted metallic uranium; e fueTypth - l e4 section contains U-238-free uranium-235 with inert diluentabsorbine th d an , g sectio similas ni . 3 typed o rt an s2 It can be seen from Table 3 that in a maximum case the efficiency of the improved compensator exceeds that of a conventional fuel compensator 1.65 times. However, this compariso t conclusiveno s ni , sinc consideratioe eth s nha been given to "fresh" compensators. It the values of economically justified burn-u f fueo p l compensator e takear s n into account e abovth , e gais i n decreased approximately to 1.34. This improved efficiency of fuel compensators appears in sufficient for the BN-800 reactor to provide the required period of reactor operation between reloadings. Therefore e developmenth , f fueo tl compensator e BN-80th r fo s0 design were terminated. However, the above developments can be of interest for perspective fast power reactor with high breeding. CONCLUSION Work performed during nearly thirty-years period has resulted in practically complete solutio problee controth e f nth o f mo l rods developmen r fasfo tt power reactors. presene Byth t time there have been develope experimentalld dan y testen di the BN-350 and BN-600 reactors: - control and safety system rods of all types (safety rods, compensating rods, control rods) with higher efficiency, high reliabilit extended yan d specified life eff0 designd . 50 ro daystim ~ e f eth o s ; have some margiincreasn a r nfo f eo their life time up to 650 eff. days; - "trap"-type safety rods allowing to exclude or substantially reduce the use of expensive high-enrichment boron-bearing materials. Work on partial experimental validation of increased-efficiency fuel compensators for their possible use in advanced fast reactors with high breeding rati bees oha n also carried out.

18 DEVELOPMENT AM) IMPROVEMENT OF CONTROL BN-35E BN-60D TH ROD 0R AN 0SFO REACTORS

A.I. EFREMOV, V.I. RYAKHOVSKIKH, V.B. PONOMARENKO, V.M. CHERNYSHOV Moscow Factory Polymetals, Moscow, Russian Federation

Abstract

This paper reports on the main direction of activities regarding improvement of technical and economic characteristics of control rods, results of the rod modernization, and perspective f furtheso r improvemen f controo t fase l th rodt r reactorsfo s BN-35d 0an BN-600.

At all time in the design and operation of Russian fast neutron reactors problems connected with the development, tests and fabrication of control protection system (CPS) rods were resolved. CPS rods are the most critical components of the nuclear reactor. CPS rods for fast reactors must be designed for operation in high neutron t operatina fluxe d an s g temperature claddinf so 600°Co t p gu . At this condition rodS sCP mus conformee tb foUowino dt g requirements: - to provide the free moving in directional tubes and the reliability of the insertion in core at an emergency; - to keep the mechanical strength; keeo t -requiremene pth t efficience leveth f o l y durin operatine gth g life. Using absorbing and constructional materials must: compatible b o t anothee - whole on th o et n e ri operating temperature region and emergency conditions; havo t minimae - eth l irradiation destructio swellind nan g under neutron fluxes. Effort developere th f o s s designing moder rodS nmaisCP o focutw nn so directions: improvemen- reliabilite th f o t operatind yan g lifrodsf o e ;

19 - development of the economical constructions. Stages of the development and improvement of rods is shown in schem. e1 1.Cassette-typ rodS esCP desig wite th hf n o AELs , containing different enriched B4 t developmenC.A rodS BN-35r sCP tfo BN-60d 0an 0 reactors were used sealed AELs with specia compensatos lga reducinr rfo g pressure

Increasin reliabilitf go y Development of economical and resource constructions

Development of optimal Development of "trap"-type construction of gas removing rod, increasin absorptiof go n and sodium filling AEL block effect due to local softening spectrum

Search for radiation-resistant construtional materials Developmen dual-purposf o t e CPS rods - producers radioactive isotopese Changing tube construction

Renouncement of using of activating absorbers in standard rods Development of repeated using enriched boron technology

Design of modern rods: resource - 550 days fluence - 3-1027n/m2 dose - 77 dpa

Scheme 1.

20 of helium, which release from boron carbide under irradiation (fig.l). e constructioTh s show i f seale no L fig.2n ne operatini dAE Th . g life of seale witL hdAE variously enriched boron carbide depend pellen o s t releaseswellins ga outles d Ga g.an t from boron carbide isn't more than 20 f helium%o , produce t s burnui da d an f boro po % n10 atomo t p u s compensated by creating of free volumes in AEL, but this reduce the part of the absorber in the rod and its efficiency. Swelling of the boron carbide takee ar n into accoun sufficieny b t t radia axiad an l l gapsL withiAE e nth e conditioth t a 0.5f no % geometry size changin boro% 1 r n gpe burnup .

Head

Joint

Lengthening tube

Operative section

Covering tube

Absorbing element

Entry nozzle

Fig Constructio. 1 . cassette-typf no e regulating rod.

21 Cladding

Absorber, europium oxide, boron carbide

Fig.2. Sealed absorbing element.

Sealed AEL with boron carbide don't provide enough long operating life, especiall t higya h intensity burnup. 2.Investigation e radiatioth f o s n behavio f theso r e AELs show, that sealed AEL with B±C can be recommended only at small boron burnup. t furtheA r improvemen rodS CP s f constructiono t BN-350r sfo , BN-600 reactors, to prolong their operative life, constructions with so-called slot filters were used for filling of the inner volume of AEL by sodium coolant

22 removind an products ga f go coolano st carriee ar t t througdou same hth e filter (fig.3). The strength factor for cladding of unsealed AEL is considerably more, than for sealed ones, because helium pressure is absent, but joints and carrying construction e samth botn e i ar sh cases. Increasin radiaf go l gaps between absorber pellet claddind an s made gar e without decreas- inpellet'f go s diamete changino t e du rcladding'e gth , mm s7 wal0. o t l because the cladding aren't affected internal pressure as compared with sealed ones, and the value of the radial gap allows to guarantee the pel-

Special unsealed "slot"-type block, providing sodium entry into the AEL anheliue dth m outlet

Absorber boron carbide

Cladding

Fig. 3. Unsealed absorbing element.

23 lets swellin 2.3o gt % geometry size changin burnu% 1 r boron-1f pgo pe 0 atoms at 6 month campaign or to 1.2% per 1% burnup of boron-10 atoms at 12 month campaign. e presencTh f sodiueo m layer betwee e claddinnth d pelletgan s provides the stable working temperature of the absorber, approximately correspondin minimuo gt m swellin f borogo n carbide. Usin unsealee gth d AELs wit boroe hth n carbid sodiud ean m laye providee ar r increaso dt e operatine th g lif 1.5...n ei 3 times operatine Th . g experience these AELs showed, that the main limit of their operating life is the interaction between boron carbid claddind ean g throug sodiue hth m layer, howevee rth interaction speed allows to provide the operating life of unsealed AEL up to 3...4 years. It was a new direction in the world practice for fast sodium reactors as foreig designd nro thesr sfo e reactors were directe develoo dt removs pga - ing systems without sodium getting into the AELs. The next operating experience of AEL showed that this design fully justified itself. At present unsealed AEL with slot filters, containing the enriched B4(7, are used in standard CPS rods in BN-350 (rods CR, SR, TC) and BN-600 (rods SR, SR-L, CR) reactors. This construction based on new constructional materials ChS-68 and EP-450 (ferrite-martensite type consideree ar ) s promisinda r futurgfo e rodS BN-35r CP sfo BN-60d 0an 0 reactors. Absorber in CPS rods for BN-350, BN-600 reactors must be possess rather high physical efficienc keept d giveya an t durinnsi d sizro f l eo g al operating t leslifno es than allowable level. Absorbing material e srod th isin sr musgfo compatible b t e wite hth constructional materials in whole operating temperature range and erner-

24 gency conditions and also to possess the high resistance to corrosion in sodium of the first contour to have minimal radiation destruction under high neutron flux. In the sealed and unsealed CPS rods for the BN-350, BN-600 reactors boron carbide enriched with boron-10 isotope up to 80- 85% pellets were use absorbins da g material . sPellet preparee sar y db t powdeho r pressin e protectiv th fren gi f o e e atmospher vacuumn i r eo . Boron carbide powder were prepare elemene th y db t synthesis. technologAe t pressinsth ho e th improve s gf i y o boroe dth n contenn i t pellets of synthesized boron carbide was increased from 68% mas. to 76%, physice th d san densit decreases ywa d from 2.41 g/cm 2.1o t 3 9 g/cm3. In this case minimal density become g/cm8 s1. 3, maximu g/cm6 2. m - 3. 3. At the development of regulating units it should be considered the following base conceptio fase th t r powenfo r reactor regulation system. In this case two main groups are formed: - regulating units operating outsid coref eo ; - regulating units locating in the core during reactor operation. This classification allows to select the most favorable absorbing materials and constructions for the different regulating units. Scram rods (SR temperaturd )an e compensation (SR-TC) assige th no t first group. For SR, SR-TC rods it is advisable to use the materials and constructions providing the maximum efficiency of the assembly desing wit enrichee hth d (80-90%) boron carbid trap-typd ean e constructions. trap-typf o L AE en I rod combinatione sth s Eu^O^ B±Cr o with effective

moderating materials base hydriden do s (zirconium, titanium, hafnium, europium) are used. In this case the local softening of neutron spectrum in the rod volume is the result of the neutron moderation on hydrogen

25 nuclei thae containear t metan di l e probabilityhydridth d e an th e f o neutron absorption is increased. The construction of such type rod allows significantly to increase the rod efficiency, to reduce the expenses of high effective costly absorbing materials e metae usinth Th f .g o l hydride particulan si r zirconiu- mhy drid modaratos ea europiud an r m oxide, boron carbide (enriched- ab s a ) sorber was found to be the most proficient. Experimental investigations of rods showed, tha trap-typr fo t , SP-CeSR BN-350r Tfo , BN-600 reactors the efficiency is higher by 10-15% than for cassette-type rods with enriched boron carbide (80%) and absorbing cladding. In this case the trap-rod demand enrichee sth d boron lestime9 «n s1. i s tha identicar nfo l size cassette-type rods. In such a manner for the trap-type rod the efficiency of enriched boron carbide using more than twic larges ea tha s cassetee-typr a r fo t e rodst I . shoul notee db increasedo t thae du t d specific efficienc e absorbth f o y - materialg in trap-typn i s e rod e burnuth s p rate f nucleo s i absorbes i r increased obtainee Th . d results show: -efficienc trap-typf yo witd ero h europium oxid zirconiud ean m hydride diameten i m m 0.8-0.83 s ri 7 efficience th f 5o cassette-typf yo e rods with 80% enriched boron carbide; -trap-type rods based on enriched boron carbide allow to have the efficiency, impracticable for cassette-type construction even with maximum permissible enrichment by boron-10; provido -t prolongee eth relative dth operatint - a ede d ro e gth liff o e creasin absorbef go r swellin keepind gan hige gth h efficienc meay b y f no generatio optimae th f no l neutron volumed spectruro e th . mn i

26 e constructioTh trap-typf no e rods, containing unsealed absorbin- gel ements with boron carbide enriched with isotope B-180-90o t p 0d u %an zirconium hydride with special hydrogen confined coating usin moda s ga - erator successfulls wa , y approbate standarn di d, CR-T cellSR f so C rods for BN-350, BN-600 reactors (fig.4).

Absorbing elements

Zirconium hydride

Fig.4. Cross sectio"trap"-type th f no e rod.

The operating experience on the study of neutron-physics characters of trap-type rods led to necessity of further development its construction for decreasing of the cost SR rods and increasing its working capacity. The above discussion, the relatively short period of the work of SR trap- type units is mainly connected with boron-10 burnup. The most part of boron-10 burn t undesou r influenc thermaf eo l neutrons indifferens i t I . t efficiencd ro e th yo t where thermal neutron absorbede sar . Thereforn ei the rod construction the second absorber is used to screen off boron-10 against thermal neutrons of the moderator zone. The such absorber must possess rather high cross-sectio thermaf no l neutron absorptio absoro nt b

27 them in the layer of some millimeters thickness and good constructional properties (fig5") heae Th .t resistance (hafnium, rhenium, tantalumd an ) rareearths oxides (Eu^Oz ) can be used as such materials. Hafnium is the best suited as the second absorber for the trap-type nuclear-physics it o t e SRdu s properties: melting temperatur e220- 0 (7°, densit y13.- 1 g /cm quantit, 3 atomf yo s0.04- cross-sectio, ? 4 10cm 2n 4i n

Zirconium hydride Absorbing elements

Titanium screen

Fig.5. Cross section of the "trap"-type rod with the titanium screen.

of thermal neutron absorption - 93 barn. Simple estimates show, that the optimum layer thickness of metallic hafnium, enough for absorbing of 90% thermal neutron, is equal to 4 mm. t shoulI notede db , tha thin ti s cas totae eth l quantit absorptiof yo e th n i boron absorbing elemen decreaseds i t . This valu abous ei t 30-50%e Th . main advantag developef eo d trap-typ significantls i d ero y more operating life, abou foln te td decrease nonuniformite th f do neutrof yo n absorption boron i n elements, significant decreasin boron-1f go 0 burnu alsd e poan th

28 rod efficiency is as good as unsealed standard PS rod with enriched boron for BN-600 reactor. e developeTh d trap-type contro consistd ro l s fro e partsmth , which contain neutron moderato - zirconiur m neutroo hydridtw - d ab nan e sorbers in the center. AEL with enriched boron carbide are mounted around the periphery, and the cylindrical graphite screen is placed between the moderator and AEL. The compensating rods (CR) and automatic regulating rods (AR) be- secone lonth o gt d group) absorbin7 , (n , g materials base europiun do m oxides have packe beeR A nd tchose materiaan type th R es C n a rodr fo l s with sealed AEL. makThiL seAE possibl obtaio et maximue nth m permissible resourc rodsf eo , whic limiteds hi , mainl workiny yb g capacity of constructional material insignificano t e du s t radiative destructiof no (71,7) absorbers. The investigations of radiative behavior of such types AEL with the absorber pellet alsd san o pulled-type AEL, obtaine draggine th y db f go e claddinth g together with powder filler, showe e higdth h reliabilitf yo these AELs and confirmed the rightful and the promising of such industrial goods. Sealed AELs based on europium oxide pellets are successfully used in standard rods (AR,CR-TC) for BN-600 reactor (fig.2). Sealed pulled type AEL standar e usee th sar n di d rods (AR BN-60f )o 0 reactor (fig.6). At present promising design, CR-TAR f o sC rod rodse ar s , fulfilling the dual function - regulating of reactor power and production of the radioactive isotope Cobalt-60 (fig.7) Developed rods consist of the external absorbing coating and the inner zone, containing an effective neutron moderator (zirconium hydride) and

29 Cladding

Absorbing material (europium oxide + molybdenum r tungsteno )

Nickel

Fig6." Pulled"-type unsealed absorbing element.

Absorbing neutron material Eu^Oo s+C

Moderator zirconium hydride

Ampoules with Cobalt-59 Fig.7.Cross section of dual-purpose absorbing rod - regulating of the reactor power and the production of radioactive isotopes.

30 an ampoule wit e starhth t Cobalt-59 isotope e ampouleTh . s wite hth start Cobalt-59 by their geometric dimensions correspond the standard gamm asource- s with Cobalt-60. sucn I h rods neutron absorbing materials base europiun do m oxide pel- usee letcobalb d productior dy sfo an ma t coref no radionuclidr sfo e hard damma-irradiation sources of high activity. This material allows to use it in the nuclear reactor regulating units, providing the production of hard gamma radiation source simultaneousld san y decreasin productioe gth n expenses. Du o optimizatiot e e chemicath f o n l composition ther s providei e d « 1.5 times more specific activity of material, comparing with that of pure cobal «1.d an t 2 times more comparing with europium oxide while havin equae gth l irradiation time. Supplies manufacture thif do s material hav higea h mechanical strength e eas , CR-Tmachinr ar yfo AR d e an Cth e rod f processingo se us e Th . provides: -high efficienc ratio t e o ydu betwee n absorbing materia effece th d - an l tive moderator in the internal rod's volume; -simultaneous production of Cobalt-60 with the activity not less than 100 Cu/g; -high speed of Cobalt-60 production due to variation of correlation between moderator's and ampoule's with cobalt volumes, and achieving the optimum neutron spectrum in internal rod's volume. At present two such rods for BN-600 reactor were designed and produced.

31 TABLE E MAITH .N CHARACTERISTIC F DIFFERENSO T MODIFICATIONF O S REGULATING ELEMENT BN-60R SFO 0

CPS rods modifications Characteristics

SR AR CR-TC

AEL-type Sealed Unsealed Trap- Pellet-type Pulled- Pellet- Pulled- Dual- type type type type purpose

Eu O + Eu O + B4C B4C B4C Eu2O3+ Eu2O3+ 2 3 2 3 Eu2O3+ Absorbing (80% (80% Mo Mo Mo Mo Mo material (80% B-10) B-10) B-10) (7,4 (6,5 (7,5 (6,5 (8,2 g/cm3) g/cm3) g/cm3) g/cm3) g/cm3)+ ampoule Co-59

Moderator - - ZrHM - - - - ZrH1;8

Setting operating time 180 360 330 360 360 360 360 360 (eff. days)

Radiactive Not less isotope 100Cu/g production

Maximum factual 130 360 330 360 360 406 406 - operating time (eff. days) lowee th Noten assemblplacee re i b parth t usinR n f A S :o t dca s ga t ampouley i r sfo irradiation of constructional materials. (Technical documentation was performed).

32 EXPERIENCE OF THF, BN-600 REACTOR CONTROL RODS DEVELOPMENT

Yu.K. ALEXANDROV, B.A. VASILYEV, S.A. ISKHAKOV, O.B. MISHIN, V.A. ROGOV, A.S. SHABALIN OKBM, Nizhny Novgorod, Russian Federation

Abstract

Durin BN-60e gth 0 reactor operation (1980-1994) controe th , desigd s ro l nwa modified with a view to improve reliability and lifetime. Control rod updating was carried out in two stages. The design of control rods modified during the first stage (1981-1985), reduco t m witai e n deformatioha maie th nf no unit excludd san e their jammin guidn gi e sleevesmodifiee th d an , d desigcontroe th f no l sleeves, allowe dcontaca t between control rods and sleeves in the core area to be excluded. 350 eff. lifetime days were reached as a result of the first updating of control assemblies. At the second stage (1986-1994), the structural material f absorbeso claddingsn rpi wrapperd ro , sleevesd san e wels th a , s a l applicatio f europiuno m oxide instea naturaf do l boro absorbes na compensatinn ri d gan control rods were optimized secone resula th s f A .o td updating, control assemblies lifetime was extended to approx, 500 eff. days.

1. INTRODUCTION

BN-600 reactor control assemblies (CPS rods and sleeves) were designed in accordance with their functional destination. Compositio d maian n n operating performances of these assemblies are presented in Table 1. When developing of control assemblies experienc f similaeo r products operatio BOR-6n no BN-35d 0an 0 reactors was taken into account. Russian specialists presented results about their operation no IAEA IWGFR meeting (1,2). Special repor n resulto t f BN-60o s 0 reactor control assemblied s an pilo t se t description of disadvantages removal was done on the seminar (3). During BN-600 reactor operation (1980-1994 years enhanco t ) e reliabilitd yan lifetime of control assemblies changes are made in their design.

2. DESIG PILOF NO T TSE

Principal design of pilot set (1980-1982 years) of BN-600 reactor control assemblie s showi s n Fig-i n s 1-4. Their main geometric size e givear s n Tablei n s 2-5 and operation conditions and results of post-reactor examinations - in Table 6. Measuremen f speno t t assembly geometr d latean y r investigations showed that their lifetim s limitei e y deformatiodb f hingo n e join S rodtguidd partCP an s f eo s sleeve wrapper placed in core during power operation. The main reason of intensive deformation of assemblies is high temperature level (500-550'C massivn i ) rodS e partCP s f hingo s e joint (-2 correspondin) 4mm o gt the maximum of temperature dependence of irradiation swelling of X18H10-type steel.

33 TABLE 1. COMPOSITION AND MAIN OPERATING PERFORMANCES OF BN-600 REACTOR CONTROL ASSEMBLIES

Name Value

Shutdown Shutdown Control Compensa- rod rod - loop rod tind ro g

1. NumbeS rodCP s f o r (sleeves) involved in BN-600 reactors CPS 5 1 2 19

2. Place of rod absorber Absorber is withdrawn Absorber Absorber arrangement relative to from core is in core moves up core at reactor power center from operation region bottom operating position in uppee on r

3. Maximum velocity of d movementro c m/ , 1 7-10-2 5-10-3

4. Feature of CPS channel (sleeve and S guidCP e tube) Complete Intermittent

3. FIRST STAGE OF MODERNIZATION

Basing on results of pilot set operation of BN-600 reactor control assemblies two stages of them modernization were carried out. Table contai5 2- s n main geometric size f controo s l rods after first modernization (1981-1985 years), and Table 6 contains operation conditions and results of post- reactor examinations. Thicknes f massivo s S rode partCP s f hingo s e joint s reducewa s d dowo t n 10 mm, conditions of their cooling were improved, height of pads contacting with guide sleeve was decreased. Wrapper diamete f bottoo r m joined lindiameted kan f entranco r e nozzle spacing werd pa e reduced additionall contron yi l rod shud san t down rods. Remova diffusiof o l n chromium-nitride coating from hing guidS rod eS frod CP part eCP f san m o sleevef so s wrappers increased plasticit f structurao y l materia d stabilitan l f theio y r deformation.

34 , -— I •' x • iE i '; HanpaB7L=Dmaa Tpyö CRDM H a M guide tubi eguide tube ! r Äi — ; < i V ffl -fuy ^ii OJ rzJILSa Guide sleeveve , /= i \ \ i ' J ' 078 ^ -* i 1 assemblie. t s (^ T 1) Ot> i •J 10

il ii

o tr— | » 1 1 C\}

B Ei_ _L

073x1

! = ^

&5 S*5

\ / ^ •—— «*^ j__— * Modernized variant MogepHgaHpOBaflHHE s* & BapnaHT^ _jt

Fig.1. Shut down rod and guide sleeve

35 CRDM guide tube TpyöM H a

Surrounded fuel assemblies TBC

Fig.2. Shut - loodowd guid d an pro ne sleeve

36 CRDM guide tube \ \ t • i2\ e \ : rmr&sa Guide sleeve '- \ —ri ^-

d ro S CP i 1 \ ! , V ssemblies ' t i A 1 ., 078

A - A /

01 L IN=c

0! O 1— 1 09t5xOT5 CVJ

>—— " •

L L M E - E

^23x0.7 073x1 •

—'-LJ-JI- 1

,> Modernized variant MojepHireiTpOBaHHffli ^-

Fig.3 . guidd Controan ed sleevro l e

37 \ . (' . I \ • in t- i4 - i/

1— ^ Ar- A y-f 089x1.5 hA

Tj«

085x1

: t _ « * 023x0.7 - i 082 080

c Modernised variant ftJOjepHgSHpOBaHHHA s. 9 1 VVv- /T

Fig.4. Compensating rod and guide sleeve

38 TABLE 2. MAIN GEOMETRIC SIZES OF SHUT DOWN ASSEMBLY

Name Piioî First Second set modern!- modernization zation

I. d wrappero Siz f o e r tubem m , 0 73x1

2. Outen ri diamete d ro f o r 0 74 the hinge region

3. Sleeve configuration Round Hexahedral

4. Inner diameter of 76 78 (outside the core sleeve, mm boundary)

5. Inner size of hexahedral wrapper within core, mm - 92

6. Rod length, mm 2100

7. Number of absober pins in rod, pieces

8. Size of absober pin cladding, mm 0 23x0.7

9. Pellet diameter, mm 20.4 19.6

10. Absorber B4C-80 r B-1%fo 0 B4C-92% for B-10

11. Leaktightness of absober pin Non-leaktight

12. Structural material of rod XI8H10-type IXI3M2BOP-type austenitic steel ferritic steel

13. Structural material X16H15-type X16H15+TÎ- of absober pins austenitic steel type austenitic steel in cold deformed state

14. Structural materia f sleevo l e X16H36-type X16Hll-type IXI3M2BOP-type austenitic austenitic ferritic dispersion- steel in cold steel hardening steel deformed state

15. Coating of head, hinges Diffusion- of rods and sleeve tubes chromium- nitride

39 TABL . MAIE3 N GEOMETRIC SIZE SHUF SO T DOWN ASSEMBLY LOO- P

Name Pilot First Second set moderni- moderni- zation zation

1. Size of rod wrapper tube, mm Of 73x1

. 2 Outer diameted ro f o r 0 74 e hingith n e region

. 3 Sleeve configuration Round Hexahedral

4. Inner diameter of sleeve, mm 76 78 (outside the core boundary) 5. Inner size of hexahedral wrapper within core, mm 92

6. Rod length, mm 2100

7. Numbe f absobeo r r pins in rod, pieces

8. Siz f absobeo e n pi r eladding, mm 0 23x0.7

. 9 Pellet diameterm m , 20.4 19.6

10. Absorber B4C-19.8 r B-1%fo 0 11. Leaktightness of absoben pi r Non-leaktight

12. Structural material of rod XI8H10-type IXI3M2B . 13 Structural material XI6H15-type X16H15+Ti-type of absober pins austenitic steel austenitic steel in cold deformed state 14. Structural material X16H36-type Xl6HH-type IXI3M2BOP-type of sleeve austenitic austenitic ferritic steel dispersion- stee n coli l d hardening steel deformed state

. 15 Coatin f heado g , hinges Diffusion- of rods and sleeve tubes chromium- nitride

40 TABLE 4. MAIN GEOMETRIC SIZES OF CONTROL ROD ASSEMBLY

Name Pilot First Second set moderni- moderni- zation zation

1. d wrappero Siz f o e r tubem m , 0 73x1

2. Outen ri diamete d ro f o r 0 74 e hingth e region

3. Sleeve configuration Round Hexahedral

4. Inner diametef o r 76 78 (outsid e corth e e sleeve, mm boundary)

5. Inner size of hexahedral wrapper within core, mm - 92

6. d lengthRo m m , 2100

7. Number of absober pins in rod, pieces 31 4

8. Size of absober pin cladding, mm £5 9.5x0.5 0 23x0.7

9. Pellet diameterm m , 8.2 19.6

10. Absorber Eu203 B4C-19.8% for B-10

11. Leaktightness of absober pin Leaktight Non-leaktight

12. Structural material of rod XI8H10-type IXI3M25P-type austenitic steel ferritic steel

13. Structural material X16H15-type X16H15+TÎ- of absober pins austenitic steel type austenitic stee n coli l d deformed state

14. Structural materia f sleevo l e X16H36-type XI6Hll-type IXI3M2BOP-type austenitic austenitic ferritic dispersion- steel in cold steel hardening steel deformed state

15. Coatin f headgo , hinges Diffusion- of rods and sleeve tubes chromium- nitride

41 TABL . MAIE5 N GEOMETRIC SIZE COMPENSATINF SO ASSEMBLD GRO Y

Name Pilot First Second set moderni- moderni- zation zation

1. Size of rod wrapper tube, mm 0 89x1.5 0 85x1

2. Outer diameter of rod in 90 e hingth e region

3. Sleeve configuration Round Hexabedral

4. Inner diametef o r 92 82 (outside the core sleeve, mm boundary)

5. Inner siz f hexahedrao e l wrapper within core, mm - 92

6. d lengthRo m m , 2874

7. Number of absober pins in rod, pieces 48 8

8. Size of absober pin cladding, mm 0 9.5x0.5 23x0.0 7

9. Pellet diameter, mm 8.2 20.4 19.6

10. Absorber Eu2O3 B4C-19.8% for B-10 11. Leaktightnes f absobeo s n pi r Leaktight Non-leaktight

12. Structural materiad ro f o l XI8H10-type IXI3M2B

13. Structural material X16HI5-type X16H15+TÎ- of absober pins austenitic steel type austenitic steel in cold deformed state

14. Structural materia f sleevo l e X16H36-type XI 6H 11 -type IXI3M2BOP-type austenitic austenitic ferritic dispersion- steel in cold steel hardening steel deformed state

15. Coating of head, hinges Diffusion- of rods and sleeve tubes chromium- nitride

42 TABLE 6. IRRADIATION CONDITIONS OF ASSEMBLY IMPORTANCE UNITS

CPS rods Guide sleeves Assembly type Set name lifetime, dpa T,'C d, mm lifetime, dpa T,'C d, mm eff.days TRN eff.days TRN

Pilot 200 37 400-530 0.2-2 300 60 510-550 0.8-4

Shut dowd ro n 1-st stag f modernizatioo e n 330 49 380-400 <0.8 330 66 430 <3

(Shut - loopdow d )ro n 2-nd stag f modernizatioo e n 500 67 380-400 <0.2 500 80 430 <0.3

Pilot 200 37 400-530 0.2-1.5 300 60 510-550 0.8-4

Controd ro l 1-st stage of modernization 330 49 380-400 <0.5 330 66 <430 <3

2-nd stag f modernizatioo e n 500 70 380-400 <0.2 500 80 <430 <0.3

Pilot 200 37 500-600 0.5-2 300 60 550-650 0.8-6

Compensating rod 1-st stage of modernization 330 49 400-420 <0.8 330 66 <430 <3

2-nd stage of modernization 500 65 400-420 <0.2 500 80 <430 <0.3 >

Importance units of assemblies: Shut - f spacinentrancdowo d d ro n pa g e nozzle Contro d compensatinan l g rod - spart f bottoo s m hinge joint Guide sleeve - cor e central plane S guidDesigCP ef no sleeve s simplicifiewa s makiny b d g them using hexahedral wrapper tube and cylindrical bushing placed above core in sleeves of shut down rods d controan l rod r undeo s r cor sleeven ei f compensatino s g sleevrodsw Ne .e design enabled to increase minimum gap between rod and sleeve, to decrease the gap height, having provided additiona p betwee ga lS rod d theiCP an sn r sleeves within core. Therefore flowrate value of coolant, cooling absober pins bundle reduced, and temperature value of absober pins cladding increased correspondingly by approx. 30°C. Change of compensating rod absorber was done from absorber on the base of euoropium oxide, having high decay heat releases and induced radioactivity to natural boron carbide. a firs s tA modernization result design lifetime (350 eff. days r controfo ) l assemblie s achievewa s d (real lifetime amount0 eff.days)33 o t s .

4. SECOND STAGE OF MODERNIZATION

Principal design of BN-600 reactor control assemblies on second stage of modernization (1986-1994 years) is showed in Fig-s 1-4. Table contai5 s2- n their main geometric size Tabld s- an operatio e6 n conditions d resultan f post-reactoo s r examinations. Radiation-resistant IXI3M2BOP-rype ferrite steel was used'for manufacturing control assembly wrapper tubes, X16H15+TÏ c.d. austenitic r absobe steefo - n l pi r claddings. Absorber on the base of euoropium oxide was replaced by natural boron carbide in contro e siz f absobelth o e rodd n claddinan spi r s unified 0 wa 23x0.( g ) . 7mm Hinge joints along d absorbelengthro f o r section were ecxlude n shudi t down rods. Design of second modernization sleeves is similar to that of first modernization sleeves (besides structural materials). Change of core to new modernized loading (core height was changed from s carrieo t 100 wa t simultaneousl) m ou dm 0 mm 0 75 y with modernizinf o g BN-600 reactor control assemblies. Neutron fluenc d fuean el element thermal load decreased in modernized core. It resulted in temperature decrease of rod importance unit down to 400°C and neutron fluence by -20%. structuraw ne f o le materialUs d introductioan s f abovo n e listed changef o s assembly design enable o increast d e control assembl 0 effy 50 .lifetim o dayst p u e.

CONCLUSION

Main results of BN-600 reactor control assembly operation (1980-1994 years) consideree ar maid dan n stage theif so r modernization directe increaso dt f realibiliteo y and lifetime of control members are given. During first stag f modernizatioeo n (1981-1985 years) changeS made CP sar n ei d desigro n that enable decreaso dt e deformatio f maio n n units d improvean , d design sleeveS oCP f s provide increasp dga e betweesleevd an d e nro withi n core s resulA . t of first modernizatio lifetime nth controf o e l assemblie -35s swa 0 eff. days (-66 dpa). Chang f structurao e l materials control assembly wrapper tube - sIX13M25OP - type steel and of absober pin cladding - X16H15+Ti-type steel in cold-deformed state

44 was done for second modernization (1986-1994 years). As result of this modernization e lifetimth f controo e l assemblie s —50wa s 0 eff. days (~80 dpa).

REFERENCES

1. MarepnajiH coBemaHHs cneima^HCTOB MPFBP MATAT3 "HoiviomaiomHe MaxepHa/i crepacHH u pery;rapoBaHHs obicrpHX , CCCP, 1973 r.

2. MaxepHajibi coBemaHHa cneimajiHCTOB MPFBP MAFAT3 " H crepxHH pery^uipOBaHHa obicrpHx peaKTOpos" OÖHHHCK, CCCP, 1983 r.

. 3 MaTepHajiH optrraHo-coBeTCKoro ceMHHapa KOHCTpyKIIHOHHbîX MaTepH&TlO 3^6MeHTOH B B aKTKBHH H peaKTOpO30 X H B , 1990 r.

Next page(s) left blank 45 CONTROL ASSEMBLY TO BE USED IN CEFR

GUANGSHANE XI , ZHANG RUXIAN China Institut f Atomieo c Energy

WANG YONGLAN Xian Jiaotong University

I SHIKUL N China Institut f Atomieo c Energy

China

Presentet VoznesenskiR. by

Abstract

This paper describes the structure and design data selected for control assembl n CEFi y n detailRi e controTh . l assembl f CEFo y R mainly consista f o s guide tube and seven absorber pins.Each pin includes a section absorber material

B4C of 510mm long and a stainless steel cladding , and sodium medium is filled in

the gap between B4C pellet and cladding. Then the characteristic properties of

the prepared BC pellet to be used in CEFR are presented, and further researches 4 pelleC 4 e alsB e don ar b r to e paperfo o egivet th n i n.

. 1 INTRODUCTION

China Experimental Fast Reactor (CEFR) contains eight control assemblien i s which, based on the conceptual design, two control assemblies serve as safety rod to perform rapid shutdown during off-normal conditions, four as compensation rod for temperature effect and fuel burn-up and two as regulator rod to perform neutronic start-up and shutdown capability and control power level for normal operations. Although these control assemblies will perform different functions, their structur s quiti e e samtheid an e r position e core onlth ar e yn i s different. These control assemblies all are independent parts, and each control assembly occupie a positios n correspondin a fue o t gl assembl e coreth n .i y

Boron carbide B4C is selected as absorber material. The B 4 C pellet has preliminary been prepare s maiit d n an dpropertie s have been measure y Xiab d n Jiaotong university. Thermal conductivit e B^th C f o ypelle s testini t n i gChin a Institute of Atomic Energy at present.

. 2 STRUCTURE

The control assembly consists of a fixed guide part and a movable part containing absorber pins, as shown in Fig 1 . e guidTh e part include a guids e tube with insid f circulao e d outsidan r f o e hexagon, a operating handle and a positioning leg. The guide tube is made of Austenitic 316(Ti) stainless steel, i.e, minor titanium is added in standard AISI316 stainles t sflat-to-fla steelou s a distancIt . s ha tf 59.0mo e d inneman r circula a diamete s ha r f 56.0mmo r .

47 «j_j— | ivsB^i. M, : ' v*Y ^ " r—- -T r -t --r -rr-t— i— r -r-t— r-f — t---. ^g— Jj^fcU^ o 510

881

(a) Absorber pin

22E1

V /". ^fm-^t"^"^^T- - uurociEEp"

(b) Control assembly

i Hold-down spring 2 Cladding peilet Operating head

5 Operating handle 6 Grid 7 Absorber pin 8 Duct

9 Guide tube Positioning leg

Fig 1 . Desig f CEFo n R control assembl d absorbean y n pi r

The movable absorber part consists of a circular duct, seven absorber pins, positioning a handlingrid d an s g head. Ther e orificar e e surface th set f n o o se the upper and lower positioning grids, to make most coolant pass through the absorbe n bundlpi r d providan e e best cooling. One control assembly contains seven pins whic e arrange ar ha trigona n i d l pitch-16.0mm e wirTh e. wrape radia e th use ar s a dl separato e absorbeth r fo r r pins within an assembly, to keep up the coolant channel. These absorber pins are fixed on the upper and lower grids, then lower grid is fixed on the leg of this assembly. So the axial position of these absorber pins is determined in the assembly. The absorber pin is designed as sealed structure".It includes a cladding made of 316(Ti)S.S, B^C pellet f 510mo s m long, hold-down spring a heliu, s plenumga m and upper and lower end-plug (see Fig 1). Table 1 gives main design data of the control assembl r CEFRfo y . A wider gap (mornal diametric gap of 0.8mm) is seleted between the cladding

and B4C pellet and sodium medium is filled in this gap. The gap accommodates the BC matrix swellin irradiationy b g t bese ge sodiuth t n , heaca m t transfer

behaviou4 r in the gap, so the B^C pellet temperature decreases.

* A vented absorber pin is also selected as a candidate for CEFR.

48 Tabl 1 e main desig ne contro th dat f o a l assembly

parameter data

guide tube material 316(Ti)S.S form inner circular, outer hexagonal outer flat-to-flat distance 59.0mm inner circular dia. 56.0mm

duct of absorber pin bundle material 316(Ti)S.S outer dia. 52.0mm inner dia. 50.0mm

absorber pin pin number / assembly arrangement trigonal pitch 16.0mm pin dia. 15.0mm inner dia f .claddino g 13.0mm cladding-thickness 1.0mm gas plenum height 200mm cladding material 316(Ti)S.S

absorber pellet

absorber material B4C dia. 12.2mm height 20.0mm stock height 510mm density 92% T.D 10 B enrichment in B4C 91%

The absorption of neutrons by B4C leads to the (n, a ) reaction which yields both lithiu d heliuman m atoms larger tha e originath n l boron atom, though somf o e e heliuth s mreleasedi l lithiual , m producte parf th o heliu d t an s m retained

cause BC matrix swelling. To accommodate this swelling, besides lower BC pellet 4 densit s selectedi y 4 s designei e wide th p , ga r d betwee e claddinth n d pelletan g . Assuming that CEFR control assembly operates for 402 full power days in this 22 3 reactor, maximum capture leve n i Bl o l.Os 4t i higheCx p 10u r / cm(n) ,a , corresponding diametric swelling of 5% (AD/D). The number of the helium atoms yielded by B * C absorption neutrons is approximately the same as the number of neutrons absorbed, and helium generation correlates directly with boron burnup. Helium gas release from the B„C matrix is temperature dependent. For the operating temperature of CEFR control pin, the percentage released is assumed to be 30% (average), when the plenum height is 200mm, heliu s pressurmga e build p 8.8MPau s . When the surface temperature of the BC pellet is more than 700'C, FeB or

FeB is formed on the inner surface of cladding4 , as a result of cladding

: a

material reaction with B„C pellet, with aggravatinc g this reactio n i sodiun m

49 Table 2 comparison of examination and design values

for chemical compositio pelleC 4 B f to n

examination values constituent unit design values 4.0 3.8 4.2

total boro tota+ n l carbon wt% 99.04 98.92 99.13 >99 total boron wt% 78.23 77.40 79.14 >77.92 Nitric acid-soluble boron wt% 0.12 0.11 0.13 <0.5 water-soluble boron wt% 0.07 0.05 0.09 <0.2

B203 ppm 492 475 503 <1500 chloride ppm 50 46 53 <75 fluoride ppm 23 24 24 <25 Ca Ppm 130 134 105 <1000 Si ppm 1500 1435 1483 <2000 Al ppm 118 135 108 <3000 Mg ppm 105 109 112 <1000 F ppm 159 143 162 <3000 freC e wt% 0.16 0.19 0.00 <0.5 moisture (water) ppm 135 124 105 <250

environment. When betweep sodiue claddingga th pelleC e 4 s d mfilleth i B n an tn i d,

e temperaturth e B4th C f o epelle s substantialli t y decreased e maximuTh . m surface temperature of B^C pellet for CEFR is not expected more than 550 °C , so this reaction products (FeB or FeB) can be a small quantity and not obviously effect

e effectivoth n e thicknes f claddingo s . 2

. 3 PROPERTIE PELLEC 4 B F O ST

The BC powder* has been prepared using higher pure boron and carbon powder

by directly4 synthetic technique,then by hot pressing and sintering, finally the

pelleC 4 B s finishei t d inte geomatrioth c surfacsizd an e e specification required.

1 Chemica3. l compositio f finisheo n d B4C pellet

Chemical composition prepared B4C pelle d beeha tn analyzed e examinatioth , n result corresponds with the standard GB5152-85 of the People's Repubic of China d ASTMC791-83an e comparisoTh . n between examinatio d desigan n n valuess i C2] listed in table 2.

Although from table 2, we can see that three of stoichiometry for B4C (i.e, with thre C ratios—3.8B/ e d 4.2an ) 0 4. hav, e bee e npropert th usen i d y

investigatio pelletC 4 B r a stoichiometr, fo n s onli 0 y 4. f selecteo y n i thid s design for CEFR.

3.2 Properties of B4C pellet The grain size of B^C pellet samples is observed by metallographical n figuri examination ) (b show2 e d an n ) respectivel(a , y microscopic examinationes y sinterinb m o grain fotw 5 gr1 n ~ witsize10 f h5~10o d s an differen nm t ho t

* Natural boron powder is used in the investigation for the preparation technique e sampleth d f non-irradiatioan o s n test.

50 ! -' v«

(a) 5-10um (2100*C, 15min) (b) 10-15um (2100'C, 60min)

2 Metallograp g Fi samplC B« f eo h

-pressing time. It has been seen from this figure that their grain sizes are

basically homogeneous e impuritie th t ,existenc no e ar s eC pelle4 B almos tn i t matrix and the form of pore appear as equiaxed.the porosity corresponds to the density measurement. The average value over the pore size is around 0.2 u m, maximum size among this pores is 3 n m. The mechanical properties and thermal performance had been examined for the grain pelleC siz B f 5~10 o er tfo samplenm . This characteristic data include

meltin4 g point, linear thermal expansion coefficient (room temperature to 1000 *C), bending strength, compression strength, young's modulus, poisson ratio, rupture toughnes d thermaan s l fatigue. These test results were liste n tabli d . 3 e

. 4 FUTURE TEST

The characteristic data of B^C pellet is based on the prepared sample during preliminary design phas r CEFfo e R control assembly. Beside e optimuth s m process technology will continue investigating, main two experimental works will be done as follows,

• Compatibility of B4C with stainless steel cladding material. A compatibility tesr B4fo tC with stainless steel cladding material wile b l investigated under sodium medium environmen d simulatean t d operation conditiof o n CEFR control pine producTh . s thicknes it e t inneth d FeB an n ro s surfacf o e

cladding wil e examinedb l . 2

• Irradiation test of BC in reactor. The main purposes of this test are to 4

obtai e resistancth n o swellint e g behavou releasd an r e rat f heliuo e mC 4 fro B m matrix o evaluatt , e operatioth e n behaviou e absorbeth f o r r pin.

5. CONCLUSION

This structure desig s undei n e conceptuath r l phasee , b althougn ca t i h changed in the detial design for the future, this structure form and main structure parameters will basically satisf r CEFR'fo y s requirements e froth m behaviour analysi e controth f o s l assembly.

51 Tabl 3 e main properties dat r Bfo 4aC pellet

density of test examination value item unit sample temp. average (%T.D) CO 123456

melting point •c 92 2445 2445

room temp. — 423.0 — 356.0 380.— 0 383.0

200 389.— 8 363.— 3 375.— 1 376.1 bending strength MPa 92 400 304.3 289.0 338.0 — — — 310.4

600 295.5 301.8 325.8 — — — 307.7

88 — — 944.— 0 752.— 0 847.0 room compression strength MPa 92 1273 1163 2143 1890 — — 1617 temp. 98 3511 — — — — — >35U

young's modula4 s37 4 GP37 a 4 37 4 37 4 37 4 37 374

poisson ratio / 0.22 0.21 0.19 0.19 0.18 0.17 0.193 room rupture toughness (MPa)^ 92 7.73 8.09 7.71 — — — 7.84 temp. linear thermal expansion coeficient 1

thermal fatigue(R) •c 140.4 _____ 140.4

The mechanical and thermal properties examinations of the trial-producted B4C pellet show that this process technology wil adoptable e b ltria th 4B C ld an e pellet wil e promisinb l r CEFRfo g t furthebu , r some researcheC pelle 4 B t r fo s will be done in the future.

REFERENCES

1. Status of Liquid Metal Cooled Fast Breeder Reactors, STI/DOC/10/ 246, ISBN92-0-155185-1, IAEA, VIENNA, 1985 2. ASTM Standard: C751-92 standard specification for Nuclear-Grade Boron Carbide Pellets

52 THE CONTROL ROD MODELLING CODE REGAIN

J. TRUFFERT Commissariat à l'Énergie Atomique (CEA), Centre d'étude Cadarachee sd , DRN/DEC/SDC/LEMC, Cedex, France

Abstract

We present her CEA'e eth s contro modellind ro l g code REGAIN. REGAINs i applie modellindo t g both normal reactor contro operationd ro l experimentad san l irradiation of absorber pins. It models long permanent periods and transient pin behaviour during shutdown. Calculation performee sar singla r witdn fo e subchannels pi h it , including the cladding, one or more different columns and possibly a shroud. The models cover both thermal, mechanica somd an l e chemical aspecte th f so problem. Written in Ada language this code is built according to an object oriented method; these choices lead "to high modularity, safety and portability. This architecture enables a clean division to specific modules : properties for different materials, thermal analysis, mechanical analysis and mathematical methods. Adding new materials(absorber or cladding) is possible either by inheritance of material properties from parent materia slighd an l t modificatio case th f eo n i r no very different materials(e.g. metal, ceramic, cermet.... and including cracks) by rewriting a specific package unit : specification and body. For boron carbide materia dean ca l t witi l h both existing design concepts: sodium bonded(diving bell)'and helium bonded. REGAIN is thus able to provide details of the absorber material and cladding temperatures, stresses and strains through the life of thé rod. Absorber material is also calculated in some detail, giving porosity, 10B concentration and capture rate.

REGAI beinw no g Ns i license calculatio C e wild usee th d b lSA an r dfo e th f no PHENIX.

1. INTRODUCTION 1.1 The problem At first it is the evaluation of the thermo-mechanical behaviour of an absorber pin depending on : • maximum burn-up of "I OB, • radial variation and, • axial variation of this burn-up, • inlet temperature and flow of coolant. But also evolution of structures for cladding and shroud: • inelastic deformation, • swelling,

53 • irradiation and thermal creep. And evolution of absorber columns: • volume modification by swelling and cracks, • gap closure, • isotopic decrease for absorber materials, • production, retention and release of helium. REGAIN take in charge two pin's architectures : helium bonded (closed pin) and sodium bonded (vented pin). Fobote th r h type, calculation r severao e made on l ar s r pelleefo t stacks( possibly annular pellet) with different diamete enrichmentB 10 r o r .

1.2 Basic assumptions 1.2.1 Geometry and axial transfer remainn Pi s alway axian sa l symmetry, For one level all components are concentric, physical and mechanical variable stemperature: , stresses, strain onle sar y radius dependent. Due to the high thermal conductivity of absorber materials axial thermal transfers are calculated for pellet stack. Ther heaa es i t generatio claddingn i , shrou pelletthermad o n d an t sbu l source in the coolant. Both normal and transient operation are treated this allow to calculate long term experimental irradiation's, operating controls rods with very slorapid wan d vertical movements. In this mode changee th lpelle e th n sti dimension propertied san e sar calculated as the control rod moves slowly or rapidly in the reactor neutron flux, and thes usee e ar calculat do t stressee eth e strain d claddine th san th n i n si d gan pellet (including creation and closure of cracks). The cladding and the pellet are allowe creeo dt p under stres pellee th f si t dimensional change sufficiene sar o t clos initiae eth l gap prograe .Th m also calculate temperaturee sth pellete th f so , cladding and coolant.

2. FIELD Due to modelling limitations REGAIN can be used in the following field: • pellet diameter [ 5 : 50 ] mm • density of 640 [ 80 : 96 ] % • cladding temperature : [ 0 : 650 ] °C • peak linear power : such as maximum temperature in 640 < 1500 C • bum up (low density -84%) [ 0 : 150 ] 1026 capture/m3 • ( high density 96%) [ 0 : 200 ] 1026 capture/mS • lifetime [ 0 :1000 ] e.p.f.d.

54 Buprograe th t heln mdca irradiatio: o t p nu % ] 0 10 : 0 7 [ • densit0 64 f yo • cladding temperature : [ 0 : 800 ] °C • peak linear powe suc: r maximus ha m temperatur 230< C 0 ° 0 84 en i • cladding deformation [0:3]% 300: 0 [ 0 e.p.f.d] . • lifetime

3. MATERIAL PROPERTIES

For standard materials: sodium(and water), stainless steel like 316 (coldworked and annealed) and boron carbide, all the properties are included in REGAIN. s easi It o introduct y e slight modifications from standard material with additional data. materialFow ne r "Ade sth a mechanis f minheritanceo " allow creatioe sth f no f propertieo t se w sne froa mparena t standard material. REGAIN with this property of changeability is very suitable for new absorber pin design(even for reflector replacement). 1 Standar3. d material properties For one typical material the set of properties are bring together in an "Ada package"_with explicit specifications lik en Figuri . e1 This allow simplifications in writing and increase legibility; for example whateve materiae elementare th r on f o l y objet calculatio f thermano l values sha the same form in Ada source language see Figure 2.

. LOCA4 L DESCRIPTION 4.1 Thermal and Physical An elementary physical point is described by the following values corresponding to an "Ada structure" :

For any kind of material TEMPERATUREK - , DIAMETER, LEVEL, THICKNESS, - m kW/m- 3 POWER, n- d SWELLING. TANGENTIAL_CRACKING - AXIAL CRACKING -nd MASSE_VOLUMIQU kg:m/- 3 E HEAT_CAPACIT J/k- - g Y THERMAL CONDUCTIVITY -W/m.K POWER_SECTOR -W

55 And more for steel dp- a DOSE, DOSE_RATE, - dpa/s D_G_ON_DDOSE, -- dérivée versus la dose K_and_ALPHA - irradiatio- ; n creep values

for 640 capt/m- - 3 CAPTURES, s 1/ - - SIGMA_PHI,

DENSITY_for_B4C - nd N_4_HE, N_10_B, N_11_B, N_12_C, -- mole/m3 ENRICHMENTS 0_B, --nd HELiUM_RETAINED - volum /volumn etp e.

for fluids m/- s SPEED, Ps, Pd , -- static and dynamic pressures PRANDTL, REYNOLDS - nd

4.2 Mechanic An elementary mechanical point is described by the following values corresponding to an other "Ada structure" :

RADIUS, - m TEMPERATURE, - K YOUNG, -- MPa n- dPOISSON, ELASTICJJMIT MP- a , YIELD.STRENGTH, - MPa ALFA_T, EPSILON_SWELLING, EPSILON_PLASTIC_and_CREE, ) I , t , r P ( EPSILON_CRACKIN )I , , t , r G ( EPSILON_ELASTId n - )I , , t , r C ( SIGMA ( r, t, I ), - MPa SIGMA_EQ, - MPa K et ALFA -- irradiation creep, EPSILON_PLASTIC_EQUIVd n - - , DG_ON_DT, CRACK_STATU . ) I --boolea , t , r S ( n

56 with CONSTANTES_PHYSIQUES_ET_CONVERSIONS; with PARAMETRES_MATERIAUX , UNITES_ET_TUBES ;

package PROPRIETES_B4Cs i

use UNITES_ET_TUBES;

TABLE_DE_COEFFICIENT:PARAMETRE_MATERIAUX.VERS_COEFFICIENTS;

function CAPACIT ETEMP_( KTEMPERATUR: E retur) n Joule_par_Kg_K;

function CONCENTRATION_10_B TENEUR_10_( B PROPORTIO: N) return RF_STD ; function CONDUCTIVITE ( TEMP_K : TEMPERATURE; DENSITE_REL : PROPORTION ) return Watt_par_metre_K ; function DILATATION ( TEMP_K : TEMPERATURE ) return RF_STD ;

function ENTHALPIE ( TEMP_K : TEMPERATURE ) return RF_STD ;

function MASSE_VOLUMIQU ETEMP_( TEMPERATUR: K E retur) n Kg_m3;

function POISSON (TEMP_K : TEMPERATURE - Utilise si t < T_solidus := CONSTANTES_PHYSIQUES_ET_CONVERSIONS.TkO ) return RF_STD ; function RUPTURE ( DENSITE_REL : PROPORTION ) return MPa ; function YOUNG ( DENSITE_REL : PROPORTION ; TEMP_K : TEMPERATURE - Utilise si t> T_solidus := CONSTANTES_PHYSIQUES_ET_CONVERSIONS.TkO ) return MPa ;

-- Pour gestion des appelS a PROPRIETES_B4C

ERREUR_FORMULATION_B4C exceptio: n; end PROPRIETES_B4C ;

FIG. Boron1. carbide properties package specification.

57 with CONSTANTES_PHYSIQUES_ET_CONVERSIONS; package body XXX_P is

procedure ACTUALISATIONS_MECAN1QUES ( LAJTRANCHE : in VERS_TRANCHE; LEJDISQU VERS_DISQUt ou n i E: s i E ) DIAM : DIAMETRE ; INTERPOLJTHER BAS: E; tk : TEMPERATURE ; densité : PROPORTION ;

begin LE_DISQUE'rangn foi I r e loop declare POINT_MIS_A_JOUR POINT: S renames LE_DISQUE(i; ) begin

RF_STD(POINT_MIS_A_JOUR* DIA0 2. M= : . RAYON; ) INTERPOL_THE INTERPOLATION_T= R: H (DIAM, LA_TRANCHE) INTERPOL_THER.= : tk T; densité := INTERPOL_THER.DENS_RELB4C ; POINT_MIS_A_JOUR.T := CONSTANTES_PHYSIQUES_ET_CONVERSIONS.T_CELCIU; ) k t S( POINT_MIS_A_JOUR. YOUNG := RF_PRECIS( PROPRIETES_XXX.YOUNG densité( ; ) ) k ,t POINT_MIS_A_JOUR.POISSON RF_PRECIS= : ( PROPRIETES_XXX.POISSO; ) ) k t N ( POINT_MIS_A_JOUR.EPSILON_GONFLEMENT - Volumique := RF_PRECIS( gonfl; v) POINT_MIS_A_JOUR.SIGMA_RUPTURE := RF_PRECIS( PROPRIETES_XXX.RUPTURE densit( égonflv}- ; ) POINT_MIS_A_JOUR.ALFA_T := RF_PRECIS( PROPRIETES_XXX.DILATATION( tk ) *CONSTANTES_PHYSIQUES_ET_CONVERSIONS.T_CELCIUS; ) ) k (t end ; end loop; end ACTUALISATIONS_MECANIQUES end XXX P :

FIG. 2. General set up of properties (extract from package body).

58 5. OBJECT ORIENTED CONSTRUCTION Representatio wholbuile s i th n f tn eo witpi h elementary objects. An elementary object is a cylinder made with only one material. Two different meshe usee s ar represen do t t this objec coarsa : t e mesr hfo therma physicad an l l calculation d anothesan r sharp mes mechanicar hfo l description; both have the same axial (possibly irregular) pitch. Optimisation leads to use a quadratic thermal radial pitch(equal section) assorted with a regular mechanical radial pitch(equal thickness). No problem for binding, along radius and axially, using thermal (adiabatic, given flux, given temperatur exchanger eo mechanicad an ) pressure( l , axial force, given axia r radiao l l displacements) interfaces. Any major componen subchane: t cladding, l , (shroud), stack(s internad an ) l gaseous volume is a cluster of same material elementary objets. The whole pin is the binding of the subchannel, the cladding, possibly the shroud and one or several(numbe t limitedno r f stack(s)o ) ; There is no software limits for the total number of elementary objects.

6. CODE DESIGN The main routine manages the assignation of basic files : input data, standard output, supplementary and graphic outputs, it also carry back run time errors at the upper level(monitor). The LECTURE procedure takes in charge the set of elementary tasks: • reading material name and properties, • readin initiaf go l characteristics, • allocation of memory, initialisation of the values, • launc normaf ho transienr o l t operations, • control of output informations, • callin savrecoverf gd o ean routines, • closin programf go .

. CONCLUSION7 S

REGAIN is a light, easy to use, code running on classical work station ( < 10 minut normaa r efo l caslevels0 1 e: stacks 2 , e.p.f.d0 tim0 ,30 5 en .i steps) . sharpnese Th totae f mesth s o l d numbehan f differenro t stacks havo en software limits owindynamie th go t c allocatio f memoryno ( limitatio causens i y db allocated memory space) REGAIN is now being licensed and will be used for the interpretation of experiments and the design of control rods, passive safety devices and core radial subassemblies.

Next page(s) left blank 59 EXPERIENCE WITH CONTROL ROD DRIVE MECHANISM OF FBTR

P.V. RAMALINGAM, M.A.K. IYER . VEERASAMYR , , S.K. GUPTA . SOOSAINATHANL , . RAJAV , N BABU Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Tamil Nadu, India

Abstract

This paper explains the principle of operation of Control Rod Drive Mechanism (CRDM) in Fast Breeder Test Reactor (FBTR) which is a 40 MWt loop type reactor. It discusses the problems faced solutionsand evolved during testing sodiumin ofduring and CRDMand air operationin reactor.in Surveillance tests carried vfiihout CRDMs pile alsoin are discussed.

1. INTRODUCTION

Fast Breeder Test Reactor (FBTR) is a mixed carbide fuelled, sodium cooled, loop type, experimental reacto MW(t0 4 f r abouo ro ) MW(e3 1 t ) capacity Bx 4CSi . control e rodth n si reactor core are held by grippers of individual drive mechanisms which are in-line with the respective control rods and housed in control plug.

The functions of control rods and Control Rod Drive Mechanisms (CRDM) are start-u o havt o t d e• an pcontrolle d shutdow reactoe th f no r • to control reactivity during normal operation of the reactor • to scram the reactor on emergency conditions and havo t e • burn-up compensation.

FBTR has operated upto 10.2 MW(t). In the present core any two control rods out of six are sufficient enoug brino ht reactoe gth r from power leve colo lt d shutdown state.

2. DESCRIPTION

Fig. 1 shows the control rod and its sheath assembly. B4C pellets are in a single pin which is vented having sodiu bondinms a g material. Grippe CRDf ro M hold heae sth controf do l rod. During normal operation of reactor, mobile assembly of CRDM and control rod act as a single unit. Electromagnet (EM) in upper part of CRDM holds the mobile assembly. On receiving scram signal, EM is deenergised and mobile assembly along with control rod falls down under gravity.

l dashpoOi t provides deceleratio f scrao m m travel lasr 0 nfo 5 t . Motor operated screw-nut mechanism up/down moveM E e sth . Grippe operates ri d manually during shutdown conditiof no reactore th .

Primary leak-tightness is achieved by metallic bellows in sodium and secondary leak-tightness achieves i silicony db e bellow ring'O d sealssan . Fig show2 . lowee sth mechanisre parth f o t m 1 with the leak-tight barriers.

61 -6217

-—HEAD -629ÇL

--——SHEATH 6623

-BODY

-7347

jo 20-2/02- — 0

FBTR1. . control Rod.

62 0 RI 3 MBCIWEiS H OUUd 5HIAIO HAH RGB ATMOSPHERE TftAMSlAHON in a ul Inter seal Ka Argon < a o

$* s 0 «INCSJJEtW CO t NPVU Î«O 6 4NO a < C« DM 1.0 WE* PAR I Q oZ •Q'RING BETWEEN GftlPP Gripper Kont roi Jubé CONTROL TUBE AND TRANSLATION TUBE

rranslationBeUows

Gripper Bellows -54/8 i P =120-12« mb f WHEN C.HDH IH FULLY (CROM at bottom limit) RAISED POSITION 4SOmmIt fOSItlOH) A P = »0-88 mb (CROM at 4*0 mm) Gripper Body

FIG. 2. Primary and secondary leak tight barners of crom. 3. TESTINR AI GN I

Eight CRDMs manufactured in India were tested in air prior to their shipment to site. They were again subjecte somo dt e important tests before installatio reactorn i .

followine Th g tests were conducte CRDMn do airsn i :

• Measurement of insulation resistances of electrical items

• Primary leak tightness test using helium leak detecto checo rt k soundnes primaro tw f so y metallic bellows

• Secondary leak-tightness test by drop in pressure to find soundness of silicone bellows and 'O' ring seals

• Measurement of straightness of CRDM with a dummy control rod

• Measurement of scram (i.e., free fall and braking travel) time of the mechanism at different drop height deceleratiod san mobile th f no e assembl dashpoy yb t

• Tests to find gripper torque, gripper operations and corresponding microswitch actuation

Air test results were satisfactory.

4. TESTIN SODIUGN I M

e CRD On testes Mwa sodiun di t t m550°C180°a a d Can . Tests were conducted with misalignment of 0 & 10 mm between CRDM & dummy control rod.

Modification e donb sodiun eo i t d mha s test facilit orden i y o achievt r e requireeth d temperature gradien CRDMn ti .

At 180° followine Cth g measurements were done:

• differential expansion between translation tub grippee& r control tube • translation operation and torque measurement • gripper operation & torque measurement with and without control rod. controe th grippen d I lro d condition, sodium temperatur raises e th e wa 550°Co do t t e Du . leakage of sodium from test vessel at weld region, dished head was replaced. Even though the vesse keps lwa t under argon purging, durinr entrai f yo g dished head replacement e coulb t dno avoided. After soakin vessee gth l with sodiu m400°t a drainind Can purificationr gfo , sodius mwa fille t 180°da temperatur d Can raises ewa 550°Cdo founs t wa dt I . tha controe tth could t lro dno be released to measure the differential expansion between the mobile part and gripper control tube.

Translation and fast drop experiments were carried out at 550°C. After 60 translations and 50 fast drops, interseal argon pressure suddenly reduced indicating leakage. Subsequent interseal pressure tests indicated main bellows failure.

64 Sodiu dumpes m experimentwa fittew d fe dt 180°an da a d Can s were carrie t undedou r argon purging. Control rod could not be released. Sodium was drained and radiography was taken in regioe th controf ngripperso & d lro . Misalignment between gripper bod controd yan head lro d . s 10.7mm wa

Then mechanism was moved and aligned over the control rod and tests were carried out at 550°Ct 180°a d Can . Contro could l ro releasede db . Normal scra mtranslatio& n operation could be performed. Only gripper torque measurement showe increasn da valuese th i eh .

At 550°C in the newly aligned condition with argon purging, 100 translation operations, 100 fast drop operations and 10 translation torque measurements were made. Ah1 were done with failed bellows.

CKDM was cleaned by vacuum distillation and inspected. The outer tube sheath of the mechanism was found to have bent on the cover gas region about 900 mm above sodium level. Bending was spread on the entire cover gas region. Reasons for this type of bending could not identifiee b d specifically.

The test results were satisfactory even with 10 mm misalignment. Also it can be concluded that CRDM can be operated with failed primary bellows for short period.

5. OPERATING EXPERIENCE IN REACTOR

Failure1 5. s leadin safeto gt y related incidents

In April '87 during a reactor start-up, control rod-F continued to raise even after removing "RAISEe th ' command. Operator action immediately brough reactoe tth safo rt e shutdown state by ordering 'MANUAL LOR incidencause e th Th .f analyses eo twa reasod dtracean s nwa o dt sluggish behaviou powee th f ro r contacto "raisingr 1 rfo logi e th cn "i circuit contactoe Th . s rwa replaced. Als preveno ot t recurrenc probleme th f eo inpun a , Reactoo t r Protection systes mwa include ordewheo t R s difference a nr LO th o ds contron ei leveld lro s exceed presesa t value.

Componen2 5. t failures

5.2.1 Increased flow oflnterseal Argon

In 1988 increasn a , e floth wn ei rat interseaf eo l argo observeds nwa . Tests were conducted deteco t mechanise tth m whic developed hha d leaalsd locato kan t leakage eth e sourcee Th . methods followed for leak check included adjusting the interseal argon pressure in a stepped manner so as to reduce the differential pressure with respect to the reactor vessel cover gas and monitoring the flow rate and injection of nitrogen in the interseal argon system and subsequent analysis of reactor vessel cover gas for the same.

Result teste th sf sindicateo d that secondary leak-tight barrier CRDM-Bf so , viz., O-ring seals were leaky and interseal argon was leaking towards reactor vessel cover gas side. Subsequently the O-ring seals were replaced and the interseal argon flow rate came down to background.

65 5.2.2 Failure ofEMofCRDM-C

Juln I y 1991, while checkin CRDMse gth , mobile par CRDM-f o t C droppem dm 0 fro12 m position and repeated operations and checking confirmed that EM showed abnormally high electrical resistance indicating coil inter-turn shorting.

Subsequent remova inspectiod lan electromagnee th f no t coil revealed insulation failure which could be attributed to ageing and a Tctnk found in the terminal leads caused during initial assembly. Exposed strand coie th l f wero s e reinsulate assembledd dan . Before installin reactor fo t gi r service, the mechanism was tested with dummy load and operations were found to be normal.

5.2.3 Failure of Metallic Bellows ofCRDM-B

During April '91, afte MW(t1 r ) operatio reactore th f no significana , t increas intersean ei l argon flow was observed. Leak tests carried out indicated leak in the metallic translation bellows of CRDM-B. The interseal argon supply to the mechanism was kept isolated to avoid loss of argon. Subsequently, primary sodium system temperatur raises ewa 375°do t carryinr Cfo t gou certain test. After lowering the temperature to 250°C mechanism ceased to operate and mobile assembly could not be lifted even manually with a force 1.5 times of what is normally required. Further analysis showed that pressure increase in the interspace of CRDM-B during temperature raisin t relieveggo argoy db n bubbling through sodiu subsequend man t temperature decrease cause vacuuda interspacee th mn i . This resulte entre th sodiuf ydn o i m throug leake hth y bellows into narrow clearance regions of the translation bellows and guide tube sheath and its subsequent solidification. This caused the sticking of the mobile assembly of CRDM-B.

Sinc translatioe eth n mechanis stucks normae mth wa f i , l proceduradoptede b o t s t i ,e wa would have become essential to take out the control rod also along with the lower part. This would have caused wastage of a healthy control rod, since once taken out of the reactor, it is not advisable to reuse it in reactor. To overcome this difficulty, special tooling and procedure was develope lowee th d r dtakeparan s usint twa nou gspeciaa l shielded flask, leavin controe gth d lro inside maximue Th . m radiation fielobserves dwa stellitee th t da d guide bush.

Another spare lower installepars twa reactorn di . Removed lower parkeps decan wa tti t ypi for sufficient period and after decontamination, components of the lower part were cut, disassembled remotely in a dismantling cell for further disposal.

The above incident indicates tha casconfirmen a ti f eo d lea translatiokn i n bellows, interseal argon supply shoulisolatee b t leake dlonno s th d a s yga CRDM insids i reactore eth .

5.2.4 Silicone Bellows Failure in CRDM-F

In the middle of 1994, global increase in interseal argon flow was observed. Tests indicated a leak in the silicone bellows of CRDM-F.

Special tools were mad insitr efo u replacemen leakf o t y silicone bellow lowen si r parf o t CRDM without remova nearbf o l y CRDM mechanisd san m box wore successfulls .Th kwa y carried out insitu in the next shutdown period and later observations showed that interseal argon flow came dow normao nt l value.

66 Miscellaneou3 5. s failures

Itère were many occurrences of stuck closed' position of limit switches used for indication and control logic interlocks. They were minimised by introducing regular preventive maintenance. Events which, can be caused due to such 'stuck closed switches affecting the control circuit logic 1 were identifie modificationd dan s were carrie preveno t t dou t unusual occurrences. occasionsw fe a n O , failuregeae th rsheaf n i drivo sn mechanismf rpi e o s occurred. These were due to over travel beyond the bottom limit and top limit. Although interlocks from limit switches are available to stop the drive motor at the bottom limit, the reason for these failures were mainly disturbance disconectabln si e connectors, deficiencie operatinn si g procedurd ean non-actuation of limit switches. This problem was solved by replacing the connectors with new types which are more reliable, modifying operating procedure and introducing regular preventive maintenance steps.

5.4 Overhauling of upper parts

The preventive maintenance programme includes complete dismantling of the upper parts and inspectio internalf no syearso onctw en .i Dashpo replaces i l toi functiona d dan l check limie th tf so switche carriee sar d out faro uppe.S 3 , r parts have been overhauled casee th l sal uneven I . n wear of teflon pads fitted to the guide tube was observed. The unevenness was corrected by machining.

5.5 Provision of strain gauge

In orde facilitato rt e on-line measuremen frictiof o t n mechanisme forcth f eo straia , n gauge bees ha n provided.

6. SURVEILLANCE TESTS

To ensure proper operation and availability of CRDMs, the following surveillance tests are don t specifieea d intervals:

• Measurement of drop time of the mobile assembly of the mechanism

• Measurement of minimum required current for electromagnet for latching the mobile assembly.

• Measurement of friction force while operating the mobile assembly, with and without control rod.

• Logic checking.

• Integrity checking of CRDM bellows and O-rings.

These check e donar s e accordin e Technicath o gt l Specifications. Test alse ar so done whenever logic modificatio dons ni therr e o deviatio a s ei n observed from normal parameters.

67 7. CONCLUSION

The experience obtained during manufacture, testing in air and in sodium, operation in reactor and surveillance tests of FBTR CRDM enhances the confidence level on overall reliability of shutdown system. This has also helped in design and development of shutdown systems for Prototype Fast Breeder Reactor (PFBR).

68 DEVELOPMENT OF PASSIVE SHUT-DOWN SYSTEMS FOR THE EUROPEAN FAST REACTOR EFR

. EDELMANNM . KUSSMAULG , . VÄTW , H Forschungszentrum Karlsruhe (FZK), Institut für Neutronenphysic and Reaktortechnik, Karlsruhe, Germany

Abstract

This paper presents a simple device for passive shut-down of liquid metal cooled fast reactors (LMR) during unprotected transients uset .I s enhanced thermal expansio controf no drivd lro e lines (CRDL) to provide for automatic scram and forced control rod insertion when pre-fixed coolant temperatures are exceeded. In this way the reactor is brought into a safe permanently subcritical state and temperatures are kept well below the boiling point of the coolant.

A prototype of such a device called ATHENa (German for: Shut-down by THermal Expansion of Na) has been manufactured and tested. The paper presents the principle, design features and thermal characteristics of ATHENa as well as results of the experiments and reactor dynamics calculations of unprotected loss of flow and transient overpower accidents Europeae foth r n Fast Reactor (EFR) equipped with enhanced thermal expansion devices.

1. INTRODUCTION

In hypothetical accident and risk analyses for LMRs it is supposed that the conventional safety syste faiy ml ma completel emergencyn i y situations instancer fo , assumes i .So t i , d that during coolant temperatur r poweo e r transient o lost f primaro e s du s y coolant flowr o uncontrolled withdrawal of control rods either the safety system would fail to activate the rod drop mechanism absorber e nonr so th f eo s would move intcore oth e (contro jamming)d ro l . Then coolant boiling or fuel melting, respectively, with subsequent severe core damage (Hypothetical Core Disruptive Acciden HCDA- t ) would occu presenn ri t design fast reactors with large mixed oxide fuelled cores.

For licensing of future fast reactors it seems necessary to completely exclude any release of radioactivity during a nuclear accident. In principle, this might be achieved in two different ways. The first possibility would consist of definitely excluding core disruptive accidents. othee Th r mighabsolutele th e tb y safe confinemen radioactivf to e materials withi reactoe nth r building. In a more realistic practical solution a combination of these two measures might be feasible, i.e. preventin mose gth t severe core disruptive accident safeld san y confining radioactivity from all others within the reactor building.

essentialo Thertw e ear fulfio st orden i l preveno rt t core meltin disruptionr go :

Reactivity control and sufficient shut-down reactivity must be available under all imagin-

able circumstances even when the conventional "active" safety system fails to shut down the reactor as postulated in unprotected transients, also called Anticipated Transients Without Scram (ATWS).

69 Sufficient heat removal capacity for evacuating decay heat has to be assured even with los primarf so y flo hear wo t sink.

This paper deals wit e firshth t item e ATWSTh . s suc , ULOhas F (Unprotected Losf O s coolant Flow) and UTOP (Unprotected Transient OverPower) are classic initiators of hypothetical core disruptive accidents. Since the failure of the engineered safety system is postulate thesn di e case reactoshue e b sth tn dowca r n onl inhereny yb t passive reactivity feedback mechanisms. The normal reactivity feedback with temperature increase in large LMRs with MOX fuel presently under consideration (e.g. EFR) is too small to outbalance the large sodium density effect and Doppler reactivity during unprotected transients (e.g. ULOF). In present designs reactivity feedback mechanisms cannot prevent the reactor from reaching sodium boiling temperature with the resulting severe consequences.

A detailed stud f thermayo l feedback effect ss show performeha K n thaFZ t tda e mosth f o t phenomena producing negative reactivity (e.g. core expansion, and bowing or flowering) are not easy to increase and would provide only minor improvement. In addition, the resulting reactivity effect e har ar o spredict d t with sufficient precisio d reliabilityan n , whics i h necessary for licensing and public acceptance. Furthermore, all of the inherent thermal reactivity effects have the disadvantage that they are present only as long as temperatures remain significantly increased. As a consequence, the reactor cannot in general be shut down by negative reactivity feedback onln stabilisee ca y b t I . elevatet a d d temperature level thin si s way. However, the reactor has to be shut down finally. Reactor shut-down by reactivity feedback woul possible db e wit h'one-waa y feedback' only. This means thanegative th t e reactivity input cause coolana y db t temperature rise beyon certaida n threshold cannoe b t reduced by a subsequent temperature decrease.

It was concluded from our study that the easiest and most efficient way to provide sufficient negative reactivity feedback in an LMR would be the enhancement of control rod drive line thermal expansion. In addition, one-way feedback can then be achieved which provides for passive reactor shut-dow casn i safetf eo y system failure.

Passive safety features in present fast reactor designs have been discussed elsewhere in more detail [e.g. Generally2] , .1 , most effor concentrates i t corw ene designn do f o m s ai wit e hth increasing negative and decreasing positive reactivity feedback effects with some penalties in cost and performance (e.g. small reactors). Also, in some cases, special devices providing additional negative reactivity feedback are built in. So, for instance, in the Fast Flux Test Facility gas expansion modules (GEM) [3] are used to increase neutron leakage during a loss- of-flow incident. However, the GEMs are not sufficient in large cores and produce negative reactivit losn yhydraulio f so c pressure only.

In other cases, devices for passive rod drop are introduced in the CRDLs utilising the Curie effec magnetin i t c material r enhanceo ] [4 s d thermal expansio ] onl n[5 disconneco yt e th t absorber element from the CDRL. This increases the diversity of the safety system with respec reactoo t r scra t doe mprovidt bu sno e forced insertio absorberf no casn si f controeo l rod jamming e ATHENTh . a device however produce a strons d irreversiblgan e negative reactivity feedbac excessivn ko e coolant temperatures whic sufficiens hi brino t t reactoe gth r into a stable safe state during unprotected loss-of-flow or transient overpower accidents even in case of control rod jamming.

70 2. PRINCIPLE AND CHARACTERISTICS OF THE ATHENa DEVICE basie Th c element ATHENf so showe aar enhancee Fign Th i . 1 . d thennal expansion CRDL consists of two individual (coaxial arranged) drive lines. The "primary" drive line corresponds to the conventional control rod drive line. However, it consists of two separate shafts linked together by a special release mechanism, which is operated by a "secondary" drive line.

Primary Drive Line

Secondary Drive Line Enhanced Expansion Ball Release Module Mechanism

Primary Drive Line (Control Rod)

Fig. 1. Basic elements of ATHENa

71 The complementary "secondary" drive line is an enhanced thermal expansion module, which at normal coolant temperature positioe effec o absorbeth e n n th s o tf snha o r element lowes It . r enlowe e fixes di th o drt primare shafth f o t y drive line, wherea movn uppee ca p sth ed u en r and down according to its thermal expansion.

In ATHEN a hydrauli c expansion modul s usedi e . (Elsewhere bimetallic devices were investigated [6]. t consist)I axialln a f o sy expandable container (metal bellows), whics hi (partly) filled with sodium. The switching temperature at which the thermal expansion switches from norma enhanceo t l adjustee b sodiue n th dca y db m filling sodiue a Th . s mha fairly large thermal expansion coefficient, whic mors hi e than four time valu e largs sa th s eea of the container material (SS). Consequently, with increasing temperature the sodium level in containee th r increase startd san s extendin whegt i n completely filled.

The additional volume of sodium increases only the length of the container, thus moving the movable end of it. This is the upper end of the expansion module as long as the release mechanis s mclosedi . s Whereache ha e uppe th ne d strok th den r e limite e releasth r e mechanis mexpansioopenes e i th d dan directes ni d downwards. The absorbee nth pushes i r d into the core. The elongation of the container depends on the cross-section of its expandable part. Therefore expansioe th , n coefficien containee th f o teasil n ca r multiplie e yb largy db e factor simply b s y reducin cross-sectioe gth bellows it f no s part extremely thin I . swa y large thermal expansion module realisede th b f n so eca . However givea r fo n, maximum internal pressure reducing the bellows cross-section also decreases the force by which the absorber can be pushed into the core.

Under normal operating conditions, i.e t sodiua . m temperature t exceedinno s g maximum design values, the (enhanced expansion) CRDL works as before without any interference from the complementary secondary drive line. However, when the coolant temperature exceeds a pre-fixed level secondare th , y drive line open release sth e mechanism thus disconnectine gth uppe d lowean r r primary drive line shafts from each other. Since o othethern s i er link betwee uppee nth r drive lin e secondareth shafd an t y drive line e absorbeth , r elemens i t disconnected for free falling into the core. In this way a fully passive reactor scram is obtained. absorbee th f I r elemen impedes i t d from frejammingd ero fallino t secondare e th ,gdu y drive line takes over from the primary by partly replacing its lower shaft. For any further temperature increas positioe eabsorbeth e th f no r elemen determines i t enhancee th y db d expansion module, which pushes it downwards with some force to overcome possible mechanical resistance in the absorber channel. The enhanced thermal expansion drive line can be designed in such a way that it forces the absorber far enough into the core to produce sufficient negative reactivit shuo yt t dow reactore nth .

The design of the ATHENa prototype developed for EFR had to fit to the existing above-core structure (ACS) geometry thermas It . l expansion coefficient, reactivity strok mechanicad ean l e choseb strengto t n d sucha h h thae passivth t e shut-down capability under worst-case conditions (EOC, control rods withdrawn) woul assuree db pushind dan g forces significantly higher tha weigh e absorbee nth th f to r element woul obtainee db d without affectin muco gto h the thermal response time of the expansion module. The expansion module is also designed as fail-safe in the sense that any leak from it would cause a rod drop in the same way as excessive coolant temperatures ATHEN] [7 n .I describes ai morn di e detail.

72 The design optimisation of the ATHENa prototype resulted in the following functional parameters:

Switching temperature 50C 0° expansion coefficient 1 mm/°C delatching temperature 505 °C maximum expansion 200 mm thermal response time ~10s pushing force lOkN.

3. EXPERIMENTS WITH THE ATHENa PROTOTYPE

The ATHENa prototype is an experimental device. It is mounted on a test rig and equipped with thermocouple displacemend san t transducer measurinr sfo g internal sodium temperature profiles and thermal expansion. It also has different types of Na level sensors to be tested as a mean f continuouslso y monitorin leak-tightnese gth expansioe th f so n module which proves its availability. This is complementary to the fail-safe feature of ATHENa which should lead expansioe drod th ro t f poi a n module should leak [7]. tes e alsg tri Th o provide simulatinr sfo jammingd gro . Thi achieves si pneumatia y db c device which tend movo st ATHENf lowee o eth d en r a upwards directioe , i.eth n .i n opposits it o et expansion. The resisting force can be adjusted through the air pressure in the pneumatic cylinder. Although the air pressure was kept constant during a single experiment, the counter- force increased with increasing thermal expansion. This is due to the spring effect (20 N/mm) of the métal bellows.

Except for the thermal time constant all essential parameters of ATHENa could be measured by slowl electrin a y n heatini cp u oven t gi . Fig .show2 levea e signalN th sa l f sensoso r (temperature difference between a heated and an unheated thermocouple) and a displacement

Na sensor [°C]

-A AA A

displacement [mm]

temperature [°C]

_L 320 330 340 350 sensoa N displacemen d . 2 ran Fig. t signals versu temperatura sN e

73 transduce e rexpansio th connecte f o e lowe th d nen o rt d module e beginningth t A . t a , temperature , sthes belo°C e0 signalwt chang33 no o sd e significantly with increasing internal sodium temperature. sodiue Ath t C mabou° 0 leve33 t l reache heatee sth d thermocouple th f uppeo e th d t en ea r sodium container. Thi clearls si y indicate shara y db p temperature decrease same Th . e signal would be obtained in case of a leak in the expansion module. At about 345 °C the module is completely filled with sodium and enhanced thermal expansion starts. In these experiments switchine th g temperature from norma enhanceo t l d expansio reduces nwa C abou° y db 0 12 t by compressing the expansion module (metal bellows) in order to avoid deformation during expansions for long periods at elevated temperatures.

In Fig. 3 results of rod jamming experiments are shown for two different resisting forces (150/400 kp). It is seen that the thermal expansion is almost linear with temperature. The

-50— p k 0 .4080 0-

-100--

-150

temperature [°C]

-200 350 400 450 500 550 600

Fig. 3. ATHENa expansion characteristics for two different resistance forces

74 sligh tnon-linea e nonlinearitth o t e rdu changs yi sodiue th f eo m density with temperature. thermae Th l expansion coefficien smalleabous i t% 5 1 t r tha desige nth n valu mm/K1 f eo t I . is assumed that this comes from an underestimation of the effective bellows cross-section t beewhicno nd measuredhha . Impedin expansioe gth constana y nb t resistance only delays onsee th f enhanceo t d expansio t doe t changnbu sno over-ale eth l expansion characteristics. The switching temperature for enhanced expansion increases by about l K per 100 kp resistance.

The thermal time constant has not yet been measured. The necessary rapid temperature increase of about 10 K/s requires a sodium test loop in which LOF-type coolant flow and temperature transients should be simulated for demonstrating the correct functioning of the whole device under realistic accident conditions. For this reason the test rig is already designed for dynamic tests in sodium.

4. CALCULATION OF UNPROTECTED TRANSIENTS IN EFR

For calculating the efficiency of ATHENa devices to prevent severe consequences of unprotected transients in LMRs the dynamic behaviour of the expansion module has to be known. Since experimental results are not yet available thermal hydraulic calculations were performed wit code hth e FLUTA N resultfouns a [8]s wa A .dt i , tha thermae tth l expansios na a function of time during LOF-type flow and temperature variations can be described fairly o first-ordeseriea wel tw y f b lo s r low-pass characteristics. This model together wite hth calculated time constant thes swa ncalculatione useth n di s wit reactoe hth r dynamics code DYANA2 [9]. Details of these thermal hydraulics as well as of reactor dynamics calculations for the "first consistent design" of EFR (1.4 m core height) have been published earlier [7]. Thereafter, extensive calculations were performe latese th r td fo cor e desigR 9/9EF D f nC 0o with l m core height. A comprehensive report on these calculations is given in [10].

followine Inth g resulte somth f thesf eo o s e calculation givene sar e . b However o t s ha t i , pointe t tha thesdn ou i t e calculations ATHEN moss t useit no n dti s a i efficien t way orden I . r to avoid any interference with the first and second shut-down systems of the reactor the switching temperature from normal to enhanced expansion was set to 590 °C. This is 45 K higher than the nominal coolant temperature. Therefore, increased negative reactivity feedback becomes effective only when coolant or fuel temperatures have already significantly exceeded normal operating values in ULOF or UTOP accidents, respectively.

It is assumed that all of the 24 CRDLs are equipped with ATHENa devices but none of the absorber rods woul dcontroo t dro e jammind pdu lro g whic extremels hi y unlikely l calcuAl . - lations were performed with control rod insertes m onlc 0 y1 d (en cyclf do e configuration)t A . these control rod positions the thermal CRDL expansion has the minimum effect on reactivity. Sinc primarR eEF y pumpt havno eo sflywheeld floe sth w halving time durin gpuma p coastdown is only 10 s. This leads to a very rapid coolant temperature increase during ULOF accidents.

In Fig. 4 the normalised primary coolant flow and reactor power during a ULOF at full power are shown. Initially, power decreases slowly due to normal reactivity feedback. About 20 s later enhanced thermal CRDL expansion becomes effective t eveBu .n then power decreases slower than coolant flow which causes a rapid coolant temperature rise. Resulting sodium temperature t channelho e th cort a ,n si e CRD e outleth weln s i L a t s shroua l d tube outsidd ean

75 100

40 Power

20 Time [s]

0 0 20 40 60 80 100 120 140

Fig. 4. Primary flow and reactor power during ULOF in EFR inside t channe ATHENho e th lgive e n sodiuI a ar . Fign 5 ni . m boiling temperatur reaches ei d but in all other channels the peak sodium temperature remains below 800 °C with a large margi boilingno t . However, even under these worst-case conditions sodium boiling can be avoided including in the hot channel. If only one control rod drops the reactor would be shut down and all temperatures kept well belo boiline wth g point. Alternativel woult yi sufficiene db reduco t t e the ATHENa switching temperature by only 10 K. This can be seen in Fig. 6 where the hot channel temperatur shows ei threr nfo e different switching temperatures. Similar resulte sar obtained for slightly increased flow halving times, e.g. 12 s [cf. 10]. Finally, hot channel temperatures might als reducee ob flatteniny db powee gth r distributio coree th f .no

It has to be emphasised that the EFR design is not optimised for the implementation of ATHENa devices seee .b Thi nn frosca m instancer Figfo , 5 . t showI . s thadifference th t e between coolant temperatures at the core outlet and at the ATHENa position in the shroud tub s muci e h bigger than that between shroud tubd ATHENae an eth o t e . du Thi s i s unfavourable flo f EFRo w .S conditionOnlAC e smalya th n si l fractio core th ef no flo ws i passin shroue gth d tubes which contain large quantitie sodiumf so . Thus temperature th , e rise in a shroud tube is strongly attenuated and delayed by coolant mixing. ATHENa would be much more efficien shrouf i t d tube temperatures were clos coro et e outlet temperature.

I UTOna P acciden uncontrollen a t d withdrawa contro a supposed s i f o ld lro . Other thaa n i ULOF the fuel temperature is primarily affected and coolant temperature rise is only

76 1000 JlNa temp.Eoi

--/-^—boiling point 900

\hot ;hannel \ 800 \

core outlet \ X 700

shrojd tube

600 -ATHENa

time [s ]

500 0 14 0 12 0 10 0 8 0 6 0 4 0 2 0

Fig. 5. Na temperatures during ULOF in EFR with enhanced CRDL expansion

1000

900

800

700

600

500 0 20 40 60 80 100 120 140

Fig. 6. EFR hot channel Na temperatures for different ATHENa switching temperatures

77 moderate. Thus, the major concern is fuel melting which may finally lead to pin failure and core disassembly with severe consequences for the environment. Due to the moderate coolant temperature rise reactivity feedback from CRDL thermal expansion is less efficient than in a ULOF. Nevertheless reactivitr fo , y ATHENs ram0/ 1 p o ratet ap su woul d limi e poweth t r transien disconnectiny tb absorbere gth s fro CRDLse mth .

However, fue t channel ho meltin e th l n woulgi preventee db d t leasonla absorbee f yi on t r dropped inte coreoth . Otherwis ATHENe eth a switching temperatur reducee b o t o s dt eha reactivite th r o C y° abou ram0 57 tp rates furthee havb o et r limited. Thi s illustratei s y db Fig. 7 which shows the peak fuel temperature in the hot channel for different switching temperatures ('Tcr' reactivitd )an y ramp casel rates al reacto e n sth I . r stabilisee poweb n ca r d at significantly high levels nominaonl% 0 y 14 (13 lo t power0 controe f noni )th f eo l rods drops into core. A single rod drop would stabilise the reactor at a power level and coolant temperature below nominal values. The reactor will be shut down when at least two absorber rods drop.

ULOe case th f th eAFo n si flattenin powee th f go r distributio improved nan d heat transfer between core outle shroud an t d tubes woul vere db y helpfu preventinn i l g fuel meltind gan would enable the ATHENa devices to bring the reactor into a permanent subcritical state even if all control rods were jamming. This would provide for more flexibility in reactivity ramp rate ATHENd san a switching temperature.

2,700-1 Fuel Melf ng

2,600-

^2,500-

— 0.25$/s, Tcr=580 deg.C "E 2,400- o. E 0.5$/s— , Tcr=580 deg.C

2,300- -• Q.5$/s, Tcr=59Q deg.C

— 1 M/s, Tcr=570 deg.C 2,200- 1.00/s, Tcr=580 deg.C

2,100- T t I I \ \ I i i 0 14 0 12 0 10 0 8 0 6 0 4 0 2 0 160 180 200 Time (s)

Fig . Maximu7 . m fuel temperatures during UTOPwitR hEF ATHENn si a

78 5. CONCLUSION

ATHENa a prototyp, e passive shut-down a coolesysteN r mdfo fast reactor s beeha s n developed and tested. It utilises enhanced thermal CRDL expansion at excessive coolant temperature disconneco st controe th t l rods fro drive mth epuso t lined h an sthe m inte oth core. It was demonstrated that such a device can be robust and reliable in providing either passive rod drop or forced insertion of control rods at a thermal expansion rate of about 1 mm/K against a resistance of 10 kN. Reactor dynamics calculations have shown that in this way sufficient negative reactivit s produceyi preveno dt t sodium boilin d fuegan l melting during ULOF and UTOP accidents in EFR under the most pessimistic conditions.

Thi trus si e eve desigR n implementatioe thougEF nth e th n hi ATHENf n o t foreseen no s awa . futurdesignn IR i f LM e s thi s properli s y taken into account HCDAs during unprotected transients in fast reactors might be definitely excluded even for large cores with mixed oxide fuel. REFERENCES

[1] Proc. IAEA Spec. Meeting on Passive and Active Safety Features ofLMFRs, IWGFR, O- arai Engineering Centre, PNC, Japan, (1991) [2] D. M. Lucoff, A. E. Waltar, J. I. Sackett, K. Aizawa, 'Experimental and Design Expe rience with Passive Safety Feature Liquif so d Metal Reactors', Proc. Int. Conf. Design Safetyd an of Advanced Nucl. Power Plants, October 25-29,1992, Tokyo, Japan (ANP '92) . WaldoB [3. J ]Padilla. ,A Nguyen. , H Jr. Claybrook. . ,D W . S , , 'ApplicatioM GE e th f no Shut-down Device to the FFTF Reactor', Trans. ANS 55, 312 (1986) . BojarskyE ] [4 Müller. K , Reiser. H , , 'Inherently Effective Shutdown System with Curie Point Controlled Sensor/Switch Unit', KfK report 4989, Kernforschungszentrum Karlsruhe, (1992) . Bilibin[5K ] , TSfuclear Reactor Shutdown Contro Assembly'd lRo , U.S. Pat. . 4,734,252No (1988) . Millington N Reacto R . D EF ] e r[6 Protectio,Th n Syste Third man d Shutdown Syster mfo Risk Minimisation', in 1. . Edelmann[7M ] Baumann. ,W . Bertram,A Kussmau. G , Väth. W d , 'Enhancean l d Thermal Expansion Control Rod Drive Lines for Passive Shut-Down of Fast Reactors', Nucl. Technology, Vol. 107, (1994) [8] H. Borgwaldt, W. Baumann, G. Willerding, 'FLUTAN - Input Specifications' , KfK report 5010, Kernforschungszentrum Karlsruhe, (1992) [9] V. Ertel and R. Reinders, 'Development and Application of the Thermohydraulic System Code DYANA / ATTICA', Proc. Int. Conf. Science Technologyd an of Fast Reactor Safety, Vol.2, Guernsey (1986) . Kussmaul[10G ] , 'Influenc Controf eo Enhanced lRo d Expansion Device Course th n sf o eo Unprotected Transient EFR'n si , KfK report 5370, Kernforschungszentrum Karlsruhe, (1994)

Next page(s) left blank 79 DESIGN PHILOSOPHY OF PFBR SHUTDOWN SYSTEMS

. RAJAV N BABU . VIJAYASHREER , . GOVINDARAJANS , , G. VAIDYANATHAN, G. MURALIKRISHNA, T.K. SHANMUGAM, S.C. CHETAL . RAGHAVANK , , S.B. BHOJE Indira Gandhi Centre for Atomic Research, Departmen f Atomio t c Energy, Tamil Nadu, India

Abstract

This paper presents the overall design philosophy of shutdown system ofSOOMWe Prototype Fast Breeder Reactor (PFBR). It discusses design criteria, parameters calling for safety action, different safety actions and the concepts conceived for shutdown systems. In tune with the philosophy of defence-in-depth, additional passive shutdown features, viz., Self Actuating Device (SADE) Curieand Point Magnetic (CPM) switch protectiveand feature like absorber Strokerod Limiting Device (SLD)contemplated.are alsoIt discusses about suitability Expansion ofGas Modulesafetythe of devices (GEM)one PFBR.as in

1. INTRODUCTION

Prototype Fast Breeder Reactor MWe(PFBR0 50 a , s mixe)i d oxide fuelled, sodium cooled pool type reactor t consistI . primaro tw f so y sodiu secondaro tw md pumpan yfouX d san IH r loops each loop having one sodium pump and four integrated steam generator modules. Two important aspects on which safety of the reactor depends are the reliability of shutdown systems anreliabilite dth f decayo y heat removal systems l engineereAl . d safeguard wele ar s l tuned towards this objective. Overall reliability of shutdown systems, of course, depends upon the well conceived design, manufacture, quality control, prototype testing, on-line monitoring and surveillance.

2. DESIGN CRITERI SHUTDOWR AFO N SYSTEM

Broad guidelines for design of shutdown system are as follows:

• Atieast two reliable, independent, automatic, fast acting shutdown systems shall be provided operatin n diverso g e principles e systemth . f Atieaso se shalon t l mee l functionaal t l requirements even in case of postulated core deformation. The reliability of each system shall be such tha s non-availabilitit t s lesi y s than 10" r reacto3pe e overalrth yead lan r non-availability of the two systems shall be less than 10"6 per reactor year.

• The design shall provide sufficient redundancy so that failure of a single most effective absorber rod of a shutdown system shall not result in impairment ofthat system to an extent tha wilt ti t mee minimule no tth m specified requirement negativf so e reactivity. thf o e e shutdowOn • n systems coul usee dreactivitr b dfo y control. However, while doin, gso its functional capability to shutdown the reactor shall not be jeopardised.

81 reactivite Th • y worth, spee actiof do delad n an actuatio n yi eacf no h shutdown system shall be such that durin operational gal l state postulated san d accident condition reactore th f so , including the most reactive state of the core,

- the reactor is rendered sufficiently subcritical and maintained subcritical under cold condition, taking into account uncertainties in the neutronic calculations/measurements, specifiee th - d fuel design t exceeded limitno e sar , - the reactor coolant system design limits are not exceeded.

• The total reactivity worth of the shutdown systems shall be such that in the shutdown state, wit absorbel hal rcore e rod reactoth e n s,th i r shal subcriticae lb l with k^not more than 0.95 such that the reactor remains subcritical under postulated fuel handling errors (e.g. replacemen mose th f t o treactiv e absorbe mosy b d t reactivrro e fuel sub-assembly, removal of absorber rods). availabilite Th • safetf yo y support systems necessar actuatior yfo shutdowa f no n system shall be commensurate wit availabilite hth y requirement shutdowe th f so n system.

• All equipment shall be designed such that its probable failure modes will not result in an unsafe condition.

• The design shall be such that all maintenance and availability testing which may be required during reactor operation can be carried out without a reduction in the effectiveness of each system below the minimum allowable requirements. desige Th n shal• suce lb h that each shutdown systeactuatee b n mca d manually fro maie mth n and emergency control rooms.

• The design shall be such that it is not readily possible for an operator to prevent a safe automatic action from taking place.

• The control logic of the absorber rods and their drive mechanisms shall be designed to prevent unintended movemen directione th n i t s whic reactivityd had .

• Maximum reactivity wortabsorben a f o h r rod, together wit s maximuit h m possible withdrawal speed, shal limitee lb d such tha fuele tth , coolan claddind an t g design limite sar not exceeded in the event of uncontrolled withdrawal of the rod.

3. CONCEPTS CONCEIVE SHUTDOWR DFO N SYSTEM

Reactor safety is assured by two independent, fast acting, diverse shutdown systems, each comprisin f sensorsgo , logic circuits, drive mechanism absorbed san r rods. Type-A system consists of a bank of 9 absorber rods while the Type-B system consists of a bank of three absorber rods. Type-A is for reactivity control as well as for reactor shutdown, whereas Type-B is only for reactor shutdown. Each rod is operated by an individual mechanism in-line with the rod. Their disposition is shown in Fig. 1. In both the systems, shutdown of reactor is achieved by droppin absorbee gth r rod gravityscray e sb th t m releasBu . e Electromagnet (EM Type-f o ) A syste mhouses i uppedn i r par mechanisf o t m controabovf o p e to argoln i plu d ngan atmosphere, whereas tha Type-f o t B syste mechanis f lowet o a md s i ren m immerse primarn di y sodiut mho pool near the top of sub-assemblies. Details of these shutdown mechanisms and absorber rods are dealt with separately in another paper [1].

82 MASS PER TYPE OF SUBASSEMBLY No. SYMBOL :SUBASSY. IN Kg $ FUEL (INNER) 85 245 O FUEL (OUTER) 96 245 » '"PRIMARY CONTROD RO L 9 200 o SECONDARY CONTROD RO L 3 200 ® BLANKET 186 320 ff*\ STEEL REFLECTOR (INNER) 72 355

$ C SHIELDIN4 B G (INNER) 69 185 ® STORAGE LOCATION 75 245 - $ RESERVE STORAGE LOCATION 24 355 ^ ENRICHED BORON SHIELDING 56 185

FIG. 1. PFBR core configuration.

83 Triplicated sensor usee sar measuro dt e parameters importan reactoo t t e r ar safet d an y connected by a '2 out of 3' coincidence logic to the reactor protection system. A hot standby channe l' coincidenc 3 aid f maintaininn so i t ou 2 ' ee logicgth . Selectio parameterf no bases i n do the well developed concepts of diversity. Physical separation of redundant channels of the safety syste mproposeds i routine Th . redundanf go t signals wil physicalle lb y independent.

Two independent Reactor Protection Logic Processing Systems are provided. One system will be base harn do d wired solid state logic circuits workin pulsn go e coding mode, wherea othee sth r system will have either rela ymicroprocessoro logitw r co achine son g standb othee th r o yrfo t logic functions. Protection systems will have facilit on-linr yfo e testing, wherever needed, without causing safety action.

4. PARAMETERS CALLING FOR SAFETY ACTIONS & TYPES OF SAFETY ACTIONS

Depending upo nature nth faultf eo reacto e th , r protection syste mdesignes i initiatdo t o etw type f safetso y action reactore th n so graduaa ; l shutdow fasa d t nshutdownan graduae Th . l shutdow effectens i Loweriny db Rodf go s (LORfase th t d shutdow)an n (Scram effectes )i y db simultaneously droppin absorbee th l gal r rods from their existing positions int reactoe oth r core. The important parameters callin safetr gfo y action givee sar Tabln i . eI

TABL . EPARAMETERI S CALLIN SAFETR GFO Y ACTIONS

No PARAMETER SAFETY ACTION LOR SCRAM 1 Ffigh Neutron Flux in start-up range - yes 2 Short perio start-un di p range - yes 3 High Neutronic Power (Log P) - yes 4 Ffigh Neutronic Power (Li) nP - yes 5 Short period in power range - yes 6 Ffigh reactivity in core - yes 7 Deviation of temperature rise for each yes yes subassembly from the calculated temperature rise for that power 8 Deviation from mean core outlet temp. yes yes 9 Deviation from mean gradient temperature yes yes 1 0(bulkFueo D t e l DN cla)du d failure - yes 11 Earthquake horizontal acceleration - X component - yes componenY - t - yes 12 Power/Primary flow rate yes yes 13 Primary pump trip yes - 14 Secondary pump trip yes - 15 Feed water pump trip yes - 16 Los electrif so c power yes yes 17 Turbine trip yes - 18 Water/steam leak into sodium in SG yes -

19 OR ineffective ™* yes

84 5. DEFENCE-IN-DEPTH

observeIs ti d that among various incidents leadin Losgo t Flof so w even Transiend an t t Over Power event, loss of off-site power and inadvertent withdrawal of an absorber rod are respectivel mose yth t probable event thed syan lea severo dt e consequences thef ,i couple e yar d with simultaneous failure of both shutdown systems, i.e., Unprotected Loss of Flow (ULOF) and Unprotected Transient Over Power (UTOP) events. ULOF and UTOP are now considered as Beyond Design Basis Events (BDBE).

Wit site Kalpakkamhs th e a , evaluation carrie indicatet dou s probabilit gri0 d12 supplf yo y failurduratioh l gri 5 f eo dd nsupplan yduratioh failur 5 life 1 eth plantf e e etimo n th th i f n eI .o philosoph defence-in-depthf yo reactoe th f i , protecte s ri d from this inciden passivy tb e means, then the overall unavailability of reactor protection system can be reduced to less than 10" per 7 reactor year.

Los1 5. Off-sitf so e power

Reactor must be shutdown, in case of a confirmed loss of off-site power (i.e., duration achievo T >2.. s) 5e safe shutdow passivy nb e means during this event followine th , g three concepts have been studied:

• Gas Expansion Module • Self Actuating Device • Curie point magnetic switch.

5.1.1 Gas Expansion Module (GEM)

GEM essentialls i ypassiva e shutdown devic insereo t t negative reactivity durin primarga y system device LOFTh basicall.es i holloya w removable sub-assembl oped an n p y to sealee th t da bottome atth argon trappeA s . nga d insid sub-assemble eth y expands when core inlet pressure decreases due to flow reduction and expels sodium from the sub-assembly. The effect of putting GEM in PFBR at the interface of core and radial blanket giving 1.5 $ of reactivity on voiding of GEM have been studied theoretically. The studies have clearly shown that for a flow coastdown incidenflos 2 w1 halvinf o t g time constant reactoe th , passivels ri peae yth ksafd coolanean t temperature reached is 903 K (630°C).

Even though passive achieve e safetb n yca d Sub-assemblie witM hGE s agains ULOn a t F event with flow halving time of a few seconds (and hence reducing the need for high moment of inerti primarr afo y pum wheels)y pfl , they hav followine eth g disadvantages:

• Inservice inspection of GEM is difficult to check argon pressure and sodium level. If argon leak the frot M nsou m GE sodiu m level fallconsequend san t negative reactivity durinF gLO event is not assured. The knowledge on dissolution and leaks of argon in sodium is not extensive.

• Noise in reactivity may be introduced due to fluctuation of sodium level in GEM sub-assembly.

• Argon under high pressur entey centrae e ma positivr th d ad core l d parth ean f reactivityo t .

85 • At low power operation, speeding up of pumps would add positive reactivity ( > 1 $). reactivitM GE y• worth depen sodiun do m s pressur it leve d an l e which varies with flow. GEMs are not effective below 25% flow.

• Operating experience is minimum.

t favouredno s i HencM . GE e

5.7.2 Self Actuating Device (SADE)

SADE, proposed by France [2], is a motor-generator system with a flywheel on the same shaft. Motor is operated by off-she power supply. The generator supplies power to scram release shutdowf o M E n mechanis casmn I losf e o off-sitf so e power flywheespeee e th ,th f do l reduces and als voltage electricae oth th f eo l generator. Afte short-timea r , this voltag moro n s ei sufficienreleaseM E d absorbee san th t r rod. Direct aching SADE syste mproposes i givo dt e supply to EMs of Type-B shutdown system absorber rods.

5.7.3 Curie Point Magnetic (CPM) switch

switcM CP h deenergises scram releas whe M demagnetises ei E nt i reachinn do temperaturga e of 873 K (600°C) under power excursion, thereby scramming the reactor [3]. Initially it was considered to have CPM switch as part of Type-B shutdown mechanism at its lower end. Refer Fig. 2.

Studies were carrie fino t t d dresponsou eswitche timth f eo , fro time m th whict ea h sodium temperature exited from fuel sub-assembly starts increasing (du time LOFo eth t t whic e a o t )e hth temperature reaches curie point (873 K) of the magnet. Fig. 3 shows the behaviour. Since the (dependins 6 3. responso t s betwee gp 6 ordee upoga 2. eth e f magnetie f timrno th o n th s ei c switc outed han r sheatmechanism)e th f ho decides i t switc,i M locatdo t CP h e juseth t abovea fuel sub-assembly. Studies are under progress to position it at proper location in control plug.

Inadverten2 5. t withdrawa absorbef lo d rro

Type-B rods coursef o , considered,e neeb t dno , since the poisee yar d outsid core eth e during normal operatio reactorf no Type-t .Bu considerede Ab o rodt e sar , since the partialle yar y inside core. They can not move w.r.t. the mechanisms, as they are firmly held with the mechanisms. But mobile assembly along with absorber rod could raise up inadvertently due to failure of logic circuit or due to operator error. Hence it has to be prevented from raising up by a physical stop which is known as Stroke Limiting Device (SLD). SLD which can be adjusted based on the power level of the reactor, is provided in each Type-A mechanism. Proper care is taken in the design so that SLD no way hinders the safety (LOR & Scram) functions of the shutdown mechanisms.

86 MAGNETIC SWITCH

-SENSOK TH 5 R1. ALLOe F % Y5 3 . 8 i N 65%

UNION JOINT

23325

ELLECTROMAGNET

23200

23125

HEA ABSORBEF DO D RRO

FIG. Magnetic2. switch electromagnetand assembly.

87 T EO A ATS/

AT SENSOR LLVEL

aoo -, AT SENSOR (CONF1G) 1 .

—————— AT SENSOR (CONFIG. 2)

750 H

700 -

650 -

600

550

6. CONCLUSION

A systematic approac bees hha n followe desige th n shutdowf di no n syste PFBRmf o e Th . design criteri bees defence-in-depta ha ns a wel d lan lait dou h philosophy, passive featuree sKk SADE, CPM switch and SLD have been incorporated to enhance the overall reliability of the reactor protection system.

REFERENCES

[1] Rajan Babu,V., et al., "Design of shutdown system for PFBR", This Technical Committee Meeting. [2] Bertagnoho, D., Mallet, Y., Barbet, M., "General Philosophy of RNR-1500 shutdown system: an example of a third line of defence - the SADE system", (Proc. Int. Conf on Science and Technolog Fasf yo t Reactor Safety, BNES, London, 1986) 235-238. ] Bojarsky[3 t al.e , "InherentlE. , y shutdow R SafSN e n system with curie point controlled sensor/switch unit" (Proc. Int. Conf Sciencn o . Technologd ean Fasf yo t Reactor Safety, BNES, London, 1986) 153-157.

88 DESIG SHUTDOWF NO N SYSTE PFBR MFO R

V. RAJAN BABU, R. VIJAYASHREE, R. VEERSAMY, L. SOOSAINATHAN, GOVINDARAJAN, S.M. LEE, S.C. CHETAL

Indira Gandhi Centre for Atomic Research, Departmen f Atomio t c Energy, Tamil Nadu, India

Abstract

This paper presents principlesthe adopted designthe in independent, oftwo fast acting, diverse shutdown mechanisms theirand absorber rods Prototypefor Fast Breeder Reactor (PFBR).h describes features of the shutdown mechanisms theirand absorber rods, passive features, present status in R & D and program on passive and protective feature as a measure of defence-in-depth, testing and verification. 1. INTRODUCTION

Prototype Fast Breeder Reactor (PFBR) has two independent, diverse, fast acting shutdown systems workin fain go l safe mode. Function shutdowe th f so n system Type-e Aar

- to start-up and have controlled shutdown controo -t l reactor power - to scram the reactor on emergency conditions - to have burn-up compensation

Functio Type-f no B syste mlimites i scrado t reactoe mth emergencn ro y conditions.

Nine Type-A rods in two radial banks facilitate adjustment of radial form factor by differential insertion of rods. Out of nine rods, three are in inner PCD and six are in outer PCD. Type-B system has three rods in inner PCD. In order to avoid common mode/cause failure and achievo t e diversity desige th , botf no systeme hthath y sucn i t wa parts shi a s relate safetdo t y action of absorber rod under scram action are entirely different.

2. DESIG ABSORBEF NO D RRO

For both the systems B4C is chosen as absorber material because of its relatively high neutron cross section, commercial availability, ease of fabrication, low cost and good operating experience in Fast Reactors. B10 enrichment is kept in the range of 50 - 70 % so as to have a margin for extra reactivity requirement futura sn e without changing numbe rodsf ro .

1 Reactivit2. y Worth

Reactivity wort absorbef ho r rods mus adequate tb satisfo et followine yth g criteria:

• Adequac shutdowf yo n margin (SDM handlo t ) e postulated incidents o likt e e LOFdu P ,TO uncontrolled full withdrawa mose th f t lo othereactiv y an ro t reason ee absorbedu r ,o d rro

89 fuel melting and slumping in a few sub-assemblies, sodium boiling and voiding in a few sub-assemblies etc. • Adequacy of reactivity worth in each individual shutdown system to bring the reactor to cold shutdown state assuming that other syste mose th failes e md tha th reactivan d f o d ero working system is also stuck. • Adequacy of SDM k the fuel handling state to cater to postulated errors such as replacement of most reactive rod by a high enrichment fuel SA during fuel handling, accidental withdrawal of two absorber rods etc.

Worth of individual rod is relatively small, so that flux distortion and shadow/anti-shadow t largeeffectno e . ar s

2.2 Raising/Lowering Speed

With the raising speed of Type-A rod being 2 mm/s, the maximum rate of reactivity addition limites i pcm/s2 o dt . This enables smooth start-u alsd opan leaves adequate margi protecno t t the reactor in the event of an uncontrolled withdrawal of a rod. In the case of Type-B rods, a speed of 10 mm/s or less is acceptable. Speed of travel of 4 mm/s is selected. Considering safety and simplicit designyn i , raisin lowerind gan g speed individuaf so l syste made mar e same.

2.3 Mechanical Design

Stationary hexagonal wrapper tube of absorber rod is like any other fuel sub-assembly externally, whereas mobile outer sheath of absorber pins is cylindrical, hi each Type-A rod, there are 19 phis each having B4C pellets in clad tubes. Fig. 1 shows the Type-A absorber rod and its wrapper tube. Pins are vented having sodium as bonding material Wrapper tube, outer sheath and clad are all made of 20 % CW D9. Type-B rod has similar design except having comparatively more clearance between absorber rod outer sheath and wrapper tube. It also has a swivel joint between head and stem of the rod.

4 Thermal-Hydrauli2. c Design

Sodium flow hi absorber sub-assemblies is such that

• Maximum cladding mid-wall temperature shall not exceed 953 K (680°C) under normal operating conditions when the rods are hi withdrawn position sodiue Th t spomho • t temperature whic between i s hi pelle C claddin4 d nB an t g shalt no l exceed the boiling point during the fall of one absorber rod into the core when the reactor is operating at full power under new equilibrium configuration.

3. DESIGN OF SHUTDOWN MECHANISM

Type-A and Type-B mechanisms are in-line with the corresponding rods in core and are housed in control plug (CP). There is a mechanical coupling between mobile assembly of Type-A mechanis rodd , man whereas therelectromagnetin a s ei c coupling between Type-B mechanism and rod. The mobile assembly of Type-A is held by an electromagnet kept above the top of CP. Schematic arrangement of Type-A and Type-B mechanisms are shown hi Fig. 2 & 3.

90 ••e

—JIOO ClflCUU»! UK COHUC— - T OH 5HIA1H

P3 P2 P1 131X MA .S F HEXA/ . CONTROL ROD POSITIONS SECTION-DO P1-CR RESTIN SHEATN GO H P2-CR HELP IN CRDM AT BOTTOM POSITION. P3-CR HEL POSITIONP CRDN DI TO T MA -

SECTION-XX PFBRFIG.1. primary control rod.

91 I*. OR

VO K) SYNCHRO

TRANSLATION SCREW MOTOR

LOAD GEH.

TORQUE UUfTER

TRANSLATION SCREW ELECTROMAGNET ROTATIVE POTENTIOMETER NUT GUIDE COLUMN

MOBILE ASSEMBLY GUIDE COLUMNS OASHPOT TRANSLATION TUBE CONTROL PLUG TOP /{/7//// y//r//

CONTROL PLUG TOP

TUBE SHEATH 27«» DIUM LEVEL

GUIDE BUSH IN SODIUM ELECTROMAGNET

GRIPPER CONTROL ROD SHEATH

ABSORBER ROD

FIG. 2. Type-A mechanism. FIG. 3. Type-B mechanism. Two Plant Protection Logic provide scram signal correspondine th o st g shutdown systems separately. On receiving scram signal, electromagnets (EM) of both the systems are deenergised. Hence in the case of Type-A system, mobile assembly of the mechanism along with absorber rod is released to fall under gravity; but in the case of Type-B mechanism, only absorber rod falls under gravity. Mor core th eef heigho tha 3 n 2/ covere s ti free rod e theth e d th y fal dnsb f an o l the decelerate e case yar Type-th f e o n dashpotl i oi ) y Adb CP systef so (providep d mto an e th t da sodium dashpots (provided in wrapper tube of absorber rod) in the case of Type-B system.

Screw-nut mechanism is used in both the systems for raising and lowering of mobile assembly. Since it is not participating in scram action, common mode/cause failure is not of specific concern here.

In Type-A mechanism, positions of EM and mobile assembly are indicated by synchro resob/er and potentiometer respectively, whereas in Type-B mechanism only potentiometer indicates positio mobilf no e assembly. Microswitche sproximit/ y switche usee contror sar dfo l of interlocks.

Type-f Sinco M eAE temperatur w systelo t a argo mn s i i d nean atmosphere carbow ,lo n steel uses i magnetis da Type-f co materialM BE t systeBu . mimmerses i t pooho ln d i sodiu d man directly holding the head of the absorber rod. Hence pure iron is considered as material for core of the magnet, whereas 2.25 Cr 1 Mo steel is considered for head of the absorber rod to avoid self weldin similaf go r metals.

Bot e mechanismth h e hermeticallar s y seale o achievt d e leak-tightness even during hypothetical Core Destructive Accident (CDA) and during Safe Shutdown Earthquake (SSE).

Temperature at top of control plug (CP) is maintained at 110°C so that sodium vapour deposition doe hindet sno free rth e movemen mobilf to e assembly. 'O' ring seal usee sar d wherever therrelativo n s ei e movemen wherr to e ther slos ei w relative movement betwee parte nth s whose movement is of no safety concern. Lip type seals like 'V ring seals or oil seals are proposed to be used between mobile assembl stationare a s th a d t yan ac yo t tub p eto sheatP levee C th f t lo h a barrier restricting sodium vapour coming into upper part. Positive pressure in hermetically sealed upper part w.r.t. reactor cover gas facilitates to have flow of argon from upper part into the reactor vessel in case of leakage of seals i.e., from non-active to active region. Argon gas supply to the upper part is bubbled through NaK to avoid oxygen entering into the system.

Major concern on the overall reliable performance of the mechanisms with the rods is the proper functionin f scrago m dashpofrictiod e releasth an d M nan te E between stationard yan mobile parts.

4. DESIGN OF CONTROL SYSTEM FOR SHUTDOWN MECHANISMS

Two independent and diverse control systems are proposed. For Type-A mechanism, Programmable Logic Controllers (PLC) wil usee lb d considerin importance gth large th ed ean numbe controf ro l components Type-r fo t B.Bu mechanism, conventiona relayM lE s wil usede lb , sinc numbee eth controf ro l component relativele sar y less. Triplicated fail-safe control logic will be used to improve the reliability of PLC.

93 5. PRESENT STATUS IN R & D

plannes i t testino I d sub-assemblieo dn t g o botf smechanismo e hth s whicsafetf o e yhar concern. In this context 1:1 scale model of EM and dashpot of both the systems have been planned. It is also planned to do experiments on seals and sodium vapour deposition in annular gap between mobile and stationary parts.

1 Scra5. m release Electromagnet

f Type-o M AE mechanism carbo,w madLo f no o efin t s steel it s dr teste ai wa , n di characteristic loan so d carrying capacity, response tim temperaturd ean e o riset s .ha ThiM sE handl loaa e f aboudo t 500argon i 0N n atmosphere. Result loaf so d carrying capacitd yan temperature rise of EM are well satisfying the design requirements; but the response time is more largo t thae especifiee du eddn th , yms currend0 limi10 f to t induce . DesigEM dn ni modifications are being don reduco et e eddy current.

Firs ascertaio t t Type-n i desige M nth E B f no system scal1 1: , e witmodew M E hlo f o l carbo testes witr n ai stee wa h absorben d i o différen an lM 1 head r rro C dt mad5 gap2. f seo between the contact faces. Increased air gap necessitates more excitation current. If the actual air ga less pi s compare assumee th do t d gap, then that will cause large magnetic holding force leading to high magnetic response time. The results are satisfactory. It is planned to test EM made of pure (550°CK 3 thed 81 steelo n )an t irohea d a M r 1 n,absorbef an d firso ai r C n i t mad5 d 2. r ro f eo in sodium at different operating conditions.

5.2 Dashpot

1:1 scale model of oil dashpot of Type-A system and its test set-up are being fabricated to find characteristics of dashpot, viz., velocity and deceleration of mobile assembly and piston and oil pressure with respect to time. Weight of mobile assembly falling on piston of dashpot is about 4500 N. Dashpot travel is 250 mm. Design modifications will be done based on the test results.

It is planned to test 1:1 scale model of Type-B absorber rod and the wrapper tube first in water at room temperature and then in sodium at reactor operating temperature. Weight of absorber rod considered is 400 N.

3 Seal5. s

Lip seals like oil seal and V ring seal are being tested in a set-up with 1:1 scale model. Tests proposee ar done b statido en t i c conditio dynamin i d an n c reciprocating conditio mobile th f no e assembly. Leak rate and friction while raising up and lowering down are studied. Test results are encouraging presene th n .I t studies, various configuratio seale beine th f sar n o g teste findo t n da optimum one. Then endurance tes differenn to t aged seals wil done l b scrar efo m actio mobilf no e assembl mechanisme th f yo .

94 4 Sodiu5. m vapour deposition

Sodium vapour depositio annulue th n ni s betwee mobile nth stationard ean y parts with different gap convectiod san n barrier beins si g studie tesa n dti set-up. tesBasee th t n resultsdo , annular gaps in the mechanisms will be properly modified. Here temperature of the mechanism is to be maintained in such a way that the top of CP is at 110°C.

5.5 Prototype Testing

It is planned to test prototype mechanisms with dummy absorber rods of both the types of system thed sodiuan n i s r firsai mascertain o i t overale nth l functionin systeme th l f al go t sa operating conditions. Fabrication of prototype systems is in progress.

6. PROGRAM ON PASSIVE AND PROTECTIVE FEATURES

1 Curi6. e Point Magnetic (CPM) switch

In order to improve the reliability of shutdown systems during Loss of Flow (LOF) even I. Curie Point Magnetic (CPM) switch which cuts poweofe fth r whesupplM sodiue E nth o y t m near the switch reaches 873 K (600°C) is considered. As mentioned in [1], studies are undei progress to position it at proper location in CP.

6.2 Stroke Limiting Device (SLD)

Inadvertent withdrawa absorbef lo durind rro g operatio considerereactof e no b o t s ri r dfo Type-A rod, since it is only partially inside core. In control logic, it is provided to restrict the upward movement of mobile assembly only a few mm continuously. But if there is a fault in logic circuit, the woulnt i eventP d avoileao TO T . do t d this physicaa , l stop adjustewhice b n hca o dt any specific power level by administrative control, is provided in Type-A mechanism. A mechanical device operated on screw-nut principle is coupled in parallel with the motor drive-screw assembly. traveBasee th n ldo positio mobilf no e assembl hencd yabsorbean e eth r rod, the position of stopper wfll be changed. When the device touches the stopper, further rotation f drive-screo t possibleno ws i that A . t position, actuatio microswitcf no h gives indicatiod nan stop motore sth . Damag drive-screo et motod wan r wil avoidee lb safety db y pins.

7. CONCLUSION

PFBR shutdown system incorporates adequate divers redundand ean t feature additionad san l passive features to ensure high reliability. Overall reliability of the shutdown system is based on

• proper design foreseeing all conditions and factors which would have effect on proper functioning, • well defined procedur fabricatior efo qualitn& y control, • testin sub-assemblief go scal1 1: e s& prototyp e system simulatin operatinl gal g conditions in reacto satisfdesigo e rt th l ynal requirementd san • design modifications based on the test results • regular surveillance tests.

workD R& s relate Shutdowdo t n system PFBf so undee Rar r progress.

95 REFERENCE

] Raja[1 , "DesignaL Babut e , a ,V. Philosoph shutdowf yo n syste PFBR"r mfo , This Technical Committee Meeting.

96 DEVELOPMEN PASSIVF TO E SAFETY DEVICE SODIUMR SFO - COOLED FAST REACTORS

Yu.E. BAGDASAROV, Yu.K. BUKSHA, R.M. VOZNESENSKI, N.V. VYUNMKOV, V.P. KÖRNILOV, N.B. KRAYEV, L.I. MAMAYEV, A.G. PORTYANOY, A.P. SOROKIN IPPE, Obninsk, Russian Federation

Abstract

In the recent years, development of passive safety devices for sodium-cooled fast reactors advanced significantly. The paper presents some research results of this subject, done at the IPPE (Institute of Physics and Power Engineering) during 1990-95.

INTRODUCTION

developmene Th f enhanced-safeto t mos e unitP th ts yi NP important problem of nuclear power development. Calculation estimates have shown a sodium-cooletha n i t d fast reactor considerable core damage mose th t ts a sever e accidents accompanie failura y db e of the main safety system can be avoided if a small effect upon reactivity by passive means is provided. The reactor with passive safety features actually acquires in this case inherent safety properties as all its safety effects and features are based on natural processes proceedin core e passive th Th g.n i e safety device additionae sar l meanmaie th no s(t safet y designesysteme ar d an )contro do t l beyond design basis accident failura maie t th sa nf e o safet y syste precludo t m ai wit ee hth sodiu m boiling and severe damages in the core. Various principle f passivo s e safety engineering realizatio e knownnar n I . Russia at present the most emphasis is being given to two principles of passive safety means actuation /1 - 3/: decreasey b - d primary coolant flow rate; increasey b - d temperatur coolanf e o cor e th et a toutle t In both case int cord insertioe e oro sth th e e taketh f no s plac gravityy eb .

1. PASSIVE SAFETY SUBASSEMBLIES (PSS) WIT HA HYDRAULICALL Y SUSPENDED ABSORBER ROD.

In 1988-89, two passive safety subassemblies based on the BR-10 standard subassemblies were developed, fabricated and put for tasting into the BR-10 reactor (Fig. 1.1 and Table! 1). The subassembly mockups were preliminarily tested at the hydraulic (water) rig. The calculation method developed has allowed to use the results obtained in water tests for sodium as well. In December 1994 the program of life time testing of the two subassemblies in reactoe th r including on-power actuatio completeds nwa teste Th .s have confirmed the design parameters of the subassemblies are recommended for use in the BR-10 reactor as standard ones.

97 core top

core niddle

core bottom

FIG. 1.1. PSS subassembly with hydraulically suspended rod for the BB-10 reactor.

98 TABLE 1.1. TECHNICAL PARAMETER PASSIVF SO E SAFETY SUBASSEMBLIES (PSS)

N Parameter Units N1 N2

1 Absorber rod cladding diameter mm 22.5*0.3 21.5*0.3 2 Absorbe weighd rro t g -242.0 -225.0 3 Absorbe efficiencd rro y %Ak/k 0.146 0.22 4 Flow rate through the nrrVh 96 81 reactor at a rod rise ) QN % ( (48) (40.5) 5 Flow rate through the mj/h 70 63 reactor at a rod drop (% QN) (35) (31.5) 6 Rated flow rate through the PSS at a raised rod m3/h 0.93 0.97 7 drod Timro p f eo int e oth core s 1.14 0.67

effece Th f absorbeo t r rods insertio systen no m reactivit f coolanyo t flow rate through the core was determined at power levels of 1-2000 kW. In this case coolant flow decreaserats ewa valu a rate e do lest o th e n f d so onetha% n.25 Mai BR-1ne th dat lifn n i 0a eo 2 reactotim1, epresenteS e testar rPS f so n di Table1.2.

TABLE 1.2. MAIN DATA OF PSS Nl, 2 TESTS IN THE BR-10 REACTOR

N Core Time On-power Accumulated Number of S oPS f cell duration operation, fluence actuations eff. days (E>0,Mev) (including n/cm2 thos- eon power) 110 3.01.89-5.01.89 0 PSSN1 110 29.03.89-1.08.89 39.55 2.6*1 2 10 38(10) 95 14.09.92-17.1192 21.96 95 15.05.91-11.08.92 151.07 PSS N 2 95 23.11.93-25.11.94 127.42 1.70*1022 125(10) 95 25.11.94-15.06.95 55.5

Tota- 2 1-21l N tim N S f operatio8eS o PS days PS r reactoe r fo th fo ,s n i rwa 1020 days. The number of rod drops during the time of the tests was 38 and 116 for RSS 2 subassemblies N d an N1 , respectively d jammincaseo e ro N .th f o sn i g subassemblies were noted. Flow rate values throug reacto e drorisd d hth ero an p t ra during the tests did not change. In 1989 work on the development of passive safety devices with the hydraulically suspendeBN-60e th r fo 0 d typdro e reacto starteds rwa 1988-8n I . 9a test absorber subassembly for the BN-600 reactor was designed on the basis of its standard shutdown absorber subassembly, and its full-scale mock-up was

99 manufacture testinr dhydraulie fo th gn i c (water Fig.1.2g ri ) . Several design versions of the subassembly were tested. At the end of 1994 rig tests of the subassemblies were completed; depending on their results, a subassembly design version for hydraulic testing in the BN-600 reactor will be recommended.

core tor)

- core middle

core botton

On—stop rod position. Lower working position of The upper working the rod after its movement e positioth f o n s it t a e drivr bth o y ( e hydraulically sus- floating-up from the stop). pended rod.

FIG. 1.2.subassemblyPSS with hydraulically BB-600 suspendedthe for reactor.rod

100 subassemble th r Fo y develope followine dth g algorith operatios it f mo s nha been adopted. At the shut-down reactor the rod in the subassembly is in the ultimate lower position - on the rigid stop of the guide sleeve. Prior to a reactor power rise the rod is engaged with the drive stem by means of its gripper and then it is lifted intuppee oth r working position primare Th . y coolant flow rat increases ei d froma minimum (0.25 Gnom) to a start-up (0.75 Gnom) one and the gripper is opened. In so doing the rod remains in the gripper as the rated flow rate of coolant through the sleev chosees i flothaa y suct n wi a twa ratha eGno6 equa 0. hydraulie m~ th o t l c force actin woullesgo d n uporo s e de thab f nth o weights nit e on t cuttin f A . o f gof three heat-removal reactor loop s poweit sd flo an wr rate automaticallar e y decreased to a level of 0.67 of the rated one. And the rod remains in the upper working position. At a signal for shut-down of the reactor the rod is automatically put into the lower working position by the drive stem at the open gripper. In this case the absorber remains in the gripper, as during the time of its movement (~1 s) from the upper working primare positio th lowee e th yon o r nt coolan t flow rate doet no s practically change. With the reduction of flow rate when changing over the primary pumps to reduced rotations (0.25 Gnom) a decrease of the hydrodynamical force takes placedecreass it t a ; dropd eweighd ro belo ro s e e th fro tw th e lowe mth r working abovpositiom stopm e et furtheth 0 A n)(8 . hols intit i brake ron d th i ean flow rate reductio fro d brake mro th e enth softly descend stope th cas n I a .n sf o eo drive start d failurro se eth fallin g intcore gravitoy th e b floa t wa y rate decrease down to <0.6 Gnom. At its fall it stops at first in the brake (40 mm above the stop) thend furthet an a , r flow rate reduction t softli , y seats agains stope th t . At reactor refuellin 0.2= G g5 ( stop) Gnom.e beind th ro cas n n n i g,e a o f th ,e o erroneous switchinpumpe th f fulo o s t n lg o capacit doed t floaro Jnt sp no e u to yth becaus) core th t onle(i mm buoyane y0 brake4 e th rise ~ th eo n si t t force actinn go it in this case is more than 2 times less than its weight. Non-floating up of the rod is provided due to a hydraulic load relief of the working part of the rod. In Tabl 3 mai1. e n hydrodynamical characteristic e BN-60th f o s 0 standard safety system subassemblie S subassemblP o tw f o d ysan teste e watee ar th g dn i ri r presented; coolant flow rates (l/s) are given scaled to sodium at an operating temperature.

TABLE 1.3. HYDRAULIC CHARACTERISTICS OF THE BR-600 STANDARD SUBASSEMBLS S EXPERIMENTAD YAN SUBASSEMBLIES LP S

S/A Qnom Qmin Type Qsusp Qnom Qfloat Qb il itis T2,S rod rod safety 2.0 1.0 0.25 2.2 S/A PSSN1 10.1 6.1 3.6 6.0 11.5 2.1 1.0 0.36 2.5 PSSN2 8.7 4.7 2.7 4.5 11.5 2.1 1.0 0.25 3.6

101

e m insertiod ro wher f o ' n t'et i - intcortota e oth e afte beginnine rth accidentn a f go ; insertiod T- tim2ro f eo n intcore oth e fro momene mth f beginnino t s it f go movement (primary flow rate reductio Gnom.)6 0. no t ; QSUS pcoolan- t flow rate through subassembly under whic suspensiod hro n in the upper working position by coolant flow takes place; Qnom - rated coolant flow rate through S/A; Qfloat " coolant flow rate through S/A at which the rod would float up from lowee th r position intuppee oth r one; Qb - coolant flow rate through S/A at which the rod would float up in the brake; Qno coolan_ m t flo we uppe th rat n i er workind througro e ghth r°d position at a rated flow rate through S/A; r Qrnin _CO olant flow rate through the rod in the lower position at °d a flow rate throug f 0.2o 5A hG/nom.S/ ; r| - margin for rod non-floating up.

In Figs effec.e 1.3th f PSS4 o t, 1. s actuation with variou worthd sro s (0.d 6an 1.2% Ak/k) is shown at a beyond design basis accident in the BN-600 reactor with total loss of electric power supply and a failure of all active reactivity control systems r varioufo s beginnine time th actuatioS f o sPS f d 14sgo nan (T<4 ) fro= | e mth beginning of an accident and of absorber insertion into the core (i2-2, 4, 7 s). It follows fro figure e mworthd th ro e s ,th tha valuer tfo s under consideration2 1 d , T-an J

900

800

700

600

500 O a 0 Time4 >s 20 FIG. 1.3. Sodium temperature corethe at outlet fonctiona as timeof since the onset of an accident.

102 800

700

£ 11 H

600

500 200 250 300 Time, s FIG. 1.4. Sodium temperature corethe at outlet functiona as timeof since the onset of an accident.

core th e outlet sodium temperature earln levea t ya l stag f accideneo t development is mainly determined by the beginning of absorber insertion (Fig. 1.3), and at a later reactivitvalue e th stagth f ey o b e- y worth inserted (Fig. 1.4). Fro mt followi Fig3 1. . s efficiencS PS thaa t ya t valu f HD.6eo % Ak/k (efficienc standare on f yo d safety rod), tima f startineo actuatios git n t-|=4s (during this time fro beginnine mth e th f go accident the coolant flow rate decreases to a critical value of 0.6 Gnom.) and a time f absorbeo r Insertion intcore oth e T2=§s (see Table core 1.3th e) outlet sodium temperature will not exceed 720°C, i.e., considerable margin (200°C) is ensured to the sodium boiling point.

SUBASSEMBLS P . 2 Y WITH THERMAL-PRINCIPLE-BASED ACTUATING DEVICES (AD).

The test PS subassernbly of this type is being developed on the base of the BN-600 reactor standard subassernbly subassernble th n I . shorteya r fuel element bundle is used, and coolant flow rate is decreased, respectively, so that the coolant temperatur fuee th l t fueeelementd a an l subassernblS sPS outlee th n i t y woule db close to each other. In the head piece of the PSS an actuating device holding the absorber rod is located. At an increase of the AD temperature it comes into action - releases the absorber rod which drops into the core by gravity. Such type AD as applie operatins it do t g condition BN-60 n sundei s i S r0 PS developmen t since 1990.

103 From Fig. 1.3 it follows that if the temperature for AD to come into action is taken as equal to 650-670°C then to obtain a margin of 100 and 150°C to the sodium boiling poin necessars i t i t y tha time tf absorbeth eo r insertion beginning (T-J) into the core (AD lag time) should not exceed 10 and 5 seconds, respectively.

2.1. Magnetic Material-Based Actuating Device

Fig.2.1 shows a mock-up design of the magnetic actuating device (MAD) develope r experimentadfo l testin sodiue s applieth a operatios it n g i m ri o d t n condition BN-60e th n si 0 PSS.

MAD type H, H, *•

dimensions are presented in mm Drivd ero

Armature

FIG. 2.1. Magnetic actuating device (MAD).

104 The magnetic system of the MAD consists of a permanent magnet made of magnet-solid alloy magnetized in the axial sense, a screen made of ferrous-nickel alloy with a Curie point of 620°C and an armature of Armco iron connected to the absorber rodscreea t .A n temperature increase abov Curie eth e poin t i loset s it s magnetic properties tha tloadD resultMA - liftinn i s g capacity reductiona t A . decrease of the load-lifting capacity below the absorber rod weight, the MAD armature disengagedropd ro se th intcor e d gravityoy th presene sb an th y B . t time testing of the MAD mock-up in the sodium rig has been conducted (~1000 hours). Its load-lifting capacity in a sodium flow within a temperature range of 300 - 680°C has been determined, and its dynamical characteristics at a fast (~12°C per second) rise of passing-over sodium temperature were studied as well. The MAD mock-up tests have revealed that it has considerable thermal inertia: s timit e . Cominconstans 4 6. gtakes i D tint osMA placactioe th f eno durin6 g~ seconds afte beginnine th r accidente th f g o coolan e th , t temperatur core th et ea outle exceedint no t g 715°C. At present preparatory work for the advanced MAD design tests is under way.

t Actuatio2.2A . n Device Functionin Basie th f Varioun so go s Physical Effects.

This device (Fig.2.2) include numbesa f thermosensitivo r e working elements functionin basie th f variousn o g o s physical effects (phase transition, memory effect PSS subassembly head

Device case Disk spring with memory effec shapf to e

Silphon

Disk spring with memory effect of shape

Gripper

Absorber FIG. 2. 2. Operating device functioning on the basis of various physical effects.

105 of shape, etc.). An absorber drop takes place at actuation of any of thermosensitive elements e maiTh .n working elemen e devicth f s bellowo i et s passed ovey b r coolant, plugge botn do h endcompresse e filled th san dn i d condition wit hmateriaa l having a phase transition with a volume increase at an actuation temperature

(aluminium with tmejt ~660° d AV~6.6%)Can . Calculatio experimentad nan l studies have shown that bellows elements havea characteristic, 8s o t 2 f o tim a ( sg ela stroke of 2 to 8 mm, a force of 450 to 10^N) which can ensure the required for device performance. The device design includes also two (upper and lower) elements, made of titanium alloy with the memory effect of shape at a temperature of 630-670°C, manufacture a disn i ford f cmo spring pack. Calculatiod an n experimental validation studies of this element were carried out. It has been shown that these elements have characteristics ensuring reliable actuatio device th f ns o e a well (th force th actuatio, e developemm strok8 a n, 6- f tim1s eo f 700H) df o e o . Calculation and experimental work on obtaining the required characteristics of the devic undees i r way.

CONCLUSION.

At IPPE work on the development of PSS subassemblies based on two principle theif so r actuatio underwayns i : - by decreased primary coolant flow rate; - by increased temperature of coolant at the core outlet. A PSS subassembly with a hydraulically suspended rod for the BR-10 reactor has been designed. Two such type subassemblies have successfully undergone life tim ee BR-1 testth n i s0 reactor including on-power actuation e testTh .s have confirmed the design characteristics of the subassemblies; such type subassemblies are recommended for routine operation. A full-scale PSS subassembly with the hydraulically suspended rod for the BN-600 reactor has been designed and tested in the hydraulic (water) rig. The test results obtained allow to recommend such a subassembly for in-pile tests. A magnetic device actuating at an excess of a designed temperature has been develope tested sodiue dth an n di m rigdevice .Th e desig bees nha n improven do the basis of the test results, and its rig tests are carried on. A device actuated by increased temperature on the base of various physical effects (memory effec f shapeo t , phase transition, compressed spring energy, etc.) has been developed insertiod Ro . n intcore oth e takes plac t actuatiof ea o y an f no thermosensitive elements. Experimenta d calculatioan l n studiee deviceth f o s , thermosensitive elements characteristic carriee sar d out.

REFERENCES

1. I.A. Kuznetsov et al. "Objectives and Main Activities on Fast Reactor Safety since April 1986". Int. Fast Reactor Safety Meeting, Snowbird, 1990, v.1 11-18. pp , . 2. Yu.K. Buksha "The Status of Work in the USSR on Using Inherent Self-Protection Feature f Liquiso d Metal Fast Reactor". Specialists' Meetin Passivn go d Activean e Safety Features of LMFRs, 5-7 Nov., 1991, Japan, IWGFR/85, pp.64-70. 3. Yu.E. Bagdasarov "The Development of Passive Safety Devices for Sodium- Cooled Fast Reactors". Sodium-Cooled Fast Reactors Sofety International Tropical Meeting Obninsk, Russia, October 3-7, 1994, v.4, pp.6.51-6.58.

106 MAIN FEATURES OF THE BN-800 PASSIVE SHUTDOWN RODS

Yu.K. ALEXANDROV, V.A. ROGOV, A.S. SHABALIN OKBM, Nizhny Novgorod, Russian Federation

Abstract

The major design characteristics of hydraulically suspended shutdown rods, as well as the principles of their operation and test results for the mock-up scram rods, in combinatio reacto0 n80 wit - simulation e r hN internath B e th lf so structure s providinr gfo the rods operation givene ar , .

1. INTRODUCTION

e analysiTh f beyono s d design basis accident r BN-80fo s 0 reactor showed that preveno t t overheatin core th casn ei f f gloso e o f powe so r primaro r shutdowP yMC n with main shutdown systems failurcommoe th o et n reason, applicatio auxiliarn a f no y system is required, which is based on the alternative, preferably passive principles of actuation. Taking into account characteric features of core structures and BN-800 above- core structures (separate transfer of guide sleeves and rods, the so-called closed FA head, defici f spaco t e abov sleevea core n r installatioeth i efo d , ro deteriorate a f no d thermal bond betwee d channero n d sodiue cor an th la resul e t m s a a outlet t d an ) of possible variants development an additional shutdown system was accepted comprising three HSRs, spontaneously inserted inte coroth f primarei y sodium flow rat s decreasedi e .

2. MAIN PRINCIPLES OF HSR FUNCTIONING

A possibility to implement an additional shutdown system based on hydraulically suspended HSR for BN-800 reactor was stipulated by sufficiently high initial level of primary sodium flowrate during usual ("cold") reactor start-up and disconnection of a loop (—0.7 Gnom), what allows to accept the required, in respect to core cooling conditions, sodium flow (-0.5 Gnom) at which HSRs are suspended in the flow and began to drop into the core at MCP-I coastdown. The cocking of HSRs is done by their lifting into upper position before bringing the reactor to power as it is done for rods of main emergency protection system with subsequent opening of grips of actuators at primary sodium flow increase up to >0.7 Gn . The suspension of HSRs in their upper position is assured by the choice of sodium flow betweep througga e e guidhth th n e sleeve channel abov e cord th e an e effective link of rod and the required (to conditions of nuclear safety) holding of corHSRe th e n saftei r actuatio durinr no g refuellin assures gi considerably db e decrease of rods hydraulic resistanc thein ei r lower positio bypaso t e ndu s holes e madth n i e lower part of extension rod communicating the increased gap between the effective lind sleevan k e with inner e extensiocavitth f o yd (Fig.l)ro n .

107 CucnncHHe PnciicrmeHHe PEIÖOT& Ha Cpaoa'i'MHaiivK' N--0 DAS n PCJKHMO rieperpysKa MOIUHOCTH ncmHoro oOoc- - G-0.2GHOM. G-0,7GiioM. N-NuoM TO^IlIBcXIIHSC A I yi npi! BA3 C--GHOM. G<0,5CiHOM. Refuelling DecouplingR HS Power Coupling operation actuation

Hanponngjomng Tpy6 CVM U n3

a el rnoraaio n .0c 1 g 0>

o T •H o 0

Fig.l HSR functioning scheme

108 To mitigate the impact of HSR rod onto the sleeve at actuation a hydromechanical shock absorbe s providei r d betwee head nro d having conical surfac r seatinfo e g onto e uppeth e sleevr th d facuppe f an o ee r extension rod.

3. DESIGN CHARACTERISTICS OF HSR

A para s f BN-80o t 0 reactor core project with zero sodium void reactivity effect developmenR HS e th calculationad an t l verification were carried out. When developing e emergencth R HS y protectio d designro r assurin s takefo i na basi t s e a nbu s th g minimal flow of sodium through the sleeve the mass of HSR rod is reduced to minimally e diameteTh . f sleevkg o r 7 e -1 d wrappepossiblchanne s ro i d d an an erl tubes i s , respectivelymm 3 7 P rodd E . an sr takefo m namel m s a n8 7 y e calculationTh s performed have shown thae velocitth t f sodiuo y n mrequirei d the gap for HSR rod suspension is ~4 m/s what corresponds to nominal flow of sodium throug gapR ,h HS i.e.~5 kg/s, thif therebo sg flok 1 wy~ goes through effective e rodth lin . f ko Bypassin effective gth betweep ega lind e extensiod nth kan an d nro channel through bypass holes assures double margin to HSR floating up in respect to pressure drop at nominal flow of sodium through the core. The results of calculation analysis of HSR movement dynamics at MCP-I coastdown are shown in Fig.2. Here the results of calculations of sodium temperature variations at core outlet at MCP-I deenergization with failure of main shutdown systems are also given. As the calculations show the time of HSR drop is ~6 s (~13s from begi f MCP-o n I coastdown) what limit e increasth s f temperaturo e t cora e e outlet to 740'C.

. EXPERIMENTA4 L INVESTIGATION MODEF O S L HSR HYDROMECHANICAL CHARACTERISTICS

r developmenFo f BN-80o t 0characteristicd ro reacto R HS r e investigationth s s of mode characteristicR HS l s were carrie togethet dou r with reactor model units assuring HSR functioning (sleeve, guide tube, actuator grip) and using water as working medium having clos sodiuo et m hydrodynamical characteristics. Chec f mode ko movemenR HS l t s effectewa d wite hel th hf sensore o pbasi th f magne o sn o s t controlled contacts. e testTh s performed allowe determino dt e hydraulieth c characteristic f modeo s l HSR and establish the laws of its movement at lowering the water flow modelling e coastdowth f BN-80no 0 MCP-I pum determind pan e conditioneth f modeo sR HS l floating up from its lower position. The basic parameters of model rod are close to expected ones. In typicae Figth .3 l experimental plot f mode so movemenR HS l t lowerina t e gth water through the sleeve are given. In total, 60 drops of model HSR was done including drops at maximum adverse radial displacements of the sleeve and guide tubes exceeding possible displacements in BN-800 reactor. The inspection of model HSR and the sleeve after tests did no reveal damages marked an d deformatio f contaco n t surfaces. Slight glos sealinn o s d ro g e coneth f o s and sleeve head was detected as well as circular attrition on shroud tube of rod effective link, mainly, in the zone of contact with lower edge of guide sleeve.

109 A/ G 1,0

0,5

Fig. R movemen2HS d reactoan t r parameters variatio t deenergizatioa n n with failurmail al n f shutdowo e n systems H-relative movemen HSRf to , N-relative power, G-relative flowrate, t-sodium temperatur t cora e e outlet Fig.3 Movement of model HSR at flow decrease H--movemen modef o t ; AP-pressurlm rod, e drop thoug e channelth h , kgf/cm CONCLUSION

An additional shutdown system is developed for BN-800 reactor on the basis of hydraulically suspended absorber rods arbitrarily dropping int core o th t lowerin ea g e flo f th sodiuwo m throug systeR e corHS hth m e e beloassure Th . 5 Gw0. s safe

shutdow reactoe th f no t MCP- ra I deenergizatio independenf nom of t driveS cu r no CP f so t r controo l equipment condition. Rig tests of model HSP performed using water as working medium in conditions maximally approaching actual ones demonstrated high reliabilit f hydraulicallo y y suspende d functioninro d g what allow o begit s n testin e experimentath g d ro S MR l in BN-600 reactor.

112 DESIGE BACKUTH A F NO P REACTOR SHUTDOWN SYSTEM OF DFBR

. OKADK A Mitsubishi Heavy Industies, LTD., Yokohama

K. TARUTANI Toshiba Corporation, Tokyo

Y. SHIBATA Hitachi ltd., Tokyo

M. UETA, T. INAGAKI Japan Atomic Power Company, Tokyo

Japan

Abstract

This paper outlines the status of the development of passive safety measures to achieve subcriticality in DFBR. Two independent reactor shutdown systems (RSSs), a a backu d primaran p S RSS e RS yprovide ar , d eac f whico h n rapidlca h y brine th g reacto a saf o et r shutdown condition s i e providebacku Th S RS .p d wita h Self—Actuated Shutdown System (SASS), which is passively released due to the Curie point effect by abnormal rising of coolant temperature. Two independent RSSs reduce occurrence th e probabilit corf yo ATWe e th damag o t S e (ULOFedu , UTO ULOHSP& ) events to less than 10 "'/ry."1 Installation of the SASS reduces the occurrence probability of ULOF & ULOHS further by two orders of magnitude. Overall

frequenc corf yo e damag ATWo t e eSdu event reduces si ordee th lesf o dro t s than 10 ~ fr9 y owing to the SASS installation. A screening study was conducted to search for promising core safety enhancement technologies based on a consideration of reactivity requirement passivr sfo e success scenario ATWn si S eventsresula s f A o t . this study, an installation of Gas Expansion Modules (GEMs) in DFBR was recently proposed to enhance the core safety further. The reactivity worth of 66 GEMs arranged outside of the core was approximately -1.5$, and it is possible to prevent the sodium boiling on ULOF by GEMs alone. In Addition to SASS and GEM, the feasibility study of the Enhanced Thermal Elongation Mechanisms (ETEM) has been conducted. As the results of some experiments and design studies, it was confirmed that ETEM could be one of the promising mechanisms to enhance core safety, and gives less impacperformance th n o t futura f eo e large scale reactor core.

1. Status of DFBR course Inth developinf eo g Japanese FBRs experimentae th , "JoyoR FB lbeins "i g operated satisfactoril e prototypth d "MonjuR an y FB es achieve ha " d criticalitf o s a y April, 1994. This leads to the prospect of the construction of the demonstration FBR

ry (reactor years)

113 (DFBR) and the commercialization of FBRs at around the year 2030. e desig Th DFBe nth r studbeinR s i relatefo d yD an g conductedR& Japae th ny db Atomic Powe . QAPC)Co r , e principawhicth s i h l organization responsible th r fo e construction as entrusted by nine electric power companies and Electric Power Developmen . Ltd.Co t . Researc developmend han DFBe th n Rbeins o i t g conductey db the Power Reactor and Nuclear Fuel Development Corporation (PNC), the Japan Atomic Energy Research Institute (JAERI) e Centrath , l Research Institut e Electrith f o e c Power Industry (CRIEPI d JAPC o an coordinatT ) . e R&D, these organizations establishe Steerine dth g Committe Coordinatinr efo Japanesn o D geR& FBRs. In designin e DFBR gth s importan i t e commercializatioi ,th r fo t f FBRo n o t s reduc e planth e t construction cost, whic s rathei h r greater than tha f ligho t t water reactors (LWRs). From this viewpoint, two types of FBRs were compared. These pooe th l e typar e e top-entrreactorth d an , y loop-type reactor ,mose whicth s t hha compact arrangemen primare th f o t y system components tope Th - entr. y loop- type was selected for the DFBR because of the following considerations: a) Major primary component sintermediate sucth s ha e heat exchangers (IHXsd an ) the pumps are outside of the reactor vessel. This facilitates maintenance and repair. systee b )Th flexibilit s mha introduco yt e such innovative technologies necessaro yt achiev commercializatioe eth FBRse th f no . Experienc) c e gaine prototype th t da e "Monju" mus fulle b t y utilized. Considering that the top-entry system is quite a new concept, the conceptual design study e evaluatioth , n stud f commercializatioyo n prospect wated an s r hydraulic tests using model f thermal—hydraulio s c properties peculia e top-entrth o t r y system were conducted. Through these studies, technicalR feasibilite FB possibilitth e d th an f yo y commercialization was confirmed, and the plant concept was also established. Also based on the results of these studies, the presidential committee of the Federatio f Electrino c Power Companies determine e basith d c specificatione th r fo s No.l DFB JanuarRn i y 1994 decided ,an promoto dt s developmeneit t aimin starr gfo f o t construction in the early 2000s. desige Th n target DFBe followss a th f e sRo ar : a) Safety level shoul t leasa gooe s dta b thas d a LWRf o t s b) Economic feasibilitl,OOOMWa n o R t abouya LW etime5 basie 1. tcos e th sth f s o t c) High burnup and long operating cycle to reduce cost of electricity generated d) Reactor outlet achievtemperaturo t C ° e0 hig55 f heo thermal efficiency ) Easiee r maintenanc repaird ean , taking advantag distributee th f eo d equipment layout Table-1 list majoe sth r specification plante th f .o s Fig-1 show appearance sth e of the top- entry loop- type DFBR.

. 2 Current desig reactof no r shutdown system DFBe th f sRo Two independent reactor shutdown system e providedar s a primar, y reactor shutdown system and a backup reactor shutdown system, to increase system reliability and to eliminate common mode failures. (Fig-2) Each system is completely

114 TABLE 1. PLANT REFERENCE SPECIFICATIONS

Item Spécifications Basic spécifications 1 Reactor type Top entry loop type reactor 2 Thermal output l,600MWt (Electric output) (about 660MWe)

Other major specifications 3 Numbe loopf ro s 3 loops (about 530MWloop) 4 Reactor outlet temperature 550 °C 5 Main steam temperaturd ean 495 °C 7169 atg pressure 6 Core and fuel (1) Core Homogeneous core (2) Fuel Pu-U mixed oxide fuel pellet with center hole (3) Burnup About 90,OOOMWd/t (Initial phase) About 150,OOOMWd/t(ffigh burnup phase) (4) Breeding ratio With radial blanke Abou: 2 t 1. t Without radial blanke Abou: t t 1.05 7 Steam generator Once—through type 8 Reactor shutdown system 2 independent systems 9 Decay heat removal system loop4 DRAGf so S 10 Reactor containment Rectangular reinforced concrete containment with liner 1 Fue1 l handling system Rotating plug, manipulator type 12 Reactor_ buildin_ g Horizontal seismic isolation building___

Reactor Intermediate vessel heat exchanger

(Primary system) (Secondary system) (Turbine system) Fig-1 Plant configuration

115 Primary reactor Backup reactor shutdown system shutdown system Control rod drive Control rod drive median i sm mechanism /

Se If-actuated shutdown system Rigid type Articulated control rod Controd ro l

Core

Fig-2 Arrangement of the reactor shutdown systems

independent from senso controo rt l rods. Parameter safete th n ysi protection system are selected to fully protect the core from all abnormal occurrences. For the detectio f abnormao n l events occurring relatively frequently which require prompt shutdown e primarth , y reactor shutdown e backusysteth d mpan reactor shutdown system have different parameters. Each shutdown system logi4 f o c t s ou work 2 y sb with diversity. e primarTh y reactor shutdown system employs rigid controe l th rod f o s gas-accelerated insertion type which were developed for the prototype reactor "Monju". e backuTh p reactor shutdown system employs articulated controe l th rod f o s gravity- dropping type. The backup reactor shutdown system is also equipped with a Self-Actuated Shutdown System (SASS), whic releasn hca e control rods after sensinga temperature rise of the coolant. This ensures the control rod insertion even if the safety protection systems fail.

3 . Passive safety measures The reliability of two independent RSSs is high enough to address Anticipated Transients Without Scrams (ATWS) as the beyond design basis events. However, if one assumes ATWS events, the reactor core would be damaged in several tens of seconds which is too short a period to allow the operator to take any action in order to preven e corth t e damage. Henc passive eth e shutdown capability, especially against ATWS events, is desired to enlarge the grace period of a large scale FBR toward its commercialization.

116 Ther e severaear l idea establiso st h passive shutdown capability t shouli t Bu d. noticee b d that suc hpassiva e countermeasure shoul selectee db takiny db g accounf o t safete th y characteristic LMFBRf so mose Th t .crucia ldesige issuth r thas nefo i e th t passive countermeasure should have less uncertaint s it evaluatio r fo d y an n developments, without affecting the core and plant design. As a result the SASS has been selected as the most promising countermeasure to introduce int DFBRe oth . The PSA study of the DFBR concluded that two independent RSSs reduce the occurrence probabilit three th f eyo type ATWf so S (ULOF, UTO ULOHSP& ) evento st less than 10 ~7/ry. Installation of the SASS reduces the occurrence probabilities of ULOF & ULOHS further by two orders of magnitude, reducing the probabilities of the three ATWS event same shows th a ) en si Fig-3 orden i Ay ° UTOf 1 ro ~ .0 P(1 A screening studs conductewa y o t searcd r promisinfo h g core safety enhancement technologies for the DFBR based on a consideration of reactivity requirement r passivfo s e success scenario ATWn i s S a resulevents s f thiA o t s. study, the gas expansion module (GEM) has been selected for a detail design study for consideres i applicatioM e th GE DFBRe shoo e dt th Th wn ni .promis measura s ea e to enhance the core safety by slowing down the event progress during the ULOF and prevent from the coolant boiling. In addition to SASS and GEM, the enhanced thermal elongation mechanism (ETEMe controth r d driv fo ro unde)lw e no lin r s i edevelopmen o improvt t e th e inherent safety characteristic larga n si e targee coreextenscalo R t Th s i e tFB . dth grace period up to a half hour against ATWS events.

Result of PSA Levé 1-1 Study

Without SASS With SASS . 2E-8 . 8a means 8.2X10 -8 a 5.7E-8 8. 2E-8 2.5E-8

5.7E-10 9.4E-10

ULOF ULOHS UTOP Overal Type of ATWS Event Fig-3 Reductio occurrencf no e probabilit ATWf yo S events installatioy b SASe th f DFBSn o i R

117 . Feature4 s of design measures 4 . 1 Self- Actuating Shutdown System(SASS) The SASS of the DFBR is installed in the backup shutdown system, which e two-segmen rodth C f o sB4 consist x si f o s t articulated type (Fig-2) e SASTh S. concep SASshowe s i t Th SFign ni consist . -4 conventionaa f so l electro- magned an t a Curie-point temperature magneti c a structur e latte alloys Th ha r . e with fino t s enhance its quick response to a change in its ambient temperature. The SASS has a flow guide which lead e sodiuth s x fuemsi l e frosubassembliee th exitmth f o s s neighboring the SASS rod to the temperature sensitive alloy (TSA). The flow guide has optimized flow holes to shorten the transport delay time of the sodium flow from the exits to the TSA.

Drive motor

UpperP R surfac e th f eo Magnet core Na level CRDM • — Upper guide tube Actuation for CRDM Coil detector \ ^ —— SASS Top of the ! Actuation Self-welding core region ~~ detector prevention CR- fitting CR guide tube Temperature sensitive alloy _„_j

Fig-4 SASS concept

Desig) holdine (1 th f no g forc . temperaturevs SASf eo S magnet Fig-5 show e desige holdinth sth f o n g force versu e temperaturth s f SASo e S magnet. The SASS rods will be released when the magnetic holding force decreases less than the effective weight of the control rod in the sodium flow. The SASS must securely hold the rod under normal operating conditions to exclude any spurious reactor scram. Hence the holding force is designed to be twice e lowermosth t a d ro t e weighe windoth th f o tw temperatur /whice C ° th 0 s i h64 f o e expected maximum temperatur durinA normae TS gth e th f le o operatio r nc unde3 e rth uncertainty condition. uppermose Ath t t windo wholdine temperaturth , C g° forc0 68 designes ei f eo o dt assuro t d release ro eSAShalth e a weighe e th e th f f b th Se o f rodo t s whe sodiue nth m temperatur f coro e e outlet increases abnormally (without scram)e uppermosTh . t window temperature is determined from the requirement for prevention of sodium

118 Temperature range duringUncertaint, f o y reactor operation temperature:3 a\ '% Maximum m temperature 1500 m. expected for ...... rmmmmmmmm r"' safe shutdown Holding forc t whicea h JJj; -/// the control rod wiII %% Upholding force |p at which the definitel e helb y d CR effective ^control rod j»aiabi._._j itely je released Minimum temperaturo t e w////////jy/////A prevent malfunctions

0 68 0 64 600 Temperature(°C) Fig-5 Relation between holding force and temperature

boiling inception (wit appropriatn ha e margin e nominath n t i subassembl) ho l e th f yo core in a typical ULOF event of DFBR (i.e., loss of off-site power without scram during rated power operation).

(2) The selection criteria used for the temperature sensitive alloy are as follows: a) the Curie point temperature should be between 650 °C and 680 °C , saturatee bth ) d flux density shoul more db e than preventive 0.5th t Ta e temperature of contro spurioud ro l s drop (64, ) 0C ° c) the ratio of saturated magnetic flux densities at 680 °C and 640 °C should be more than two, and d) magnetic hysteresis and resistance should be negligibly small. Three candidate Ni-Co-Fe alloys were selecte alloe th yr dfo whic h will satisfy e magnetith c holding force characteristi f Fig—5e chemicaco Th . l compositiond an s magnetic characteristics of the selected alloys are shown in Fig-6. They were confirme havo dt negligiblea e magnetic hystersi resistanced san . Laboratory testf so thermal aging, metallography and strength of the materials have been conducted. The thermal aging test in high temperature air has been continuing for more than 3740 hours until 24000 hours.

4 . 2 Gas expansion module (GEM) Application of GEMs in fest reactors was investigated in the USA over many years and employed in the design of the PRISM plant, which was reviewed by the USNR procese th Cn i Generif so c Approva Nucleaf o l r Plante Designparth a f s o tA . PRISM design activity effectivenese th , againsM GE tf o mains - pump stop accidentn si a small fast reacto demonstrates rwa USDOy db passive th n Ei e safety tes FFTn i t Fn i 1986.

119 20Ni-42Co 23Ni-38Co

300 400 500 600 700 800 TEMPERATURE(°C) Fig-6 Magnetic characteristics test for SASS (Saturation Magnetic Flux Density Curves)

GE facilita Ms i s containe i y s e uppewherga th e n di reth portio duca d f no an t the lower end is open to the high pressure plenum. At the time of an accident, the neutron leakage is increased by the expansion of the gas with decreases of the pump discharge pressure and negative th ' e reactivit insertes yi s showa d a s A Fign ni . -7 e supposedesige resular th f M o nt surrouno studydt GE e th ,e entir dth e peripherf yo e cor th s showea Fig-8n ni e reactivit Th . y wortGEM6 6 f ho s arranged outsidf eo the core was -1.5$, and it was possible to prevent the sodium boiling on ULOF by GEMs alone. GEM GEM

Neutron

Increased Neutron Leakage

Negative Reactivity

Pumpn O s Pumps Off

Fig- 7 Operating principle of GEM

120 Pin Diameter 8. 5mm Pin r Assemblspe y 217 Assembly Pitch 158. 1mm

O Core Fuel(Inner9 19 ) <§> Core Fuel (Outer) 96 Core Tota5 29 l ® Primary Control Rod 24 ® Backup Control Rod 6 Contro Totad 0 3 Ro l l s ExpansioG$Ga > n Modul6 6 e (GEM) O Radial Blanket 150 ShielS SU d2 Layer> f 4 17 s B4C Shield 3 Layers 306

Fig-8 Core layou DFBr fo t R

GEM reactivity disturbance due to the fluctuation of the inlet plenum pressure during normal operatio partiad nan l load conditions imposes another restrictioe th n no GEM s conclude designwa t I .d tha o limi t e treactivit th t y disturbance within a n allowable range, the sodium level in GEM must be raised high in the upper axial blanket region reactivitM , wherGE e slope yth e th wort f eo smals hi l enoug accepo ht t the sodium level fluctuation. To satisfy these conditions imposed from the plant operations, it has been proposed to keep the flow rate of the primary system at 100 % rated flow whenever the reactor is at criticality.

Enhance3 . 4 d Thermal Elongation Mechanism(ETEM) The ETEM is installed at the upper part of the control rod. The ETEM will insert the neutron absorber into the core following the temperature rise of the coolant, and then it gives a negative feedback effect, which improves the self—control capability. Two type f ETEo s M concep e showar t n Fig-9i n . Control rod n operatini s g condition e core hunETEMd th ar s ean n g i e expose e ar sflowin th y db g sodium from the neighboring Subassemblies, so that ETEMs will easily respond to the temperature changes in the outlet of Subassemblies. (1) Linear Hydraulic Expansion Device (LHED) e workinTh g principl LHEe converth o t f eDo s i e volumth t e expansio liquif no d to the linear elongation. Liquid Sodium or Sodium—Potassium Alloy (NaK) with large thermal expansion coefficients are likely to be used as the working fluid. (2) Amplificatio Differentiaf no l Expansion Device (ADED) The working principle of the ADED is that the multiple links work to convert and amplif e thermath y l expansion difference, betwee e temperature—sensitivnth e cylinder and the low—expansion rod to axial elongation. Austenitic stainless steel for the

121 Linear Hydraulic Expansion Device Amplificatio f Differentiao n l Expansion Device

Temperature Sensitive Chamber (Filling Liquid Metal)

Spring Temperature Sensitive p—'u Cylinder Amounf to

Control Rod

Fig- 9 Schematic diagrams for two types of ETEM

temperature— sensitive cylinde molybdenua d an r expansiow tungstelo a mr e o th r nnfo rod are employed as the thermal expansion members. A feature of the mechanism is tha larga t e elongation rate wil expectee b l e earlth yn di stage during ATWS events, sinc e temperature—sensitivth e e cylinder wil enlargheateo e t b l s a e t firs da eth o s t thermal expansion difference.

ETEM design requirements have been determine e s followsdbasie a th th f o sn o , ATWS analyses: Thermal elongation characteristics: 120mm/200°C (forBCRs) 30mm/200°C (forPCRs) BC R Backu: Primar: R pPC y contro contro d d ro l ro l Response time constant: Less than 10 sec.

. Evaluatio5 performancf no e of measures ( Performanc SASS/GEf eo M system against ULOF) Analysi1 . 5 s conditions The performance of the SASS/GEM system of the DFBR on a ULOF events was evaluated under the following conditions: a) Nominal hot channel conditions •Peak linear heat rating = 410 W/cm •Initial maximum coolant temperaturC ° 7 62 e=

122 b) SASS conditions •Temperature time constant of the SASS magnet = 1.0s •Coolant transport delay time = 1.5s •Initial coolant temperature around the magnet = 571 °C •SAS delatchind Sro g temperaturC ° 0 68 e= c) Plant conditions •Flow halving time of the primary pumps = 6.5s •Pony motor flow = 15 % of rated flow casn I e 2 o . f SAS5 S alone On the ULOF events, even if the reactor shut down systems fail, the SASS rod will be released when the SASS temperature sensitive alloy reaches 680 °C about 8 seconds after the pump trip. As shown in Fig-10, in case of the SASS alone (without GEMs), maximum coolant temperatur , whic°C 0 hs restrictei e 88 satisfie e b e o dt th s non- boiling conditio preventd nan core sth e damage.

(SASS Alone)

Coolant Boi Temg Iin p

20 30 Time(sec) (SASS-t-GEM) (For Reference:GEM Alone)

P P 900 g CoolanTemin I p i Bo t D. Coolant BoiI ing Temp Q. .» 800

O O O 600 Nominal Hot Pin ,5 Nominal Hot O Coolant Temp Coolant Temp 500 20 30 40 20 30 Time(sec) Time(sec)

Fig-10 Core temperature change on ULOF

5 . 3 In case of SASS & GEM Fig—11 shows the relationship between the sodium level in GEM and GEM reactivity worth. On the ULOF events, the rapid negative reactivity is inserted by GEMs with decreases of pump discharges after pump trip, and it makes the reactor power decrease e SAS d wilTh e releasero b Sl. d whe e SASth n S temperature sensitive alloy reaches 680 °C about 10 seconds after the pump trip. The maximum coolant temperature is 790 °C as shown in Fig-10, which is 90 °C less than in case of SASS alone.

123 135

\ (Upper Blanket Region) g 100

LU GEM Reactivity CP -50

Even in case of GEMs alone (for reference), maximum coolant temperature reaches 880 °C as shown in Fig—10, and it is possible to prevent the sodium boiling on ULOF.

6 . Future development plans for SASS and GEM (1) Future development plan for SASS Future plans for experimental studies consist of materials tests and tests for the components and whole system of SASS. (Materials tests) a) Thermal aging tests of the TSAs & iron core from JFY'93 to '96 b) High-temperatur eelectro-magnee testh f o t t coil unde conditioe th r cyclif no c suppl electrif yo c current c) Sodium corrosion test of the TSAs & iron core d) Irradiatio nTSAse testh f o t, iron corH coi e- Joy& n i l oMK (Components & system tests) a) Magnetic holding SASe forcth f Seo b) Temperature respons TSAe th f eso c) High temperature & thermal shock tests in sodium d) Whole system test in Joyo MK- II ) Developmen(2 M t GE pla r nfo e resultBasee preliminarth th n f do so y desig ne applicatio studth r e yfo th f no GEM system in the DFBR, the main technical problems have been identified as follows: a) confirmation of the neutronic calculation method for the evaluation of the GEM reactivity worth by conducting criticality experiments b) validatio improvemend nan calculatioe th f o t s expansioga n e modeth r nfo l behavior of GE conductioM y - reactob of - t r nou experiment s

124 7. Conclusion

(1) The PSA study concluded that the two independent reactor shutdown systems of DFBR reduce the occurrence probability of ATWS to less than 10 ~ /ry7 . (2) The passive reactor shutdown system SASS was introduced to reduce the risk of core damag ATWo t e Sedu eventsinstallatioe Th . SASf no S reducee sth occurrence probability of ATWS by two orders of magnitude. furthea r (3Fo ) r enhancemen reactof o t r core case safet th ATWf eo n yi S events, bees employmenha n M proposeGE f o t reinforco dt passive eth e shutdown capability against ULOF events. (4) GEM can provide additional diversity in the shutdown system capability against ULOF alsd oan , complemen functioe th t SASSf no . e outlinTh ) futurf eo developmen(5 M e SASe GE plan th d r San describeds fo twa . (6) In Addition to SASS and GEM, ETEM could be one of the promising mechanisms ennhanco t e cor leses safetha s d impacperformance yan th n o t futura f eo e large scale reactor core.

REFERENCES

(1) TJnagaki, M.Ueta, Y.Shibata, K.Tarutani, K.0kada, : "The Developmen Demonstratiof o t n Fast Breeder Reactor (DFBR)", SMiRT 13th Grando Ri , Sulo ed , Brazil, Aug. 1995 (2) TJHoshi, KHarada, H.Endoh, LIkarimoto: , "Development of Actuated Shutdown System for a large scale FBR", ICONE-3, Kyoto, Japan, Apr. 1995 (3) S.Kotake, S.Kasai, K.Nakai, SJtooka, : "Development of Enhanced Thermal Elongation Mechanism (ETEM) for a Large Scale Mixed- Oxide Fuel FBR", ICONE , Kyoto-3 , Japan, Apr. 1995

125 IRRADIATION PERFORMANCES OF THE SUPERPHENIX TYPE ABSORBER ELEMENT

B. KRYGER . GÖSSED , T Commissariat à l'Énergie Atomique (CEA), Centre d'Étude Saclaye sd , DRN/DMT/SEMI/LEMA,

J.M. ESCLEINE Commissariat à l'Énergie Atomique (CEA), Centre d'Études de Cadarache, DRN/DEC/SDC.LEMC,

Cedex, France

Abstract

Several aspects of the irradiation behaviour of the SUPERPHENIX type absorber element are presented in this paper. A large programme of irradiation tests performes wa PHENIdn i assesXo improvo t t d absorbese an eth design pi r n whose main characteristic firse sodiua th t : r loa sfo e mdar bonde vented d an wit n dpi h high borof o highl d ) B4% nan C10 t ) y a pellets enriche0 TD densit 9 % o t .6 p y(9 d(u

We present and discuss the main post-irradiation results obtained by this programme wich concern e behaviouth s f boto r h B4C pellets (fragmentation, swelling, helium release, thermal conductivity evolution stainlesd an ) s steel clad (embrittlement by carburization, mechanical interaction).

It appears that the residence time of the first load of SUPERPHENIX control rod s clearli s y limite y mechanicadb l interaction betwee clade d nth an B,d 4Can particularl relocatiny yb f smalgo l fragment f Bso 4 t beginninCa initiae th f lif go n e li gapirradiatioe Th . n performe residence th x PHENIn fi di o t e firse d timth Xle tf eo loa f controdo e.f.p.d0 l rod 24 analyse e o st . Th effect e th f so s limitin residence gth e time have enable proposo t s du extension ea thif n o measuresso timtw y eb firse tTh . one is reduction of the capture rate in boron carbide. This measure was brought into operatio boroe meath y enrichmenn0 b n 1 % f lowerint no a Be 8 4th 4 C f t gpelleto a t s e loweth e pin n ri e seconth par .f Th o t d measur s preventini e e fragmenth g t relocation by adoption of a thin stainless steel shroud enclosing the pellet stack. The efficienc f theso y e measure s proveswa severan di l irradiation tests (ANTIMAG experiments) in PHENIX. A burn-up of 220 x 1020capt/cm3 was achieved without any dimensiona diametern l pi chang e shroue th f eTh o . d faile t couldbu d nevertheless preven pelley an t t cladding deformation.

Thus, these results hav residencea enablex fi o t s ee.f.p.0 du tim64 r f defo o thire th d SUPERPHENI e loath f do X control rods.

The achievement in the future of lifetime up to 1 000 e.f.p.d. will require the developmen botf o t h advanced absorbing material designsn pi d san .

127 1. INTRODUCTION

e neutroTh n absorbing material e choseabsorbeth r fo n r elementf o s SUPERPHENIX is boron carbide. It is desirable for economic reasons to match the enduranc absorbee th f eo r elements wit hfuele thath steady-statn I f . o t e operating conditions fuee th , l elements have been designe achievdo t eresidenca e0 tim64 f eo e.f.p. e.f.p.d0 2 32 rund( f o s. each). Irradiation experiments were performen i d PHENIX to test the improvements made on the initial absorber pin design in order to increase the residence time from 240 e.f.p.d. (1stload) up to 640 e.f.p.d. (3rd load).

In the first part of this paper we present the criteria wich determine the life of absorbee th rsubsequene pinth n I . t par basie th t c results issued fro irradiatioe mth n tests performe n PHENIi d e presenteXar d with those gathere e literatureth n i d . Finally the improvements made on the initial design which led to increase the performance absorbee th f s o described e ar r pin, .

CONTROE TH . 2 L RODS SYSTEM SUPERPHENIF SO X

2.1. Description

The control rod SUPERPHENIn si ] hav Xmai[1 o etw n functions : wic e har

1 ) the compensation of reactivity effects due to thermal or power variations and to fuel burn-up,

2) the safety function by negative reactivity introduction to assure the reactor power decrease or shut down when incidental or accidental situation occurs.

In orde o increast r e diversificatioeth o improvnt e safetyeth e previouth , s function assuree systemso sar tw y (SystèmP db SC , Commande d e e Principa- l Main Control System) and SAC (Système d'Arrêt Complémentaire - Complementary Shutdown System). Each system includes several subassemblies (S/A) wita h movin containind ro g e par th absorbee - tgth - connecter mechanise on o dt r mpe S/A.

The SCP assures the above two functions and is divided into two groups SCP1 SCPd an 2 eac wite hdifferena on t typ f mechaniseo m agai r diversitnfo y reasons. The SAC assures the safety function, manual or automatic shutdown. SCP1 includes 11 S/As, SCP S/A0 2 1 S/A 3 sanC S d(Fig.1)SA . reaco T requiree hth wortd dro h efficiency wit limiteha d dan numbeA S/ f o r mechanisms in the specific environment of SUPERPHENIX core, the absorber 10 material is enriched boron carbide (B4C, B enrichment = 90 at %).

The SCP subassembly (Fig.2) is realized with two parts, the outer hexagonal shell and the absorber rod. The external shape of the S/A is similar to the fuel S/A absorbee Th oneitseld . ro bundlen rcomposes i fpi e innee th th , y rd b shell e th , upper guid liftine th tube gd heaan , d connecte mechanisme th o dt bundle Th .s eha 31 pins and 12 dummy pins on the periphery to improve the cooling. The pins are

128 sodium bonded and the gap is determined to accomodate the B4C swelling to avoid functioa expectee s i th mechanica p f no ga e d th o s l interactio d f lifo e an d en t na residenc eabsorbere timth f eo .

The pins are vented through the upper and lower plugs to release helium directly in the primary circuit. A schematic view of the absorber pin is given on figures.

O SCP 1 OUTER CORE SCP2 SAC

FIGUR SUPERPHENi: E1 X CORE

129 REACTOT RA REACTOR SHUTDOWN ON OPERATION COUPA EA 1 \ 5 i. M \i L y Course possible IUS Absorber

COUPB EB

ABSORBER c FUEL FUEL I iioo 1000 1558 I ^ 5300 l

SECTIONC

iOEO

"3S; ? °» -^— ; ,^ /'Maille do néseain V hexagonalJ . i \1y

50 ^ X, > ^

300 ooo

r V \ 1 .

FIGURE 2 : SCP SUBASSEMBLY - MAIN CONTROL ROD OF SUPERPHENIX

130 UPPER PLUG

UPPER VENT

PLENUfl

SPRING

PELLETS

REFRflCTORY PELLET WIRE SPRCER (WITH/WITHOUT)

REFRflCTORY PELLET (SUPPORT)

LOWER VENT

LOWER PLUG

FIGUR SCHEMATI: E3 C VIE ABSORBEF WO N RPI

131 2.2. The absorber element end of life criteria

e theoretica e contros determinei th Th f liff o d o e ro d l s en it y llos b df o s reactivity worth. Praticaily the control rod end of life occurs earlier than expected by neutronic calculations e reasoth , n o beint e e absorbefailure th gth du f o en pi r problem f materialo s s damage e lif ee Th absorbe . th tim f o e r e elemenb y ma t determined by one of the following criteria [2] :

- Los f boroso whic0 n1 h reduce effectivenese sth n a controe o t th f d o sro l unacceptable level.

- Swelling of the B4C pellets leading to pellet-cladding mechanical interaction (PCMI).

- Excessive clad embrittlement due either to radiation damage or to pellet- cladding chemical interaction (PCCI).

- Excessively high pellet temperature. The swelling rate of boron carbide is essentially independen f exposuro t e temperatur temperatura o t p eu 500°C1 e~ , above whic e swellinth h g increases rapidly with temperature. Pellet meltint a g 450°2 avoided- e alsCs b i o t .

3. BEHAVIOU ABSORBEE TH F RO R ELEMENT

3.1. BC thermal conductivity and structural behaviour 4

The thermal conductivity of boron carbide reduces very rapidly with burn-up. In unirradiatee th d stat conductivite eth y decrease 1 witT t hsa increasing measurement temperatur. eT

Figure 4 shows measurements performed in different laboratories on irradiated 3 B4C at low burn-up (up to 40 x 10 capt./cm ) and low temperature (500°C). For these irradiation conditions, a drastic reduction in thermal conductivity (more than a factor of five relative to unirradiated material) is observed. Works performed at HEDL [3] indicate that the reduction in thermal conductivity at high irradiation temperature significants a t no 575° 1 s si t materia e A .C th l exhibits about hale th f conductivity of unirradiated B4C.

In the majority of boron carbide pellets cracking generation by thermal stresses occur firsn so t goin poweo gt thed size an f rnfragmentth eo s decreases t leasa , t partly due to the reduction in thermal conductivity. Fragments of quite small size (< 1 mm) can be generated at high ratings and burn-ups.

3.2. Swellin heliud gan m formation

Swelling is attributed to the retention of helium produced by the 10B (n,a)7 Li reaction. The literature reports a general consensus on the therory that lenticular bubbles filled with responsibl e heliuar s mga botr efo h swellin microcrackind gan f go

BC [4-7]. Figur showe5 s collected dat materialn ao f variouso s densitiee th n i s 4

132 thermal conductivity (W/m.K) at 500 °C 20

——• — 16 - - UKAEA (GB)

——— D ——- HEDL (USA)

12 -- O CEA (F)

o CEA (F)

4 --

burnup /cm3.10"20

10 20 30 40 50

FIGUR THERMA: E4 L CONDUCTIVIT IRRADIATEF YO D B4C VERSUS BURN-UP

range of temperature (600 - 1 500°C) indépendant swelling. Swelling is found to be linea burn-un ro p Net, i.e. heliun 0 , o capt/cm 1 mx 5 . concentratio 1 .~ . No o t p nu 22

0 1 m/He highex th r 5 ea sloprfo Fro1. t densitf mo ege e Figw y , .5 ,-29 (96 % TD) pellets. These small values of the extra volume added by each one of the created helium atoms are very close to the values that are obtained with helium

0,4 T

0,3 -

0,2 -

0,1 -

0 50 100 150 200

FIGURE 5 : EVOLUTION OF B4C SWELLING VERSUS BURN-UP

133 under extremely high pessures r Waal ,covolumde e i.e.th n so t ,Va e eth tern i mB equation [8]. Thus, it has been proposed [9] a simple, zero-order, linear model for swelling :

AV/V« BNct

where B is not an adjustable parameter but just the covolume term B in the Van der Waals equation. The simplicity of this model is eloquent. At the first glance, the model disregards the bubble microstructure or shape and simply assumes that the helium is in its most compressed state, i.e., the covolume term in the Van der Waals equation.

The fact that grain boundaries are preferential traps for helium makes the main source of early intergranular microcracking.

Figure 6 shows measurements of the helium retained as a function of burn-up. transiene Inth t swellinperiow lo f e heliudo th g m releas s substantiali e , thet i n decreases to low values and then increases again for burn-up > 100 x 1020 captures/cm3, although the swelling continues unabated.

8x1021 ' FULL RETENTION

c so I5 O O

m FRANCE

BURN-UP (CAPT/cm3)

10 20X1021

FIGURE 6 : MEASUREMENTS OF HELIUM RETENTION IN B4C

3.3. Pellet-cladding mechanical interaction

Pellet swelling is considerably faster than that due to void swelling in the cladding material. Therefore, dependin initias it n glo value, pellet-claddin wilp glga ten closo dt e

at a certain burn-up. Gap closure and continued B4C swelling then results in mechanical pellet-cladding interaction inducing cladding strains and stresses which are potential causes of absorber pin failure.

134 The SCP first load absorber pin, whose main characteristics are presented in table 1, is designed with a gap equal to 11,8 % of pellet diameter allowing a free volumic swelling for B4C up to about 40 %. If we consider the swelling law of B4C (96 % TD) (fig. 5), it can be seen that such a pellet-cladding gap should allow the control rod to reach a burn-up equal to 250.10ao capt/cm3 without mechanical interaction. The irradiation experiment PHENIn i s X called PRECURSAB were conducte orden i d o t r determin life e eabsorbeth e timth f eo r pin t appeare.I d tha residence tth n pi ee timth f eo was clearly limited by the mechanical interaction between the B4C and the clad and particularl presence th y yb f smaleo l fragments thermae (duth eo t l stress crackind gan

TABL CHARACTERISTIC: EI E FIRSTH F TSO LOAD OF THE SUPERPHENIX SCP ABSORBER PIN

Design Sodium bonded and vented pin

Material B4C Boro enrichmen0 n1 t % 9t 0a

Density 96% TD

Pellet diameter 17mm

Column length 1 145 mm Cladding material 316Ti

Cladding diameter 19x21 mm

Pin length m m 1 131

intergranular microcracking), which relocate at beginning of life in the initial gap, leading to clad failure when burn-up goes beyond 150 x 1020 capt/cm3 (fig. 7). As a consequence of B4C dispersion in the gap we also observe that the B4C columm length after irradiation is lower than that expected from swelling. Figure 8 shows the evolution of B4C diameter combinatioe increasth o t e edu swellinf no relocationd gan , inducin prematurga e pellet- cladding mechanical interaction.

3.4. Pellet-cladding chemical interaction

welIs i t l known tha presence tth f boroeo carbod nan stainlesn i s steel produces severe embrittlemen lone th gr proposefo d an t d lifetimes these penetration effectn sca be limiting factors. Figure 9 illustrates the carbon penetration in the cladding on the experimental vented and sodium bonded pin PRECURSAB A1 irradiated to 250 e.f.p.d. in PHENIX.

135 o FIGURE 7 : EXPERIMENTAL ABSORBER PIN PRECURSAB A1 IRRADIATED

IN PHENIX. RELOCATION OF B4C FRAGMENTS (2 AND 3) IN GAP AND SUBSEQUENT CLAD FAILURE (1). IL

1,13- sweltin relocatiogr - n

1,10-

swelling

1,05-

CAPTURE RATE 50 ico 150 -•» ^ ""J FIGUR DIAMETRA: E8 L EXPANSIO COLUMC B4 F NNO DU SWELLINO ET RELOCATIOD GAN N

136 PRECURSABA1 (M316 CLADDING) 13 2 Effective Diffusion Coeff. Deff = 4,5 .10- cm /s Clad Internal Temperature 565°C (Max.) EFP0 Tim25 De=

100 200 PENETRATION

FIGURE 9 : PENETRATION OF CLADDING BY CARBON DIFFUSION

4. ABSORBER PIN DESIGN FOR THE THIRD LOAD OF SUPERPHENIX

irradiatione Th s performe PHENIdn i residence th x Xfi o havt efirse d timele th t f eo load of SCP to 240 e.f.p.d. The analyses of the limiting effects on the residence time have allowed to propose an increase of this time by two cumulative ways. The first one is the reduction of the capture rate in boron carbide. As in a pin the shape of the capture rate is very sharp it is possible to reduce the maximum rate by decreasing the number of boron 10 atoms in the lower part without significant effect on the rod worth. This can be achieved with use of less enriched boron carbide or by a lower density.

The solution which has been chosen for the second SCP load (320 e.f.p.d.) and the following loads, consists to adopt a 48 at % of boron 10 enrichment for the lower pinse parth f , o t {Fig. 10).

The second improvement made on the initial pin design, is to prevent the fragments relocation. It consists of enclosing the pellets stack inside a confining shroud (thin tubular sheath) made with similar material as the cladding.

Four irradiation tests carrie PHENIn i t dou Xcapsulea (ANTIMAn i 3 d , an G2 HYPERBARE 1 and 2 in a control rod) have allowed both improvements to be validated successfully. ANTIMA Gexperimen3 t achieve maximue dth m performance (657 e.f.p.d. and 220 x 1020 capt./cm3) without clad deformation. The behaviour of this experimental absorber pin is illustrated by figure 11 on which we can observe presence th beneficiae resula th residuaa s f f a o e o t p ga ll shroude effecth f o t . Carbon local measurements from X-rays microbobe analysis, revea similaa l r levef o l carburization (1,8 w % C) at surface of shroud and clad.

137 60 80 100 120 heighfrom c mn n pi i t botto e th f no FIGURE 10 : AXIAL REPARTITION OF BURN-UP IN THE FIRST AND SECOND LOADS OF SUPERPHENIX SCP ABSORBER PINS

1 mm

FIGURE 11 : ILLUSTRATION OF THE USE OF A TUBULAR SHROUD IN EXPERIMENT ANTIMA G(653 7 e.f.p.d10x 0 20 capt/cm22 .- 3)

138 Considering the overall irradiation experiments carried out in PHENIX, the improved absorbe whosn pi r e main specification givee s beear s tabln ha ni n, e2 introduced in the third SCP load for a residence time of 640 e.f.p.d.

TABL CHARACTERISTIC: I EI THIRE TH DF SLOAO D OF SUPERPHENIX SCP ABSORBER PIN

Design Sodium bonde vented dan n dpi Double enriche shrouded dan d column

Material B4C Density 96% TD

Pellet diameter 17.4 mm Column length m m 5 114

Boron 10 enrichment 90 at % in upper part (995 mm) 48 at % in lower part (150 mm) Cladding material 15-15TÏC.W. Cladding diameter 19,8x21.6 mm Shroud material 316Tior15-15TI Shroud diameter 17.6 x 18 mm Pin length 1 318mm

5. CONCLUSION

The design of the SUPERPHENIX SCP absorber pin was a significant evolution from the PHENIX one and the performances required were far above. A lot of irradiation experiments were performe suppordo t designe th toverale th f I . l desigs ni well confirmed and functions required totally fulfilled, the analyses of the irradiation tests have led to limit the residence time of the SCP first load at 240 e.f.p.d. But significant improvements made on the initial design (adoption of two axial boron 10 enrichment zones in the B4C column and a shroud to prevent relocation of B4C fragments), allo residence wth thirP e.f.p.d.d0 SC fixe e loa64 e b tim t th ddo a t f e,o wich correspond residence th o st efuee timth lf epino .

To achieve in the future a residence time of 1 000 e.f.p.d., it will be necessary to tesendurance th t absorbee th t burn-up f a eo n 10aboux o t pi r 0 2p 0s u capt./cm30 t 3. Suc increasn ha f lifeo e timburn-ud ean p will requir solutionfinew o t dne delao st e yth advent of both mechanical and chemical interactions between B4C pellets and cladding.

139 REFERENCES

[1] FRANCILLON , JALLADEE. , , LANGUILLEM. , , ROUCHESA. , controe Th , lF. , rods in SUPERPHENIX 1, KTG metting, BONN (1991).

[2] KELLY, B.T., KRYGER, B., ESCLEINE, Ü.M., HOLLER, P., Int. Conf. on Fast Reactor and Related Fuel Cycle, KYOTO (1991).

] HOLLENBERG[3 , G.W., BAKER, D.E., DOE/CEA-DEBENE Semina n Faso r t Reactor Absorber Materials, SUNNYVALE (1983).

[4] PITNER, A.L., Nucl. Technol (1972)0 41 , .16 .

[5] JOSTONS, A., DU BÖSE, C.K.H., COPELAND, G.L, and STIEGLER, J.O., . NuclJ .(1973/74) 6 Mater13 , .49 .

[6] HOLLENBERG, G.W., MASTEL BASMAJIANd an , B. , . Ceram, J.A.Am . J ,. Soc. 63,376(1980).

[7] HOLLENBERG, G.W, and CUMMINGS, W.V., J.Am. Ceram. Soc. 60, 520 (1977).

[8] COGHLAN, W.A. MANSOURd an , , L.K. . NuclJ , . Mater. 122-123 (1984)5 ,49 .

[9] STOTO, T., HOUSSEAU, N., ZUPPIROLI, L, and KRYGER, B., J. Appl. Phys. 68 (7), p. 3198 (1990).

140 OPERATION EXPERIENCE OF THE BN-600 REACTOR CONTROL RODS

V.V. MALTSEV, V.F. ROSLYAKOV, G.V. BABENKO, A.N. OGORODOV Beloyarsk NPP BN-600, Russian Federeation

Abstract

This paper reviews the major experimental results on different types of control rods with boron carbide and europium oxide which have been tested in the BN-600 reactor during the operating period.

INTRODUCTION

Shime rods (SHRsa compensatio e user ar )fo d n a f o n inlet core temperatur d thermaan e l power reactivity effects and burn-up effect. There are 19 SHRs in the core. Automatic power control rods (ACRs e user ar )fo d automatic maintaining and change of the preset reactor power level. There are 2 ACRs in the core. Scram rods (SCRs) are used to scram the reactor in emergency situations (all SCRs) or to decrease its power abruptly by 1/3 of preset power level in the mode of trippin e coolinon g g loop (one rod, named SCR-L ). Ther e are 5 SCR and one SCR-L in the core. During the operating period 5 different types of SHRs and 3 types of ACRs with boron carbide and europium, oxide as an absorber and 7 types of SCRs with boron carbid f variouo e s enrichment n absorbelevea s a l r were tested in the BN-600 reactor. Basic ideas behind rod testing were the following: e utilizatioth - w structurane f o n l materials more irradiation resistant for longer liftime;

- the validation of "neutron trap" design using ZrH2 a moderatos d f shapannulao ro rs wel a es d a lan r absorbers of annular and rod shape to decrease the absorbing material amount in a rod; - functional testin f sealeo g d absorbers.

141 SHIME RODS

Shime rods (SHRa compensatio e user )ar fo d n a f o n inlet core temperatur d thermaan e l power reactivity effect s wela ss reactivit a l e burningyth changeo t -e du s up of the fuel in fuel assemblies. The basic design SHR characteristic e th operatio d an ns conditione ar s presented in Tablesl, 2, and 3; principle design is shown in Fig.l. Rods, define s 1161a d , 1161A d 1161an , n TechnicaBi l documentation, were designed as SHRs for BN-600. Their principle distinction is the absorbing material used. For

the rod 1161 the absorber is made of B4C while in rods d 1161an 1161BA Eu203 +Ms used i oe peculiarit Th . l al f o y the rods is the chromium-nitride coating of the hinges preventing them from selfwelding with a case during operating under the influence of high temperatures. The design of the 1161A and 1161B rods is the same. The differenc s onl i en absorbe i y r manufacturing technology. The absorber of the 1161A rod is made of the hot-pressed pellets while the absorber of the 1161B rod is made of powdere basith cs A d .basevarian ro n europiu o de th t m oxide was adopted and for the 1st charge SHRs of the 1161A and 1161B type were produced and installed. Operation experience showed that the main factor limiting the life-time of the said rods was the sufficient shape-changin e botto e th part th f m o sf o g hinge under the influence of irradiation. That led to the jamming of a rod during the 3rd recharging and produced some difficultie e extractioth o rodsr tw e fo sTh .f o n following measuring of the spacing part diameters of the bottom hinge of the said rods showed their increasing up to 2.4%. To prevent this the second set of rods was improved e foolowinth in g manner (prior inserting e intth o reactor): - spacing part diameter e bottoth f mo s hinge were decrease do 89.2-89. t fro 0 9 m ; 3mm

142 TABL . BASIE1 C DESIGN CHARACTERISTIC PERMANENF SO T CONTROD LAN SCRAM RODS FOR BN-600 REACTOR

Characteristics Shimd eRo Control Rod Scrad mRo

1161 1161A, 1161B 11611 B 157A 1663 1663-01

1. Hinges amount, pcs . 2 2 2 2 4 2. Working links amount, pcs . 1 1 1 1 2 3. Absorbers amount in a working link, pcs. 19 48 8 31 7 4. Case diameter, mm 89x1.5 89x1.5 85x1 73x1 73x1 5. Absorber cladding diameter, mm 15.6x0.1 9.5x0.5 32x0.7 9.5x0.5 23.X07 6. Case material OX18H10T 06X18H10T 08X18H10T 08X18H10T 1X18H10T 7. Absorber cladding material 3M-847 3M-847 3M-847 3M-847 3M-847 8. Hinge material 12X18H10T 12X18H10T X18H10T X18H10T X18H10T 9. Absorbing compound length, mm 780 780 740 410+410 414+300

10 . Absorbing compound B4C Eu203+Mo B4C Eu203+Mo B4C 11 . Concentration of B10, I' natural - natural - 80 natural 12 Absorbe. r type sealed sealed non-sealed sealed non-sealed CO pellets pell., powder pellets pellets pellets TABLE 2. OPERATING CONDITIONS FOR PERMANENT CONTROL AND SCRAM ROD BN-60F SO 0 REACTOR

Characteristics Shimd Ro e Control Rod Scram Rod '

1161 1161A, 1161B 1161B 1157A 1663 1663-01

1. Physical effectiveness, not less than dk/% , k 0.27 0.27 0.27 0.27 0.28 0.62 0. 0 2. Fuel burn-up ß'10, % 24 _ _ 18 ' - 12.5 (per year) 3. Operatine g th tim n i e reactor (resource), eff.d. 180 180 180 365 365 730 4. Rated sodium discharge 6 3. throug e case 6 th h 3. , kg/c. 3.6 3.6 3.6 1.7 5. Maximum absorber cladding temperature, °C 620 493 493 595 498 608 6. Case inlet temperature, 380 380 380 380 380 7. Operating media liquid sodium TABLE 3. CONTROL AND SCRAM RODS IRRADIATED IN BN-600 REACTOR

W Type Amount, pcs . Maximum damaging Operating dose range, time, eff. d.

1. SCR 1663 65 10-47 88.0-365.3 2 .R 1663-0SC * 1 12 10-45 93.6-331.6 3. SCR 1641 2 18-19 161.2 4. SCR 2233 4 37-45 290.4-348.4 5. SCR e3.1583 1 31 192.2 6. SCR e3.1584 1 46 331.6 7. SCR 63.1469 2 33-49 190.0-285.5 8. SCR 263* 7 1 64 502.3 9. SCR 2633 1 63 502.3

10. SHR 1161 3 18-19 93.6 11. SHR 1161A 40 19-50 88.0-240.1 12. SHR 1161B 78 14-79 97.5-391.1 13. SHR 1161B 99 14-56 97.5-422.8 14. SHR 2635 9 66 502.3

15 .R 1157AC A 23 15-45 97.5-348.4 16 R .241AC 5 2 45-68 348.4-510.1 17. ACR 2631 1 66 502.3

SCR - scram rod SHR - shime rod ACR - control rod r trippin fo Scra- * d e coolinro mgon g loop.

145 A - f- Supplant

D-D

) 6 ) a FigShim. 1 .witd ero casha e design. a) 1161A rods with a round case ) 1161b B rods wit hexahedroa h n case

146 - tip diameters were decreased from 80 to 78 mm. Beginning from July/ 1982 n ordei , o decreast r e irradiationa e massivl th hea n i te hinge o t joint d an s improve the conditions for the case contact, the design e spacinoth f g elements, hinges d shanan , k tail parf o t 1161A and 1161B rods was changed. In 1982-1983 8 rods of 1161E-3n type were tested. For their hinges ferrito-martensite steel 311-450 was used instea f o 12X18H10d T steel e measurine Th th . f o g dimensions of the spacing rod parts made of this steel showed that after operating cycle they incresed slightly (appr e diamete e workinth . th t 0.1% f bu o r)g link jacket pipes (12X18H10T steel) swellehinge th eo t ddimensions . Despite high irradiation resistance and absorber efficienc e considerablth y e disadvantag R wit- SH Eu hf o e based absorbee higth h s i rinduce d e activitth d an y residual energ e accumulatioyth releaso t e du ef highl o n y active Eu-152 and Eu-154 long-live isotopes with hard gamma-irradiatio. n 6 1 spectru d a an half-lif( m3 1 f o e years correspondently). Becaus f tha o ee advancen 198 i tth s 4 wa R d SH typ f o e introduced, defined as 1161B (Pic.3.1), the working link of that rod comprised 8 absorbers, 23 mm in diameter and n lengthi m m .0 Seve74 n absorbers were uniformly spread along the rod circle and one absorber was put into the middle. Fourteen displacers were spread in the empty space between the absorbers - 7 in the inner row and 7 in the outward row. Thus for 1161B rods absorbers unified with csram rods were used (non-sealed ones with an absorbe f boroo r n carbid f naturao e l concentration)e th ; coolihg of the scramble cases was improved due to the utilizatio f jackeo n n diametei tm m pip 5 r8 e insteaf o d e structuraTh . 8mm 9 l materia r the s fo exactlli m e th y r 1161d 1161fo an sam ABs a e rods. This modernization allowed to increase the life-time 5 eff36 o .t p daysu R . SH SucR desige SH hoth fs i n currently used in the core of BN-600 reactor. In 1990-199 n advancea 1 R 263SH d 5 wit a hlife-tim e of 550 eff. days was tested. The main distinguishing e utilizatio th feature rod th s i sf o ef 311-45o n 0 steel

147 for jacket pipes, hinges, and shank parts, and the use of austenised yC-68 r absorbersteefo - l s claddingsd ro e Th . was designed to operate for 550 eff. days. Although it manage o operatt d e withou a complait2 day 50 sr onlfo n y due to the reactor operating conditions. In the operation course the shape-change control of s regularwa R thSH ye performe o correct d t life-time resource. The diameter of the spacing belts of the bottom hinge and that of the spacing part of rod extension was measured. The measurements took place in a "cooling" pond and in a "hot cell". Total 58 rods were tested including 19 - 1161A, 28 - 1161B (of them 6 - 1161E-3II) , 10 - 1161B, 1 - 2635.

CONTROL RODS

Automatic control rods (ACRs), of 1157F type, are designe o automaticallt d y maintai d changan ne prese th e t reactor power level. The basic design ACR characteristics and its operation condition are presented in Tablel, 2, e principlth ; 3 d ean desig s showi n n Fig.2i n . e samth e d disadvantageha t varianThR 1s eAC f o t s a s R witSH he oxidth e e SHRabsorberth sr fo s wel A .s a l there was a rod jam in a case during a recharging. As a result the rod design was sufficiently changed. Mostly it relate e absorbee hingth th o ed t d an rjoint e absorbeTh . r was changed from a pellet to a powder made (just like for e thhing th s econcerne wa s e1161a r e fa Bth d rod s A ). chromium-nitride coatin s excludedwa g e weighth ,s wa t decrease e coolinth d gan d conditions changed. Five 1157A rods were studie o controt d e shape-changinth l e th d an g results obtaine dSHRsr werfo e sam .s th ea e 241d 263an 5 1 rods were e reactotesteth n s i a dr experimental ones. The 1st differed from the permanent ACR in the construction material selected for cap, shank, hinges d jackean , t pipe (more irradiation resistant 311- 0 steel45 ) s chargemanufacturewa d ro d e inte Th th .o 2 effreacto31 .r days fo re maximu th , m increase th f o e bottom hinge diameter was 0.1%. As far as the 2nd ACR is f o concerned e us e e differenc th th t ,onl n no i y s wa e

148 Supplant

Supplant

a) 0)

Fig. 2. Control rod with a case design. a) 1157A rods with a round case ) 1157b A rods wit hexahedroa h n case

149 irradiation resistant 311-45 d HC-6an 0 8 steel t alsbu s o

Eu2U3+Mo was replaced by B4C. This ACR was operated for 2 eff50 s dimension. it t changed dayd no an sd di s .

SCRAM RODS

Scram rods (SCRs) were designe o scra t de reacto th m r in emergency situations (all SCRs) or to decrease its powe f rpreseo abruptl3 1/ te mod th f levey o b eyn i l tripping one cooling loop (1 SCR). The basic design characteristics of permanent SCR and its opereating condition e presentear s ; principl3 n Tableld i d an , e2 , design is shown in Fig.3. Beginning from the 1st microcampaign, SCRs 1663 and 1663-01 were used as permanent rods. .They were practicall e sam n th designyi e e basith , c difference between them was the amount of B10 content in absorber composition. Durin e firsth g t year f operatioo s n there also were some difficulties in recharging due to the swelling of the bottom shank of the bottom link - 2 SCRs were jamme n 'rechargini d g box. That resulte n SCRi d s modernization e maiTh .n changes were a hingmad o t e joint, absorber bottod an , m link: the cooling condition a hing r efo s joint were improve d chromium-nitridan d e coatin s excludedgwa ; e absorbee lengt p workinth th to f - o hn i rg lins wa k , becausmm 0 33 e o thit sm m , frodecrease4 mm 41 m4 8 y b d part did not substantially affect the total rod effectiveness (the position of the bottom border of the t changed)no s wa ;d ro absorbe e th n o r - the diameter of the bottom link mostly affected by the irradiation was decreased from 74 mm to 68 mm. Shape-changing contro s performe wa 0 l1661 r 3fo d - 1663-01 3 r e bottofo Th . SCRd m an sshan k diametes wa r measured. The results obtained coincided with that obtaine r SHRACRsd fo dan s .

150 78

Fig. 3. Scram rod with a case design. a) 1663, 1663-01 rods wit rouna h d case b) 1663, 1663-01 rods with a hexahedron case 1 - directing pipe of executive facility directin- 2 d ro e g th cas f o e 3 - rod 4 - fuel assemblies surrounding the rod

151 Seven types of rods were tested as SCRs. The basic ideas in their design were: the utilization of advanced irradiation stable structural materials to prolong the life-time (2637, 2633 type) ;

e validatioth - f "neutroo n n trap" design using ZrH2 s a moderatoa s l d we f shap annularo o s ra ed an r absorber d shapf annulao sro o decreast ed an r e amounth e t of absorbing material in a rod (2233, e3.1583, e3.1584, e3.1460 type); functional testing of sealed absorbers (1641, e3.1460 type). e experimentath l Al l rods were functioning during the defined time withou y faultsan t .

152 THE EXPERIENCE OF POST IRRADIATION INVESTIGATIONS OF THF, BN-600 CONTROL RODS

V.P. TARASIKOV, R.M. VOZNESENSKJ, V.A. RUDENKO IPPE, Obninsk, Russian Federation

Abstract

The paper present resulte sth post-irradiatiof so n examinatio BN-60e th f no 0 reactor control rodsrode Th s. wer operation ei r mornfo e thadays0 bum-ud n50 an , s pwa 18.3.1021 capt/cm3.

1. INTRODUCTION

e controTh e lreacto th rod f o s r BN-600 were developed accordin o theigt r functional purpose and the experience of operation of the similar rods of the Bor-60, BN-350 reactors. e e desigBN-60th Th f o n 0 contro e lfirs th rod tf o sloadin s considerei g e n detaii dth n i l reports of Soviet specialists at the Meetings of International Working Group on Fast Reactors of IAEA. Thus thin i , s pape t won'i r adducede b t . The life-time of the control rods of the initial complete set was ~200 EFPD and it was limited by the dimensional changes due to irradiation of the parts of the rods working in the core. According to that, the design of the control rods and wrappers was changed with the o increast m ai s reliabilitit e d life-time e experimentayan th r Fo . l confirmatio e desigth f o n characteristics of the rods and the efficiency of their design modifications the earring out of the post irradiation investigation of the spent rods was planned. Two stages of rods upgrading could be distinguished according to the rods modifications. At the first stage (1981-1985), the modifications of the parts of rods that allowed to decrease its swelling were made. The use of hexagonal guide tube practically excluded the temperature irregularit perimetee th n yo guidf o r bowine e th tube d gsan cause. it y db In 1984-1985 four experimental compensating rods (CpR) with natural boron carbide (S/A Mil61B) were e BN-60testeth n i d 0 e life-timreactoth r f fo 310ro 6 eEFP 40 , D (damage dos 4 dpa-6 e , burn-up 5.2*1021 capt/cmt cel f Institutho o l e th f 3Physicn )o eI . s d Powean r Engineerin 116R Cp 1 d wit above sealee gro th hth (IPPEe f eo th d rodo d tw an )s absorber pinoperatiod san n EFP4 tim9 f eo D were investigated. resulte Th f investigationo s s permitte substituto dt witR rode th ehCp s e E^Oth y sb boron rods e life-tim e boroTh . th f no e rod s determinei s s limitei 5 EFPt i 36 y d b ds Da dan efficience th structurae th f o y l material. During the second stage of upgrading (1986-1994), the post irradiation investigations of standard rods were carried out 199n .I regulae 0th r(S/ R safetS Ad yro 1663 ) with timf eo operation of 312 EFPD, dose -44 dpa, burn-up 7.3*102i capt/cm3 was investigated in IPPE. The life-time of this rod was determined as 365 EFPD and also it was limited by the efficienc f structurao y l material. At this stage of upgrading, the structural material of the absorber cladding and elements of the rod was replaced by the material with higher irradiation resistance (HC-68 and 311-450 steels, respectively). In the regulating rod RR the natural boron carbide was used e rod, SR-Th .SR s L (th r edro safetfo powe n p i d e hea ro ycasn on i rtf eremovao l loop cutting off), RR, CpR were in nominal operation 502.3 EFPD (dose -70 dpa, burn-up of the enriched and natural boron carbide 18.3*1021 and 8.7*1021 capt/cm3). The post irradiation investigation f theso s e rods showed that after operatio e rodth n s save e margith d f o n performance. In this pape e resultth r f posso t irradiation investigation e experimentath f o s l rodf o s the reactor BN-600 with the life-time of 500 EFPD are given. For the comparison the results of investigations of the rods of the BN-600 and BN-350 reactors with shorter life-time are also given.

153 DESCRIPTIOE 2TH . RODF NO S DESIGN

The principal differences of the upgraded rod from the regular one are: workine upgradeon e s th gha • sectiond dro ; s structuraa • l material e 311-45th s 0 steel (fo rparte ducts th f hinger o s fo feetd ,d an san ) the HC-68 steel (for the cladding) are used; • in the control rod the boron carbide with higher B10 content (92% at.) is used; • in the regulating rod the natural boron carbide was used instead of absorber on the base of Eu203. maie Th n design characteristic e rod presentee th sar f so Table th . n dei I

TABLE I. THE MAIN DESIGN CHARACTERISTICS OF THE EXPERIMENTAL RODS

N? \ type, SR SR-L RR CpR parameter \ serial N° (S/A (S/A (S/A (S/AM2635) JvJ?2637) N°2633) N? 2631) 1 length of the rod, mm 2100 2100 2100 2874 2 diameter of the duct of 73 x 1 73 x 1 1 7 3x 1 x 5 8 working sectionm m , 3 externald diametero e th f o r 74 74 74 88 e hingesth m n m o , 4 number of the absorber pins 7 4 4 8 5 diameter of the absorber 23 xO.7 pin, mm 6 absorber B4C B4C B4C B4C at.(B% )10, (92% at.) (natural) (natural) (natural) 7 lengt f absorbeho r pin, 800 850 850 930 mm 8 diameter of absorber, mm 19.6 9 width of axial gap, mm 28 45 45 28 10 material of the duct, parts of the hinges and tips, mm 13Cr-2Mo-Nb-P-B (311-450) 11 materia claddine th f o ld gan replacer 16Cr-15Ni-2Mo-2Mn-Ti-P-B (HC-68)

f one-sectioo e us e nTh desig borod nan n carbide with higher enrichment permitteo dt f increasleso se swellinus e efficienc ee th e resul e rodth th f th s ~20%y go stb f A o ystructura. l materials the calculated life-time of the developed rods is raised from 365 EFPD up to 550 EFPD. 3. THE CONDITIONS OF TEST

The experimental rods were in operation in the BN-600 reactor for -502 EFPD. The condition f operatioso presentee nar Table th . n di eII

RESULTE TH . 4 POSF SO T IRRADIATION INVESTIGATIONS

Durine reacto e rodth e testth n gi th sf ro s (during charge, discharg d operationean ) there wer troubleo en s with moving efficience Th .e rod th sf o ycorresponde desigo dt n values accounting boron burn-up,

4.1. Structural material

No changes of the rods construction able to make worse their performance were detected. There were no mechanical deformations; the hinges saved their original mobility.

154 TABLE IL THE MAIN CHARACTERISTICS OF RODS IRRADIATION

N» parameter \typd ro f eo safety rod safety rod regulatind gro compensatind gro (S/A (S/A (S/A JJÔ2631) (S/A M?2635) M°2637) M?2633) 1 tim f operationeo , days 502.3 2 burn-upx ma . at % , BIO 20.5 30 . 48 44 B, 19 6 9.5 9 *1021 capt/cm3 18.3 5.5 8.7 8.2 3 max damage dose, dpa: rod structural material, 67 67 71 65 claddinn pi g 22 25 68 59 d ro 4 e x fluencth ma n o e construction (E>0.1 MeV), n/cm2 1.56*1023 1.6*1023 1.62*1023 1.55*1023 5 max calculated temperature of boron 1040 670 835 725 carbideC ° , 6 max calculated cladding temperatureC ° , 485 445 445 435 7 d x ro increasma f o e diameter, mm 0.4 0.25 02 0.28

changee Th s which took constructioe placth n ei e rod unimportante th ar sf no maximue Th . m increas f rodeo s diamete betweep decrease ga 0.2-0.s th e , wa rth nf 4mm eo guid d ero tub d ean 4-8%s wa . The material of ducts and feet retained sufficient workability after irradiation. Its mechanical propertie characterizee ar s hige th hy db leve f strengto l markea d han d margif no

residual ductility (at the dose of 70 dpa at the nominal temperature - a=1000 MPa, cr~800, u y e uniforth m relative lengthening ou=0.8%) e brittlTh . e fracture sampleth f o e s wasn't observed. claddinge Th s changed their sizeresula s sa f swelling o t e maximuTh . m swelline th f go claddin e , rod SR-L th s 0.7d SR 5.7%sn wa i an g, , 1.6R 3 RR, 5. respectivelyCp , , e Th . researche e mechanicath f o s l propertie e nominath t a s l temperatures showed thae th t irradiatio e increas f claddinth o n o t d f ultimat o ed an gle a e MP strengt 5 83 o ht fro0 m50 decreas f ductiliteo . 2% y o frot 3 m2

4.2. Absorber material

e absorbeAth s r materia e boroth l n carbide produce direcy db t synthesi f carboo s d nan boron was used. It is used in the form of cylindrical pellets with the density ~2.1 g/cm3, made of boron carbide powder by sintering and pressing. The pellet stack is enclosed in the stainless steel cladding with the axial and radial gaps to compensate its swelling. In the investigate e venteth s d dSR absorbe r pin e used sodiue ar s Th . m floods throug e weeth h n pi p pluto ge durinth d loadinggro e releasth ; f heliuH producee o als C d . oman B carrien i d s up through this weep. conditione Th f absorbeso r test presentee sar Table th . n deII i The results of these and earlier investigations of the safety rods of BN-350 and BN-600 reactors permitted to reveal basic laws of boron carbide performance under irradiation (in case of vented pins) for the validation of life-time of fast reactor safety rods. The obtained results of boron carbide performance under irradiatio givee nar n below.

155 4.2.1. Dimensional changes of B4C pellets (swelling)

e swellinTh f borogo n carbid e rangth f studien eo i e d temperature d burn-upan s p (u s to 19*1021 capt/cm3, 450-900 °C) directly depends on the burn-up at the constant irradiation temperature swellinC 4 B e rat f Th eo . g (th ediameten ratipi f o r changin percentagen g(i o t ) the quantity of burned up boron atoms expressed in per cents) depends on irradiation temperature non-linearly. The minimum rate of B C swelling is -0.1-0.4 and it was observed temperature th e rang it nth A f temperatur eo . °C e0 lowe50 d e ran betwee thaC ° n0 500°n47 C it decreases with the temperature increase; and at the temperatures higher than 500°C it increases. The ratio of pin diameter changing for the enriched boron carbide (60-90% at.) within the range of temperature of 400-850°C does not exceed 0.6% per % at. of boron burn-up (fig.l) r naturaFo . l boron carbid t doe i et exceef burn-uo no s . at d p1.0% withir %pe e th n range of 450-750°C (fig.2). e investigationTh f boroo s n carbide froe ventemth d pins shown thae interactioth t n between sodium penetrating inte pelleoth t crack d moisturan s e durin dismantlinn gpi d gan pellets washing leads to additional dimensional increasing of pellets. The investigations of absorber from the experimental rods SR, SR-L, CpR, and RR shown that pellet-cladding gap is retained. The maximum increase of pellets diameter was, respectively, 6.1, 2.9, 7.6 d 7.9an , % (the permissible increas e- 10.2%) .

0.8

-Safety system 263A ,S/ 7 -Safety syste BN-35f o m 0 -Safety system of BN-350

900 IRRADIATION TEMPERATURE , °C

FIG. 1. Linear swelling rate of boron carbide (60-90 at. % B-10) as a function of irradiation temperature.

156 -SS-Loop S/A 2633 -CR S/A 2635 -RR S/A 2631 -CR S/A 11613/02 -CR S/A 1161B/03

700 750 IRRADIATION TEMPERATURC ° , E

FIG. Linear2. swelling rate boronof B-10) % function a carbide at. as irradiationof (20 temperature.

4.2.2. Structural changes of boron carbide

During irradiatio pelletC e crackinB nth e s complea th occurs s f go ha t I x. character showin e participatiogth f differenno t cracking phenomen thin i a s process. Accordine th o gt characte f crackso r , three main cracking phenomena coul distinguishede db . The first is connected with the thermal stress effect. The cracks caused by it were directed inside the pellets from its surface. This type of cracking is not connected with the absorbers burn-up because the decrease of the thermal conductivity under irradiation completes as a whole at the small burn-up and the thermal stress level is fixed in a short time after reactor e fulgointh l o t gpower e longitudinaTh . l crackin s observei g n multitudi d e whe radiae nth l temperature changin abous gi t 110-12. 0°C The second is associated with the appearance of the stresses connected with the irregularity of the burn-up along the pellet cross-section. It is displayed in the form of cracks directed along the side surface of pellet and belongs only to the absorber of safety rods operate softenee th n di d spectrum. e thir Th s connectei d d wite violenc th e habsorbe th f o e r structure effectee th y b d products of the B^(n,a)Li' reaction and displayed in the form of fine cracking of the absorber becaus s saturatioit f eo micry nb o fractures e micrTh . o fractures hav n arbitrara e y directio d reflecan n t technological imperfection e absorberth f o se presenc th ; f sodiuo e m apparently promotes to the activation of this process. This effect is displayed with the achieving of the burn-up level of 4-6*1021 capt/cm3. t shoulI notede db pelleC , B thate crackinth t g during irradiation does t leait sno o d t fragmentatio e venteth r dfo n sodium bonded absorbe e rresul th pins f s crackingo tA . e th , stress relaxatio f pelletno s occurs pellete Th . s extracte havn d pi fro evisibla e mth e cracking net, but remain as if intact. During the investigations of big number of vented pins it was observe caso n df destructio o e e claddin th e absorbe f th o ny gb r fragment t intgo s o pellet- cladding gap.

157 4.2.3. Thermal conductivity of boron carbide

e thermaTh l conductivit f boroyo n carbid e importanth s i e t characteristic, influencing s oirradiationit n stability parameters (helium release, structura dimensionad an l l changes). The irradiation leads to the essential decrease of the thermal conductivity; here it is observed initially sharp and then smooth (with the damped rate) decrease of the thermal conductivity. For the synthesized boron carbide the initial thermal conductivity is 15-25 W/(m*°K). At the burn-up -1.5*10 capt/cm the decrease of thermal conductivity is about W/(m*°K))4 ~ o t 70p f initia% (u o e increasin e Th . on l f irradiatiogo n temperature reduces the rate of the thermal conductivity decrease. It could be affirmed, however, that there is no 21 outstanding recoverthermaC B f o yl conductivit e burn-upth t a y s more than 0.5*10 3 capt/cm and the irradiation temperature not more than ~870°C (the maximum average B C temperature in the reactor BN-600 is 700°C for the safety rods and 550°C for CpR, RR).

BORO% BURN-U. Nat , P

FIG. 3. Thermal conductivity as a function of boron carbide burn-up.

4.2.4. Helium retentio releasd nan e

problee Th f heliumo m pressur e claddinth n eo g e ventedoe t existh no sn i td absorber pins. Thus, the data on helium release are not presented here in detail and helium retention is considered only in the meaning of influence on the structural and dimensional absorber changes. It could be noted, that the helium release depends on the technological parameters (density, granule value, impurity presence irradiatioe th d an ) n conditions (temperature, burnup). Wit increasine hth burn-ue th -1.5*1 o t f go p pu 0 capt/cm initiall increasine yth f go helium release occur thed an t sni decrease s wit e achievinth h e practicallth f go y stationary value at the burn-up level -4* 1021 capt/cm (Fig. 4). For example, the maximum helium release from the natural B C was ~55% at the burn-up -1.5*10 capt/cm and irradiation

158 BURN - UP , at. % BORON

FIG. Helium4. release from boron B-10),% carbideat. irradiation(20 temperature 500- 800 C.

21 3 temperatur e burn-uth et a ~70 p; ~4*10°C 0 capt/c e heliuth m m releas s -20e th wa e f %o produced one. The helium release from the enriched B C under the same conditions was less.

4.2.5. Pellet-cladding interaction

The stainless steel (type 16Cr-15Ni, thickness 0.7 mm) is used as a cladding in the absorber pins. The pellet-cladding interaction leads to the severe embrittlement of cladding materia resula s a f lcarbo o t borod nan n penetration effects evaluato T . e thickneseth e th f o s interaction layer test g date usualle ri th f sar , a o yshowe s pose usedwa th tt y I .dirradiatiob n investigation, that the irradiation decreases the temperature level of the beginning of pellet- cladding interaction by 100-120°C compared with the level obtained under the rig conditions. The intensity of the pellet-cladding interaction rises with the burn-up increase. In the BN-600 absorber rod with moderator the interaction layer between B4C and cladding (3H-847 steel) after irradiation for the 7500 hours was up to 100 mkm for the softened spectrum e temperaturTh . s abouwa e t 430°C e averagth , e burn-u 17*10^— p 1 capt/cm . Durin e investigatiogth BN-60e th f no 0 control rods wit ^IC-6e hth 8 steel claddind gan with operation time of 500 EFPD (12000 hours) in the temperature range of 390-520 °C, no pellet-cladding interactio observeds nwa .

159 5. CONCLUSION

. The BN-600 experimental control rods containing boron carbide were in operation for 500 EFPD (burn-up up to 18.3*1021 capt/cm3 and irradiation dose up to 70 dpa). According to the results of their operation and post irradiation investigations, it could be concluded the following: • There wert detecteno e y changean d f rodo s s construction abl o makt e e worse their performance. There wer o mechanican e l deformations e hingeth ; s retained their original mobility. The maximum increase of rods diameter was 0.2-0.4 mm. • The rods retained the margin of performance; the pellet-cladding gap is retained; there is no pellet-cladding interaction (materia f claddino l g- HC-6 8 steel) e structurath ; l material has sufficient margin of the mechanical characteristics. • At the reached level of absorber burn-up (18.3*10 1 capt/cm3), the obtained earlier characteristic s irradiatioit f so n resistance didn't change evaluato T . e life-timeth f safeto e y e swellinth rod y b s f absorbego e followin th s possibl i e t i us r go t e values lineait f o sr swelling rate: 0.6% per % at. of boron burn-up for the enriched B4C; and 1.0% per % at. of boron burn-up for natural B4C. calculatior Fo absorbee • th f no r temperatur burn-ue th t ea p more than 1.5*1021 capt/cmt 3i is recommended to use the value of thermal conductivity equal to 4.0 W/m°K at rated reactor power. • The helium release at the burn-up of 1.5*10 * capt/cm in the range of absorber operation temperatur s 55%wa e . Wit e raisinhth f burn-ugo e heliupth m releasf o e reduce% 20 o t s produced.

160 IRRADIATION BEHAVIO BOROF RO N CARBIDE NEUTRON ABSORBER

. KAITT O Nuclear Fuel Research Section, Advanced Technology Division

. MARUYAMAT . ONOSES , . HORIUCHH , I Material Monitoring Section, Fuel Materiald san s Division

Oarai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Ibaraki, Japan

Abstract Boron carbide pellets were irradiate o 23Ot 0 cap/ 1 Xp u d m burnup at maximum temperature of 1400°C in "JOYO" MK-II core. Pin puncturing tests, densit d thermaan y l conductivity measurements were performe n theso d e pelletsirradiatioe th d an , n behaviors were evaluated. It is concluded that the high burnup boron carbide pellets take different irradiation behavior fro burnumw lo tha f pto pellets. If the burnup and the irradiation temperature are greater than about 100X102Scap/m d abouan 3 t 1000°C, respectively heliue th , m release is enhanced due to remarkable microcracking of the pellet, which cause swelline sth g rate decrease.- 1.Introduction Boron carbid extensivels ei y controa use s da materiad lro fasr fo lt reactors, because of its superior properties such as great neutron absorption capacity, high melting temperature, light weighd an t chemical stabilit t elevatea y d temperatures. However e heliuth , m produce 10y dB(n,a)b 7Li reactio releases ni d fropellete th m caused ,an s ths pressurega increase th e f insido e e absorber pins e retaineTh . d helium in the pellet causes the pellet swelling, and causes the absorber- cladding mechanical interaction(ACMI). Since the life time of control rode mainlar s y dominate y heliub d m releas d pellean e t swelling s importani t i , o evaluatt t neutroe th e n irradiation behavior of boron carbid higo t hp e u burnup . On the last specialists meeting in 1983[1!, the irradiation behavior of boron carbide pellets in "JOYO" MK-I core was presented, in which burnup leve s limitewa l o onlt d y abou 10X 2630 tcap/m e irradiatio3Th . n behavior of the pellets up to about 80 X 10 cap/m burnup were also 2S 3 reported from each country. Since that meeting, results of

161 post-irradiation examination have been accumulated in Europe up to about 150Xlo26cap/m3[2!. In PNC, the experimental fast reactor "JOYO" was improved to MK-II cor n 1983 i ed irradiatio an , n tes f boroo t n carbide pellet s beeha s n executed using Absorber Materials Irradiatio (AMIRg developmenr nRi fo ) t of long life control rods. Recently e post-irradiatioth , n data were obtained up to burnup level of 230 X 1026cap/m3 at maximum irradiation temperatur f o 1400e . C °Thi s paper describe e resultth s f o s post-irradiation examinatio f boroo n n carbide pellets irradiatet a d such high burnup and high temperature region. 2 .Expérimental procedure 2 . 1 Specimens and irradiation condition Chemical compositions of boron carbide pellets are shown in Tablel . These pellets were pressingt formeho y b dd boro an ,o carbo t n n ratios 4.0+0.2.n i wer t ese Specification borof o s n carbide pellet e showar s n n Tablelli B enrichment.10 e variear s t threa d e levels betweend an 7 3 . 92at d pelle%an t densitie d 95%T.De variean ar s 0 e averag9 dTh . e grain sizes of all pellets are under Tablel chemical Compositio Borof o n n Carbide Pellet (wt%)

Item Specification Analysis

"B Enrichment 37at% 50at% 92at%

Total B e75 76.7 77.8 76.8 Total C S23 22.4 22.4 23.0

B/c Ratio 4.0±0.2 3.85 3.96 3.99

Soluble B <0.2 0.08 0.05 0.05 Soluble C 1.51 1.7 2.0

Fe ^0.8 0.08 0.04 0.02 Ti gO.l 0.006 0.03 <0.01 Si ^0.15 0.20 0.06 0.07 Al Ê0.20 0.09 0.09 0.03 Cl+F <0.01 <0.005 <0.005 <0.005 Co ^0.005 <0. 00025 <0. 00025 . 0002<0 5 Cu eo.oi 0.0009 <0.0003 <0.0003 Mn SO. 01 0.000224 0.0003 0.0002 Na go. 01 0.0002 0.0003 0.0002

These pellets were inserte n stainlesi d s steel capsulesd an , irradiated as AMIR test and "JOYO" MK-I and MK-II driver control rods. Irradiation conditions of boron carbide pellets are also shown in tablell e maximuTh . m burnup reache 10X e maximu7 26th cap/m22 sd man 3 irradiation temperatur pellet a e t cente s 1380°Ci r . Where burnupe th , s d irradiatioan n temperatures were estimated using MAGI d codan e HEATING-5 code, respectively. 162 2.2 Experimental method After completion of irradiation, puncturing tests of the capsules were conducte o estimatt d e heliuth e m quantities released from boron carbide pellets. Then visual inspection, density measuremend an t thermal conductivity measurement were performed on the boron carbide pellets.

Tablell Specification and Irradiation Condition of Boron Carbide Pellet

*i *2 "B Pellet Average Irradiation Burnup Enrichment Density Grain Size Teinperature UO^cap/n1 ) —— (at%) (%T.D. ) (/*m) (°C) AMIR-1 37 95 <5 33-51 580-920

AMIR-2 37 95 <5 92-132 590-880

AMIR-3-1 92 95 <5 115-159 1020-1330

AMIR-3-2 92 95 <5 168-227 1180-1310

AMIR-4-1 50, 92 95 <5 5-100 500-1380

MK-I 92 95 <5 -50 -610

MK-II 92 90 <5 -80 -1330

*1 Calculated by MAGI code *2 Calculated by HATING-5 code(at Pellet Center)

(1)Helium release Helium quantities released from boron carbide pellets were estimated by measurement of helium gas pressure inside the capsules through the puncturing tests of capsules. The released helium quantities are normalize e heliuth n mi d volum t standara e d pressur d temperaturan e e

3 3 for unit volume of boron carbide (m -STP/m -B4C) . (2)Pellet swelling To estimate the swelling of boron carbide pellets, density measurements were conducted by means of picnometric method using mercury. The unit of the swelling was converted from density change to diametral chang eAD/D%{ ). (3)Thermal conductivity The pellets were machined into a plates with 3mm in thickness. Then, the thermal diffusivit s measureywa d usin lasee th g r flash methode Th . thermal conductivity K was evaluated using the following equation with the heat capacity of unirradiated pellet. a • p C • p = K

where, p is pellet density, Cp is heat capacity and a is measured thermal diffusivity.

163 3.Result d discussioan s n (1)Helium release Result measuremenf so f releaseo t d helium quantities froboroe th m n carbide pellet e show ar sFig.ln i n . Released heliu mucs i m h lower than the helium generation at the burnup level below 100 X 1026cap/m3. In

300

o 250 mi He Generate 200 -# wi * t 150

100 Irradiation tV' Temperature

0) K 50

i i 0 50 100 150 200 250 300 Burnup (1026cap/m3)

Fig.l Burnup Dependence of Helium Release

further irradiation, the helium release is accelerated, and release rate becomes almost equivalent to the generation rate under the condition of over 150X 1026cap/m3 and over 1000°C. This result is almost e estimate th e sam th s ea d result USAn si 131. However reportes i t i , y db Europe121 thaheliue th t m release rate becomes almost equivalene th o t generatio 10nX 2650 ratcap/m t a e 3, whic s lowei h r burnup level thar ou n result. Irradiation temperature dependence of fractional helium release is shown in Fig.2, which makes clear that helium release increases linearly with temperature within burnup range of 150X 1026cap/m3. When burnup exceeds 150Xl026cap/m3, influenc burnuf eo p remarkable b ten o t d e rather than temperature dependence. FigureS shows the apparent view of irradiated boron carbide pellets. Boron carbide pellet brokee ar s n into small pieces by neutron irradiation. The higher the temperature and the burnup e smalle,th brokee th r n pieces became s showA . n Fig.4i n , there are many microcracks inside the grain, some of them reached to the

164 Irradiation Temperature

Fractional Helium Release (%) H- IQ • H-> O C U rfitO . W 1^0 o o o o c o i —————— ————————————————————————————— "9 o 1 1 1 1 H- oo p k> D O ^ ^••D>Oc?;ro Fra c t : Irrad : -J o 1 H o S-i cnouioo^d n "3 o o o o H V'TJPv ff. t*. Hf o P> HI O 0* & g {î. H- °og H P> rr vo H- o ïu o o 0» > I H A 3 SH- l f** H i

^•atur e D Rele a (D D H M u 1 pi H c^n o rt O 3 1 O H- C VO m o Vy p*tL-f S3 00 CO 00 Hi v v *>o XX (D r-, D • O X (->(-> D «t. A H" O O O w w 8 *P V . . (T . <7\ g 0° N / X XW H- (h U VJ W PO> "ßd) «P) o s- "O "v. "x, ! o 0° * ^-33 (D 0 O HI H O

Fig.4 Hicrophotographs of Irradiated Boron Carbide Pellets (SEM image)

grain boundary. The grain boundary of boron carbide irradiated up to high burnup appears to be brittle, as shown in right hand side of Fig.4. From the results of examination described above, dominant process of helium release seems to be helium diffusion in crystal grain . At higher burnup region, helium release is accelerated by 11 4 51 increasing of helium release paths due to occurrence of microcracking in helium accumulated pellets. Pellet density, grain size and 10B enrichment are also the fabrication parameters with an effect on helium release. However, from the microstructure as shown in Fig. 3 and 4, the effects of those parameter e considerear s negligible b o t d higt a e h burnup region where the pellet cracking is a dominant factor for helium release. (2)Pellet swelling Figures show e burnuth s p dependenc f diametrao e l swellin f boroo g n carbide pellets. Swelling of boron carbide pellets tends to increase linearly with e swellinburnuth t a pg rat f abouo e A D/D%/100 3 t X 10w burnup26lo cap/m t a .3 When burnup exceed X 100 2610 cap/ms e th 3, swelling rate tends to decrease gradually. This behavior corresponds to e heliuth e considereb m n releasca t Fig.ln i i de d thaan , t increasing of released helium quantities contributes to decreasing of swelling rate e swellinTh . g rate reporte n eaci d h country 112s aboui 1 t 5AD/D%

166 dp 4 Q

-HH 13

4J 0> Irradiation CM Tempera ture

0 20 0 15 50 0 10 250 300 Burnup (102Scap/m3 )

Fig.5 Burnup Dependence of Pellet Swelling

/100X 1026cap/m3 ,e greateth whic s i hr value thar resultsou n e Th . decreasing behavio f swellino r g rate beyond 100X 10 cap/m s alsha o 26 3 never reporte e irradiatee dstudth th even o yn i n d pellets reached about 150X1026cap/m Europen i 3 . Relationship between swelling and helium retention in the boron carbide pellets was estimated, because the primary factor of swelling is considered to be helium accumulation in the pellets111. As shown in Fig.6, helium retention increases linearly with burnup up to about 100X 102Scap/m3, and beyond that burnup helium retention tends to saturate. Figure? show e relationshith s p between helium retentio d swellingan n . Although the data are relatively scattered, swelling and helium retention has proportional relation. This result supports that the main factor of swelling is accumulation of helium produced by 10B(n, a )7Li reaction in the boron carbide pellets. According to the report by Zuppiroli et al[61, temperature does not influence on swelling below 1500°C. However, since the swelling has the proportional correlation to helium retention which depends on temperature, it is considered that the swelling has a temperature dependence even below 1500°C. (3)Thermal conductivity Figures shows temperature dependence of thermal conductivity of irradiated boron carbide. The thermal conductivity of irradiated

167 800

o mi "B 600 CM

g 400 -H JJ Ö 01 4-1 OJ Irradiation 200 Temperature

0) K

0 50 100 150 200 250 300 Burnup (1026cap/m3 )

Fig.6 Burnup Dependence of Helium Retention

5.0

4.0

Q ^ Q

3.0

•H rH rH Burnup,Temp. l 2.0 O -50,^1000 4J

0) D -100,>1000 04 1.0 • ~150,>1000 A ~200,>1000 • ~250,>1000

0.0 0 80 0 60 0 40 200 1000 Helium Retention (m3 -STP/m3 -B^C)

Fig.7 Relationship between Helium Retention and Pellet Swelling

168 specimen t dependeno s n temperatureha so d , while thermal conductivity of unirradiated specimen tends to decrease with temperature, that is a typical phenomenon of phonon conduction. These results agree with those reported by Gilchrist171 and Mahagin et al181. The phonon mean free path (mfp) 1 of irradiated materials can be expresses da 111

n unirradiatei p mf e p associateth where mf ds i e material0 ,th 1 d l d d san

with irradiation induced defects. If the mfp ld is very small compared

with 10, I is effectively equal to ld and the thermal conductivity is determine e irradiatioth y db n induced defects alon eBecaus. vers i y1 e

small compared with 10, it is considered that thermal conductivity of irradiated boron carbide has less temperature dependence such as unirradiated specimens .

30 Burnup (1026cap/m3

•vg, g 20 Unirradiated

15 U

Io o 10 gfC Ï-I fi0)

0 200 400 600 800 1000 Temperature (°C)

Fig 8 .Temperatur e Dependenc f Thermao e l Conductivity

Burnup dependenc f thermao e l conductivit f boroo y n carbid t rooa e m temperatur s showi e n Fig.9i n . Thermal conductivit f unirradiateo y d boron carbide correspond o about s t SOWm^K"1. Thermal conductivity decreases rapidl y irradiatiob y e thir on f unirradiate o do t n d value till abou 10X 5 t2S cap/m f burnupo 3 . The t decreasei n s gradually with

169 300

MK-O I (>95%T.D.) MK-IA I (90%T.D.} • AMIR (90%T.D.) g \ S

•rH 4J u 3 •o o u

g(0 1-1 (D

30 40 50 Burnup (1026cap/m) 3

Fig.9 Burnup Dependence of Thermal Conductivity (Measurement temperature:25°C)

burnu d reacheunirradiatef po an % s10 d valu about ea t 50Xl026cap/me 3.Th tendency that thermal conductivity decrease still remain about sa X 50 t 1026cap/m3.

4.Conclusion Irradiation effectpropertiee th n o s borof o s n carbide pellets were evaluated with the post-irradiation data, which reaches burnup of 230X 10 cap/m at maximum temperature of 1400°C. The results in this study 26 3 are summarized as follows. (1)Boron carbide pellet e crackear s n piecei d s with neutron irradiation. Particularly, boron carbide pellets are cracked remarkably over the burnup of 100X1026cap/m3 and over about 1000°C. (2)Helium release shows rapid increase with burnup over about 10OX 1026cap/m3, and is enhanced by the occurrence of pellet cracking. (3)The swelling rate of boron carbide pellets takes about 3% per 100 10X 26cap/m t burnua 3 f beloo p w 100X 1026cap/m d beloan 3 w e 1000Th . °C swelling rate decrease burnut sa beyonf po d 100X1026cap/m3, which might be due to the accelerated helium release by microcracking. (4)Thermal conductivit f boroo y n carbide pellets shows great degradation with neutron irradiation. The thermal conductivity of irradiated boron carbide pellet littls sha e temperature dependence.

170 It is concluded that the properties of boron carbide pellets irradiate t higa d h burnup leve e verar ly different frow m lo tha f o t burnu pirradiatioe burnue levelth th d f an pI . n temperatur e greateear r than about 100X 1026cap/m d abouan 3 t 1000°C, respectively e heliuth , m releas enhances ei remarkablo t e du d e pellet cracking e swellinth d an ,g rate decreases. These dat n higo a h burnup boron carbide pellets suc s heliua h m releas d pelleean t swellin e verar gy valuabl develoo t e e lonth pg life control rods. ACKNOWLEDGMENT e authorTh s express sincere appreciatio e memberth o f Fuelt no s s Monitoring Section(FMSo executewh C e puncturinPN th d n i ) g testo t s estimat heliue th e m release behavio borof ro n carbide pellets.

REFERENCES

[1]IAEA-IWGFR Specialists Meeting on Absorber Materials and Control Rods for Fast Reactors, Obninsk, Russia, (1983) [2]B.T.Kelly, B.Kryger, J.E.Eicline, P.Holler; FR'91 International Conference on Fast Reactors and Related Fuel Cycles, Proceedings vol.Ill PI.10, (1991) [3]J.A.Basmajian, A.L.Pitner; HEDL-SA-1189-FP, (1977) [4]C.E.Beyer, G.W.Hollenberg, S.A.Duran; HEDL-SA-1883-FP, (1979) [5]G.W.Hollenberg; HEDL-SA-2236-FP, (1981) [6]L.Zuppiroli, D.Lesueur; Philosophical Magazin evol.6A 0 No.5 539-551 (1989) [7]K.E.Gilchrist; High Temperature High Pressure vol.17 671-681, (1985) [8]D.E.Mahagin, J.L.Bates, D.E.Baker; HEDL-SA-578, (1973)

171 THIRD SHUTDOWN LEVEL FOR EFR PROJECT

. FAVETD . CARLUEB , C FRAMATOME Direction NOVATOME Lyon

S. DECHELETTE Electricité de France, Cedex

France

Abstract presentatioe Th n describe evaluated dan thirda d shutdowR EF n e leveth f o l project, which has been developed implementing the following four new device systems: systea m whic• h terminate powee absorbee sth th o t r r electromagnets afte losa r f so primary pump electric power supply; devica contror efo • l rods drive lines, which passively enhances thermal expansion; devica e whic• h overcomes contro jammingd ro l motorizey b , d insertio f absorbeno r rodsd an , • a mechanical stroke-limitation device, which passively terminates the withdrawal of faultee th d control rod.

presentatio e goale th f Th followine so th e nar : g

. to present the EFR safety approach which leads to improve the preventative measures agains occurrence th t f wholeo e core accident basie judginr th sfo d incentivee an , gth s f implementatioo f additionano l shutdown devices (so-called third shutdown level), . to present the chosen devices and their preliminary design, preseno t . efficience th tthire th df o yshutdow n leve casn i l f transienteo t protecteno s d basie byth c shutdown systems.

o independenTw t shutdown system e providear s EFRn di e principle desig.e Th th d nan s options concernin basie gth c shutdown system achievo t s e reliabiliteth y requiremene ar t presented.

rise kTh minimisatio striv o comprehensivt a nr s i approacefo R EF r hbalancefo d ean d safety concept by harnessing favourable features of liquid metal fast breeder reactors. For the shutdown function a so-called third shutdown level has been introduced as an extra safeguard. This basically consist f additionaso l engineering features which include passiv activd ean e measures capabl bringinf eo reactoe g th saf a o ert conditio casn i postulatef eo d total failure of the two basic shutdown systems after design basis transients.

The devices are designed in order to achieve the two principle third shutdown functions :

disengago t . absorbee eth r rod thao s t falthey l yma intcore e oth , mechanicallo t . y assis insertioe absorbee th t th f no r rods.

173 The main, third shutdown devices which will be described in the presentation are the following : . SADE system which passively terminate e poweth s r e absorbesupplth o t y r electromagnets after a loss of primary pump electrical power supply, . delatching by control rod enhanced expansion device (CREED) which provides an increased thermal expansion on the rod driveline. At a certain threshold of expansion a delatching mechanism initiates passively rod release, . bulk rod insertion (BRI) which provides shutdown by motorised insertion of absorber rods casn I .jamminf eo rodse I provideth BR f , go s drive-in forces whic e mucar h h higher than gravit onld yan y limitestrengte th drive y th d b f eho mechanism, . rod disconnection initiated by the mechanical stroke limitation device which passively terminate withdrawae sth faultee th f o dl rods. e thirTh d shutdown leve assesses i l regarn di postulateo dt d transients. r eacFo h transient t wili ,verifie e b l d:

. that the third shutdown level provides an additional, efficient and reliable line of defence, . the fulfilment of the success criteria by the third shutdown level.

1. INTRODUCTION overale Th l safet reduco t y s risi e objectivf operatioes th ko R a EF w levea r lo o efo ns t a l reasonably practical (ALARP principle) and no higher than that expected for future LWR'S Europeae inth n Countries concerned. Apart from exotic initiators (e.g. fast structural collapse or insertion of a large amount of gas into the core) a core melt accident in a LMFBR with state of the art safety features and engineering could only occu f eithei r e shutdowth r n system e decath r yo s heat removal systems fail totally. In general, postulated failure of these systems in past LMFBR'S resulted into core melt scenarios. Therefore, an improvement of these safety functions - i.e. shutdown and decay heat removal - would result in a genuine enhancement of preventive safety. The improvemen shutdowe th f o t n functio(referencR EF presentes f i o n ) followine 1 e th n di g sections.

. 2 STRATEG RISR YKFO MINIMIZATIOR EF R NFO For the shutdown function a so-called third shutdown level has been introduced in EFR as an extra safeguard. This basically consists of additional engineering features which include passiv activd ean e measures capabl f bringineo reactoe gth safa o et r conditio casn ni f eo postulated total failure of the two basis shutdown systems after design basis transients. In orde ensuro t r erobusa weld an tl balanced preventative safety concept extenden a , d line of defence concept (referenc applieds i ) e2 . Basically mediue strono on tw d , mgan linef so defenc requiree ear orden di relegato t r intolerable eth e natural consequence accidenn a f so t sequence int residuae oth l riskstrono Tw . g line defencf so genera n i e ear l contributee th y db reactor protection systems. Other such features as plant protection system, natural behaviour and operator action may represent medium lines of defence. The low frequency of initiating events may also be equivalent to a line of defence. In the frame of the EFR risk minimization effort, this basis target has been extended : at least, one additional line of defence should be introduced to interrupt event sequences which would have the potential to cause serious core damage. The extra safeguards against failure

174 of shutdown have been introduced for the more frequent initiating faults. If a low frequency faul diverso protectes i ttw e eth shutdowy db n systems overale th , l protection agains whola t e core acciden alreads i t y better than basically required.

Exotic initiators (e.g. fast structural collapse or insertion of a large amount of gas into the core) are already excluded by appropriate design measures. In these cases, a deterministic approach resulting from robust design feature attenuatind an s g natural behaviour justifieo t s relegate these events into the residual risk.

Table I presents the lines of defence basically required to relagate whole core melt scenarios intresiduae oth typicae lth risr kfo l events additionae Th . l line f defenco s r furtheefo r risk minimizatio describee nar sectio e listed th an Tabln n di i n3 . eII

TABL EI BASIC LINE DEFENCF SO REPRESENTATIVR EFO E INITIATORS

INITIATOR LOD'Require) b + Sa d(2 Loss ot station service power RPa 2 S: SADE • a (LOSSP) Coast down of all primary PPS b RPS 2a pumps (LOF) by other cause than LOSSP

Break of one LIPOSO (Q - 42 Prevention . a RPS 2a %) Simultaneous breaf ko two LIPOSO's same th et a ipum)~ Q p( Prevention 2a RPS 2a 26 %) differen«it n) t pump— s(Q Prevention 2a + b RPS 2a 24 %) Los t secondarso y pumps (LOHS) Loss of feedwater or other PPS b RPS 2a malfunctiosysteS W/ me th f no in all loops (LOHS)

Withdrawa singla f o d l ero runawad Ro y protectiob n RPS 2a

Withdrawa l rodal sf o l Global powei limitation b RPS 2a Total instantaneous, blockage Prevention 2a RPS b Core compaction by SSE Prevention 2a Detectio seismograpy nb a h Design with benign response of core and components • b Core support failure Protection equivalent to (2a + b) LOD'S by design standards, minimal ris f bucklingko , in-service inspection, leak before break argument, self redundant structure (e.g. secondary skirt), slow development detectablS RP y eb Voidin fissioy gb s releasnga e (severe earthquake or BOB earthquake, severe core Preventio f avoidancno initiaton a f e o r failur fo rlarga f eo e number transients (ULOF, ULOHS, f simultaneouo s cladb - failures4 a 2 , UTOP) with prolonged high temperature conditions)

RPS . Reactor Protection System : PlanPPSt Protection System BDB : Beyond Design Basis (e.g Limiting Event) BDB/RR Beyond Design Basi s/ Residua l Risk SSE . Safe Shutdown Earthquake ULOF Unprotected Loss of Flow ULOHS Unprotected Loss of Heat Sink UTOP Unprotected Transient Over Power LIPOSO piping connecting the primary pump at the diagrid D LO strona g Lin Defencf eO e : mediu D mLO Linb Defencf eO e

175 TABLE II ADDITIONAL LINES OF DEFENCE FOR RISK MINIMIZATION

Initiator Additional LOD'S for Risk Minimisation ULOSSP SADE. BRI, CREED ULOF other than ULOSSP CREEDI BR . ULOHS CREEDI BR . Withdrawa singla f d o l ero SJLD wit released hro D DN , Withdrawal of all rods SLD with rod release SLD + CREED for part load Subassembly accident propagation DN r ABNDo D dependin initiatoe e th th n d go ran sequence of events

Passage of gas through the core Reactimeter Radial core movement

Core support failure Reactimeter, disconnection of rods after limited core drop

ULOSS : PUnprotecte d Los Statiof so n Service Power

ULO : UnprotecteF d Los Flof so w

ULOHS : Unprotected Loss of Heat Sink

: InsertioBuld BRkRo I n

: SLStrokD e Limitation Device

: DNDelayeD d Neutron Detection

ABND : Acoustic Boiling Noise Detection

The fulfilment of the success criteria by the third shutdown level is verified for postulated failures of the main shutdown systems (total failure of the reactor trip systems or of the absorber systems) after serious imbalances of produced and removed power (failure of the pumps, faulty withdrawa f absorberso l , los f maio s n heat sink) which could lea coro dt e disruptive accidents (ULOF, UTOP, ULOHS), result f transieno s t analysi e givear s n i n section 4.

176 . 3 DESCRIPTIO DEVICEF NO FEATURED SAN IMPROVEMENR SFO E TH F TO SHUTDOWN FUNCTION

The third shutdown level consists of passive and active measures being capable of bringing e reacto th a saf o et r conditio a postulatecasn i n f o e basio dtw ce failur shutdowth f o e n systems differeno Tw . t type f failurbasie so th cf eo shutdow n functio consideree ar n d:

• failure to de-energise the scram magnets,

• failure of rods to drop into the core.

Failure to de-energise all electromagnets is minimised by redundancy and diversity of the trip systems. Because of the diversity of the electromagnet locations (in the cover gas and under sodium), only rod jamming in the core remains as a cause of mechanical rod failure. Regardin hige gth h degre extreme f diversiteo th rode d th san ef yo misalignemen tubd an t e deflections which can be tolerated, this type of failure is judged to be extremely unfrequent.

Regarding these type f failuresprinciplo so tw e th , e function thire th df so shutdow n : leve e ar l

• to disengage the absorber rods so that they may fall into the core,

mechanicallo t • y assis insertioe absorbee th t th f no r rods.

The main features of the third shutdown level are presented in the following

3.1 SADE System

SADe Th E system would passively terminat powee eth r typsupple f absorbeon eo f yo d ro r electromagnet afte losa r f primarso y pump electrical power supply trie th p f i signal, d sha faile initiato dt drod ero p (see figur. 1) e

These electromagnets are electrically fed by a generator which is driven by a motor provided wit hflywheela desige flywheee th Th . f no s suci l h thagravite th t yrode droth sf po occur s lesn i s than abou seconds0 1 t . This valu consistens ei t wit halvine hth gprimare timth f eo y pump coast down and is small enough to avoid sodium boiling in case of loss of station service power combined with a failure of both shutdown systems.

2 Delatchin3. Controv gb Enhanced Ro l d Expansion Device (CREED)

CREED is a passive mechanism for the delatching of rods into the core in response to core outlet temperature increase (Figure2>). It essentially consists of an element which provides an increased thermal expansion on the rod driveline. At a certain threshold of expansion, a delatching mechanism passively initiates rod release.

CREED characteristics are : threshole th C ° delatchinr 0 dfo • 60 - C abouf ° go 0 50 t • a time delay of the enhanced expansion related to the core outlet temperature of about 14 s at nominal flow.

177 From From main grid standby grid oo

27kV~ L ^busbarV Tk o6 s

To primary pumps secondary pumps feed water pumps ~ V 0 38 380 V~

Fly wheel Fly wheel

Generato) = r f ~ V~~j I——( = J Generator

Motor Motor 2= 4V 24 V=

I d12 d12 _ _ From ^:^ From 6-contact /d31 fxHd32 d 2 — reactor 6-contact /d31 d32 d 2 reactor system 1 system2 I r n trip system1 trip system2 d21 d22 d3 ~^ d21 /d22

rodD s5DS 4 OSD rods

FIGUR E1 PRINCIPL SADF EO E Sodium I eve I

Martens i t ic sta s steelies n I

Sod i um f Iow through the expansion jacket

Expansion jacket Austenitic stai n I ess steeI ; Protective incone ackej I t

nconeI

E !ectromagnet

d DShearo D d

FIGURMAGNET HO D E2 T RO CREE PUS D DEVICF HDS OF D - E

179 3.3 Rod Disconnection Initiated by the Mechanical Stroke Limitation Device

The primary purpose of the Stroke Limitation Device (SLD) (Figure 3) is to terminate the withdrawa faultee th f o ld rod casn i s f inadverteneo withdrawad ro t reduco s d rise ean th lk of unacceptable fuel melting to an acceptably low level. Conditions in the affected fuel pins may further deteriorate after terminatiowithdrawald ro e th hasD f o nSL ,. thereforee th , additional functio releasd ro f o ne shoul stroke dth e limi reachede b t .

4 BulInsertiod 3. kRo n (BRI)

In order to prevent rod jamming, the shutdown function by gravitational drop of absorbers complementes i motorisey db d insertio f absorbeno r rods initiate trie th p y signalsdb .

In case of jamming of the rods, BRI provides drive-in forces that are much higher than gravit onld yan y limitestrengte drive th th y db f e ho mechanisms.

. 4 ASSESSMEN RISE KTH MINIMIZATIOF TO ADDITIONAY NB L PREVENTION

4.1 Success criteria

The following success criteria are used to assess the efficiency of the available elements and features :

A/ Short term transient phase

. Avoidanc f widespreaeo d sodium boiling, . limitin fuee gth l melt fractio overpowen i n r transients, . core support structure integrity must not be jeopardized, . maintenance of coolable global geometry.

B/ Long term

. Safe neutronic shutdown at an asymptotic equilibrium temperature of 500°C, . Decay heat removal systems must reac maintaid han safe nth e neutronic shutdown equilibrium temperature.

4.2 Efficiency of the third shutdown level

threr Fo e typical faulty transient Unprotectes- d Los Flof so w (ULOF), Unprotected Transient Overpower (UTOP) and Unprotected Loss of Heat Sink (ULOHS), the effects of CREED, I havBR e d beean n D parametricallSL y studied. Both CREED effects enhanced thermal expansion and delatching, have been separately assessed.

4.2.1 Unprotected Loss of Flow (ULOF)

naturae Th l behaviou untrippen a r fo r d los f floso w rapidly lead larga o st e increase th f eo coolant temperatur e shorth tn i eterm , wherea e reactoth s r powere th decrease o t e du s

180 Stroke 1 imiter Main motor motor with e v i r d pos11 i on encoder

Switenable coug in pI C stroke I imiter shaft Ana logue to ei ther shaft stroke I imiter motor or ma. i n shaft 3

n driv i a M e Stroke l i m i ter shaft shaft

Upper analogue nut Stroke I imiter stop nut

Lower Stroke l i m i ter analogut nu e limit switch

4 1 rat i o är ive between main drive shaft and analogue shaft

Th e D CPBsketc I undeSX e s SKETCH3DWG_G4i VA archivehth r e th n do O SK 5

FIGURE 3 SLD - MOVEABLE STOP TYPE SLD DESIGN

181 reactivity feedback effects f sodiuI . m sink boilin avoideds gi furthee th , r developmene th f o t acciden decas e linkei t th o yt d heat removal performance, which depend accidene th n o s t scenario. Computer simulation have shown thaULOe th t F initiate LOSSa y db P e leadth o st highest core temperatures.

In this case transiene th , t analysi corsR performe EF eincludin- e th r fo dga contro l rods drivelines enhanced thermal expansion effec f abouo t4 mm/°0. t C wit a hspecifi c flow dependent tim7 (primare+2 constan1 1 y = flo t t w rate) second - sreveale d that sodium boilinavoidee b n Figurg d ca d- Tabl an I . These4 eII e result assumptio e basee th sar n do n that there is no delatching, and all the rods are equipped with CREED.

Improvement of risk minimizations is obtained if the absorber rods are delatched by CREED.

The long-term effect of both CREED functions enhanced thermal expansion and delatching on the asymptotic equilibrium temperature (AET) can be seen in Table IV. The improvement of these passive shutdown features is sufficient to keep the AET below 700°C and to avoid structural failures for about ten days at least, which provide grace time for operator action, manual shutdown for instance.

Assuming thacontroe th t drivelind ro l e enhanced thermal expansio s weani CREEd kan D delatching is not efficient, it was derived that boiling for ULOF can be avoided by the motor l diversal f shutdowd o drived ean an mm/rodn1 D sninsertio t s a CS rod l al s f (DSDno )(BRI) at 8 mm/s.

TABLE III. SUMMAR TRANSIENF YO T RESULTS All cases except (1) include CREED model

Case Pump halving Minimum% Minimum rod Maximum Time of Margio nt time (s) flow insertion % sodium maximum boil temperature °C temperature (s)

1 10 0 0 boiling 150 -

2 10 0 0 923 88 29

3 10 0 5 886 88 67

4 10 25 0 873 30 85

5 10 10 0 918 76 36

6 20 0 0 884 230 69

4.2.2 Unprotected Transient Overpower (UTQP)

Only faults, leading to an overpower, that can be controlled by the reactor shutdown systems consideree ar the d callee dan yar d slow transient overpower.

Such faults are caused by the unscheduled withdrawal of one or several control rods, occuring at full power or part loads. They provoke slow increases in power either widespread in the cor r restrictiveleo y locate vicinite th faultea n di f yo d control rod.

182 CREED PARAMETER Ste• S . expansio pm insertiom 0 n10 ratn- mm/de5 e2 =0 gC thermal time constant * (1 - 2/w s (w = normalised primary flow)

CSD rod position = 0 % . pump halving lime = 10 s . no pony motors present 1100 ———————————— i————— — l————— — l————— — i————— — ;

1000 Sodium boiling temperature I

Q; Q 900 0) 5- 800 p(aU. 700

2nd line trip 600 C m 1st line trip

400 0 27 0 24 0 21 0 18 90 0 15 120 300 Tim) (s e

FIGURE 4 MAXIMUM SODIUM TEMPERATURE DURING LOSS OF PRIMARY FLOW TRANSIEN VARIOUR TFO S LEVEL REACTOF SO R SHUTDOWN TABLE IV EFFECT OF CREED ON ASYMPTOTIC EQUILIBRIUM

AET [°C] ASSUMPTIONS = T OLANT) CO BOEC EOEC 930 800 O No CREED, no jamming of rods

O CREED with enhanced expansion abovT A p/ e 600°C:

delatching^o n , rods free movable a) Ap / AT = 0.4 mm/°C 674 694

0 mm/°1. = CT A / p A b) 633 651

~jc delatchiny insertiob p A f failind no g an w f onlgo fe ya CSD rods, other rods not delatched but free movable 616 555 Ap = -1 $ 547 446 A-1.= p $ 5 437 346 $ 2 - A= p

* delatching of all CSD rods at T = 625 °C during the transient, thermay insertiob p A lf no expansio falliny b a d f go nan few rods 500 500 Ap = -1$ 450 450 A-1.2= p $ 5 * rod delatchins CSD fallinall gand gof Cold shutdown

BOE C: Beginnin equilibriuf go m cycle

f equilibriuo EOEd En C: m cycle

e completTh e withdrawa e singl on frod f ero mo l half inserted initial position n (stara f o t equilibrium cycle) with automatic core global power compensation would cause a 25 % local overpower for an inner ring control and shutdown rod (CSD) and 60 % to 70 % for an outer ring CSD - best estimate analyses because UTOP is beyond the design basis - showed that no fuel melting would take place in case of the inner ring CSD complete withdrawal whereas t certainli y oute e woulcasn th i difficuls i f r eo o t ringd i s demonstrato A t .t e tha casn i t e of fuel melting clad failure allowing some interaction between the coolant and molten fuel will not occur, it is conceivable that the fault could propagate into a core melting.

184 The simultaneous withdrawal of all CSD rods from full power operating conditions at the beginnin f cyclgo e wit speeh a mm/ 1 f do s would s reactivitcausil 3 ea y ramp complete Th . e withdrawal would insert 8-9 $ (best estimate). The overall reactor power would double with i insertion 0 8 a , which correspond totae th l o reactivitst y insertio evene th f nti occure th t sa en f cycledo .

It is clear from these examples that a UTOP caused by the simultaneous withdrawal of all CSD could initiate an intolerable CD A (Core Disruptive Accident).

In cas f failurdesige o th f eno basis line f defenceso thire th , d shutdown level protectios ni provided by independent systems.

e strokTh e limitation device (SLD s )specifiei o mee t ddemande th t n unprotectea f o s d withdrawarods D designes i t CS I . l al sto o drodf t e o lp th s motio , whicn aftemm h5 allow1 r s a reactivit overpowen Figur. a d % y 0 an insertioshowe 5 5 i - r5 0 4 pea s4 f f nthao k o t peak fuel temperature are near the melting point at this stage of the transient history.

I nwithdrawae casth f single o on f eo l rod t woul,i d limi overpowee th t . % onl o t r3 y2-

Bulk rod insertion (BRI) is unlikely to be effective if the absorber rods system has failed to respon previouo dt s commands fro reactoe mth r contro protectiod an l n system terminato st e withdrawale th .

CREED, with same characteristics as in the ULOF analysis, was not introduced to cause rod insertio withdrawad ro r nfo l faults becaus t befor t woulei ac reactivitt e eth no d y insertion reaches about 1 $, the overpower peak being then 250 %. Although CREED would prevent coolant boiling quite extensive fuel melting would occur.

U o O O O Typical for max. rated fuel a0,

0 6 0 4 0 2 80 Added Reactivity (cents)

FIGURE 5 UTOP - ALL CSD RODS WITHDRAWN

185 4.2.3 Unprotected Loss of Heat Sink

The loss of the main heat sink (LOHS) may be initiated by :

• secondary pump coast dow seizurr no e • loss of feed water.

LOHS can also be initiated by LOSSP but in this case transient behaviour is ruled by the loss f primaro y flow (see section 4.2.1 ULOF).

Assuming that the reactor trip is not initiated or fails, the thermal radial core expansion provide largsa e negative reactivity feedback effect, which lead passiva o st e power reduction. In the long term, all temperatures in the hot and cold pools slowly increase according to the large thermal capacity of the primary system (see Figure 6). Equilibrium is reached if the direct reactor cooling syste mavailabls i removae th r efo f deca o l y power.

CREEd an I e independenDar BR verd an ty effective measure thire th df o sshutdow n level.

f triI p signal r rodso s magnets de-energizatio rodD n sfailDS bule d th ,k an insertioD CS f no enable exerting insertion force overrido st e even limite jammingd dro . Assuming tha l rodal t s insertable ar wit d insertioe ean hth n spee defines da ULOf do F (see section 4.2.1)r thafo s i t f equilibriuo d en e th m cycle insertes i $ , -1,2 - d o 5withit firse nth t minute afte faule th r t initiation a core disruption is definitely excluded for ULOHS. ULOHS is indeed much more benign and developes more slowly than ULOF.

Assuming thagravite th t y fal f absorbeo l r rods fail excludint bu s g irrevocable jamminl al f go CSD and DSD rods, the asymptotic equilibrium temperature (AET) is about 500°C for BRI insertion of -1 $, which is achievable with the insertion of 2 or 3 rods, if no driveline expansion effect from the other rods is assumed.

beneficiae Th l effect f CREEo s consideres i illustratee DI ar BR o tabln dN i . heree4 e Th . smalles i T rAE than 700°C wil T smallee b l AE . r than 500° Cmorf i e than about r -1,fo 3$ end of equilibrium cycle and -1,7 $ for beginning of equilibrium cycle are inserted by the delatching of only a few rods.

5. CONCLUSIONS

The implementation of new devices into the EFR project leads to a significant improvement shutdowe oth f n functio: n

. SADE syste mimplementes i d becaus simplicitys eit , attractive passive featurew lo d an s s higit cost hd reliabilitan s n casi y f LOSSo e P combined with failur f boto e h trip systems.

. Bulk rod insertion with forced drive-in allows to shut down the reactor even in case of mechanical failure with resistance against insertion of aborber rods by gravity.

. Mechanical stroke limitation device limits the maximum reactivity insertion. The repositionning of the stop device is necessary for compensation of burnup reactivity.

186 REACTIVITY (CENT) TEMPERATURE (C) POWE) R(% -800 -4 . 0 80 . 0 60 400 1000 0

H2 m

ULOH- FIGURR EF S WITHOU E6 T CREED Control rod enhanced expansion device could prevent sodium boiling in case of ULOF and ULOHS, even if the delatching function is not assessed. However, the thermal enhanced expansion function of CREED is not required because delatching function and BRI are efficient for all possible shutdown failures.

REFERENCES

[1] C.H. MITCHELL, M.G. PELLOUX, "EFR Shutdown System", Proceedings of an international Topical Meeting, Obninsk, Russia, October 3-7,1994

[2] M. LAVERIE, "Réglementation, options et critères de sûreté", Proceeding of the LMFBR Safety Topical Meeting, Lyon, France, July 1992.

188 IMPROVEMENT OF THE PERFORMANCES OF THE

"POISONED" CAPRA CORE BY USE OF "B4C

. GASTALDG O FRAMATOME Direction NOVATOME Lyon

G. VAMBENEPE Electricit France éd e SEPTEM, Cedex

J.C. GARNIER Commissaria l'Energià t / e Atomique, Durance

France

Abstract approacheo Tw s have been considere Francn CAPRe di th y eb A tea optimismo t efasa t reacto plutoniur rfo r minom(o r actinide) burnin "dilutione th g: "poisoningd "an " options III. At present, "dilution" is the preferred design as it allows better core safety parameters (mainly the Doppler constant).

This paper shows a route, based on the use of a moderator material, "B4C, to improve the core safety parameters (Doppler constant and sodium void reactivity worth) of the CAPRA B4C poisoned core. The aim is to obtain a convincing alternative to the "dilution" option.

The introduction of the moderator inside the fuel bundle of the CAPRA B4C poisoned core allows to strongly increase the Doppler constant (+ 62 %) and to reduce the sodium void reactivit. y%) wort3 2 - h(

The "poisoned" CAPRA core with B4C has similar plutonium burning performances and better core safety parameters tha "dilutione nth U " option mais It . n advantage linkee e sar th o dt larger diameter fuel pin : sincrease d fuel lifetim "standardd an e " fuel bundle (397 pins instead of 469).

CAPRE TH A PROJECT

According to the French cycle strategy based on irradiated fuels reprocessing, the CEA launched the CAPRA project to demonstrate the feasibility of FRs optimised to burn plutonium. n 1993i e CAPR A Starteth , CE y Adb project involves also EuropeaD R& n organizations (AEA and KfK) and design companies (European Fast Reactor Associates consortium). Other close cooperation relatee programmD th n R& do s e have also been established with JAPAN (PNC), RUSSIA (Obninsk), SWITZERLAND (PSI) and ITALY (ENEA).

CAPRe Th A project mus consideree b t frence th n dhi context wher foreseee eth n plutonium yeare stocth sn ki 202 0205- importants 0i fasa , t reactorSo . , plutonium burner meaa s i , n allowin regulato gt amoune eth f plutoniuo t m produce LWRsy db .

189 BACKGROUND

Unlike natural boron (19.82 % 10B isotopic abundance) which is a neutron absorber material, B C is a moderator as B only participates to the neutron slowing down process (scattering

n 4 U without absorption). Moreover, witBhCU pins uniformly distributed in the fuel bundle,

no significant local spectral effect4 is produced so that any impact on the power distributio avoideds ni .

MAIN DATA

Capro maie tw Th nae datth core f ao s use r thidfo s stud collectee yar d hereafter. Thee yar related to the so called "09/93" preliminary core versions.

"Dilution" B4C poisoned Core thermal power (MWth) 3600 3600 Maximum burn-up (GWd/toxide) 170 170

Numbe f fissilo r e sub-assemblies 367 367 Number of diluents (steel1 5 ) (natura1 5 l B4Q Number of absorber rods 33 33 Contro Shutdowd an l n (CSD) rods 24 24 Diverse Shutdown (DSD) rods 9 9 Numbe fuee f pinth o rl n bundlsi e 469 397 Number of fissile pins 320 320 Number of inert pins 149 77 Sub-assembly pitch (mm) 181 181 Fissile height (cm) 100 100 core Th e layou s showi t figurn no . 1 e

CAPRA "POISONED" CORE WITH MODERATOR

U moderatore Th , B4C 121,introduces i empte th n dyi inert pins. Pins with moderatoe ar r uniformly distributed inside the fuel bundle (Figure 2).

CALCULATION METHOD physice Th s characteristic differene th f o s t cores were determined from equilibrium cycle diffusion/depletion calculations performed usin ERANOe gth S code cylindricaa /3/, l (RZ) core model and twenty-five groups cross sections.

In particular e sodiuth , m void reactivit s calculatewa y d with cross sections processed individuall r floodefo y voided an d d cells e sodiuTh . m void reactivit e Doppleth d an yr constant were calculated for the End of Equilibrium Cycle (EOEC) configuration using the exact perturbatio obtaio t s distributione nth wa n theorym ai componente e wels th a ,Th . s a l s by reaction type, of the reactivity effects.

190 \_y Inner core Diluent Outer core

Figure1 CAPRA Core layout

191 sixte bundlf On ho e

MODERATOR

FUEL

Figure2 CAPRA heterogeneous fuel bundle with moderator

MAIN RESULTS

The main results are collected here below.

CAPRA CAPRA CAPRA "Dilution" B4C poisoned B4C poisoned with moderator Fuel residence time (EFPD) 3x286 4x231 4x231 Feed enrichmen: t l (VolE ) % ., 40.5 40 40 E2 (Vol. , %) 44.5 40 40 Burn-up reactivity losr spe 8730 6265 6410 cycle (10-5 AK/KK') Plutonium consumption 65 65(*) 62 (kg/TWeh) Minor actinide 8.8 8.8 9.5 production (kg/TWeh) Sodium void reactivitt ya 1512 1520 1170 EOEC (ID'5 AK/KK') Doppler constant at EOEC -435 -299 -483 (ID'5 AK/KK')

(*) The amount of natural boron placed in the diluent sub-assemblies, a volume fraction of about 15% choses i , tha o plutoniune s th t m consumption rat equivalens ei thao t t t of the "dilution" core. 192 Burn-up reactivity loss

U significano n B s 4Cha t impacfuee th l n enrichmenbum-ue o t th n o pd reactivitan t y loss. Heavy nuclide mass balances

moderatoe e effecth Th f o t plutoniun o r m consumptio impace nTh . n rato t smal%) s ei 5 (- l minor actinide productio effece s mornth i t t ebu remain , importan%) 8 s small+ ( t .

Sodium void reactivity and Doppler constant

U B4C allows to reduce the sodium void reactivity effect and to increase the Doppler constant:

VARIATIONS WITH RESPECT TO THE "DILUTION" OPTION (%)

CAPR POISONEC 4 AB D CAPRA B4C POISONED WITH MODERATOR Na total void worth + 0.5 -23

Doppler constant -31 + 11

In comparison wit "standarde hth poisoneC 4 B " d cor gaine vereth e sar y importan: t the Doppler constant is increased by 62 % the sodium void reactivity worth is reduced by 23 %.

CONCLUSION

The "poisoned" CAPRA core with B4C has similar plutonium burning performances and better core safety parameters than Uthe "dilution" option. Note however that the safety n parameters of the "dilution" core could be significantly improved by using B4C. So, the main advantages of the "poisoned" core with moderator are linked to the larger diameter fuel pin : sincrease d fuel lifetim lowed ean r numbe fuee f pinth o lr n bundlsi e (397 insteaf do 469). f "BMoreovero 4s e fillinCa us e g th ,materia l doe t introducno s e supplementary technological problems.

U e "poisonedTh " CAPRA core with B4C mus furthee b t r investigate usiny db a threg e dimensions core model (power distribution, fuel lifetim controd an e efficienciesd ro l d an ) other aspects such as plutonium multi-recycling, minor actinide transmutation or higher burnup havconsiderede b o et . Then interese "dilutione th , th f o t " option remains t frow bu ,m no statee b n d ca tha t "poisonede i th t " CAPRA core with interestinn Ba Cs i g alternative. H 4

193 REFERENCES

III J.ROUAUL. al d Tan Physics of Plutonium Burning in Fast Reactors : Impact on Burner Cores Design Topical Meetin Advancen go Reacton si r Physics, 11-15 April 1994, KnoxvilleA US ,

121 G.GASTALD. al d Oan EFR - Sodium void reduction - Core with moderator Internationa- AR 4 S'9 l Meetin gAdvance- d Reactor Apris1 2 Safety - l 7 19941 , , Pittsburgh sylvanin ,Pe a

. DORIATY . . J al / d /3 Han ERANOS 1 - The Advanced European System of Codes for Reactor Physics Calculations Joint International Conference on Mathematical Method and Supercomputing in Nuclear Applications Apri3 2 - l 9 1991 , 3Karlsruhe- , Germany.

194 EXPERIENCE IN DEVELOPMENT, OPERATING AND MATERIAL INVESTIGATIO BOR-6E TH F 0N O REACTO R CONTROL AND SAFETY RODS

V.B. PONOMARENKO, A.I. EFREMOV, G.I. GADZfflEV Moscow Factory Polymetals, Moscow

V.D. RISOVANY, A.V. ZAKHAROV, T.M. GUSEVA Research Institut f Atomieo c Reactors, Dimitrovgrad

Russian Federation

Abstract

The present paper is a summary of all results obtained over 25 years of control rods operation, and involves the following data: (1) design and characteristics of the different type f controso l rods tested durin BOR-6e gth 0 reactor operation mai) (2 n d an ; result post-irradiatiof so n examination.

1. Introduction unique Th e data presente behavioun do absorbinf ro desigd gan n materiald san five modification reactof so r control rods under high fluence irradiatio obtaines nwa d ove e BOR-6year5 th 2 r f so 0 reactor operation. Every type of rods was developed in terms of the experimental data obtained unde BR-5e rth , BR-10, BOR-60 reactor irradiatio absorbinf no g material specimens and their subsequent examination e RIA th PhEd t t Rcella an sn i ho welIs a ss a l performanc whole th f eo e comple priof xo r irradiation tests (technology development of material d absorbinan s g elements (AE) production, compatibility testingf o s absorbing and structural materials, joint effort of rods with drives, etc). The present paper is a generalisation of all results obtained over 25 years of control rods operation. The paper involves the following data: - desig characteristicd nan differene th f so t type control rods tested durine gth BOR-60 reactor operation; - main results of spent rods investigation.

2. The BOR-60 Control and Safety System e BOR-6Th 0 contro safetd an l y system involves seven rods: three safety rods (SR)automatio tw , c control rod shiso (CRmtw d rod)an s (RR-1, RR-2). boundare th t a p cor f radiau yd o t ean se le reflectorar s CR e , (RR-1e Th th n i - ) cente(RR-2e d fourtth an coref n i w ro - )h ro . Durin reactoe gth r operations SR e th , above loweR ar coree C eth e r parth ; t (1=200-25 submerges i ) core 0mm th t ea n di all times; as fuel burnup, RR-2 is moved upwards from its extreme position and RR-2 is withdrawn from the core while bringing the reactor to power. During the reactor operation the safety rods undergo a small thermal load and subjectee ar relativa o dt e light radiation effect majoe Th . r requiremen reactivits i t y

195 worte uni f volumeth o t n h i e requirementTh . r radiatiofo s n resistanc head ean t remova lese ar ls stringen attributed an t d lowee onlth o yrt par rodf o t s located close to the active core. cor e subjected th ean n i Durine o ar dt s reactoe gth RR d r an operatio s CR e nth intensive irradiation. The requirement for effectiveness of those rods per volume unit is suitabl small-sizee th r efo d reactors wher numbee eth rod f volumd ro san smale ear l This requiremen less i t s stringen large-sizee th r fo t d reactors e majoTh . r featurf o e these rods should be high radiation resistance. While bringing the reactor to power, the shim rod (RR-1, RR-2) compensating the temperature change should have the maximum effectiveness and only a part of shi should mro radiatioe db n resistanc ewithdraws i becaus d ro e eth n fro reactoe mth r firse zonth t t reactoe staga th f eo r operatio smale th d l nneutroan n flux sevel Al . n rods operate in the hexagonal channels of 42 mm inner diameter and are cooled with sodium flux at inlet temperature from 340 to 350°C.

3. Rod Design t firs A accordancn i t e wit e BOR-6hth 0 draf e rod th f controt o s safetd an l y system were developed on the base of boron carbide with sealed absorbing elements (AE). desige Th n (Fig.S.la) involved head, tail-en OCrlSNilOTd dan i steel carrying tube 39.5x0.8 mm diameter in which seven (SR, RR - 1057A-0) or four (RR - 1065A-0) absorbing elements are located. The absorbing element is OCrl6Nil5Mo3Nb steel cladding 12x0. diametem 4m r filled with boron carbidr eo chromium deboride pellets 10. diametem 8m enrichmentB 10 r wit% 80 he Th . absorbing elements are sealed and filled with air or helium in the manufacture and have the gas collector. The reliability of rods was provided for the corresponding matching of the radial gap between the absorber and cladding as well as the formation of necessary plenum for gas collection. The next stage of control rods development involved the manufacture of vented designs filled with boron carbide pellets. Unlike the above design the gap between claddin borod gan n carbide pellet filles swa d wit collectos t hga sodiuno e s th t rwa mbu available (1057D-0, Fig.3.lb). To provide the large operative reactivity margin for increase of microcampaign, the single-absorbing vented design with boron carbide pellets of 80% 10B enrichment was developed (1309, Fig.3.1 , c). The rod involved head, tail-end and OCrl6Nil5Mo3Nb steel carrying tube 36x0. diametem 8m whicn i r h boron carbide pellets 32-33 mm diameter were located. uppee pluth d n en rgI speciaa l vented useunis betweep alloo dwa tga t e wth n boron carbide pellets and cladding to be filled with sodium immediately in the reactor. When using the rod as SR, in the lower part of design the ampoules may be used for irradiation testin differenf go t design materials. At first time the europium oxide containing rods were widely used in the BOR-60 reactor. Their desig s similai n o rodt r f boroo s n carbide with sealed absorbing elements. The europium oxide pellets 11 mm diameter and density no less than 7 g/cm3are used as an absorber (1065D-O, Fig.3.1,d). increaso T effectivenese eth neutrof o s n absorptio e "trapnth " typ desigd ero n based on europium oxide and zirconium hydride was made (1847, 2121) (Fig.3.1e). It involves head, tail-end, two carrying tubes 39.5x0.8 mm and 27x0.5 mm diameter filled with annular europium oxide specimens 37. 27.d 6an m oute2 inned m an r r diameter and the moderating element located along a center of rod. The moderating

196 element (ME) is a cladding 18x0.4 mm diameter in which zirconium hydride blocks 17 mm diameter are located. The claddings are made of OCrl6Nil5Mo3Nb steel. The inne re absorbin gapth f o s d moderatinan g g element e fillear s d with heliud man sodium, respectively.

FIG. 3.1. Control rods design of reactor BOR-60: 1309. RR 1057D, RR 1057A,SR, RR SR, c- a-b- SR,

197 1 1

FIG. 3.1. (cont.) 1065D, d-"trap"CR CR e- 2121.

198 . Physica4 l Effectivenes f Controso l Rods e physicath Dat n o a l effectivenes rodS CS s f involveo s d boron-containing absorbing materials are presented in Tab.4.1. Application of the single-absorbing element design in the RR-2 rods (B4C pellets 10.diameterm 8m ) increase e effectivenesth s compare% 30 y b s d wit e sevenhth - absorbing elements design (B4C 10. diameter)m 8m desige th .s f Amonno CR e gth 80% boron carbide enrichment has the most effectiveness. The CR-trap containing europiue th m oxide absorbe zirconiud an r m hydride moderato effectivenese th s rha s by 30% higher than that of the control rod based on europium oxide. Measuremen e controth f effectivenesd o t ro l s after operatio e BOR-6th n ni 0 reactor showed it was practically unchanged at least up to 4% burnup.

TABLE 4.1. EFFECTIVENES BORON-CONTAININF SO RODS GC S

sc Radius, Number Absorber Average IOB Effectiveness, % ~ rod cm ofAE type neutron actual energy, content, Calcul. Measurement E, KeV g

RR-2 0 7 B4C 480 395 1.82 1.90 ±0.02 '

RR-1 16.2 7 B4C 405 380 0.95 1.06 ±0.02

SR-1 13.5 7 B4C 440 393 1.25 1.25 ±0.02

SR-2 11.9 7 B4C 452 392 1.41 1.34 ±0.02

SR-3 18.0 7 CrB2 384 275 0.65 0.84 ±0.02

CR-2 23.4 4 B4C 250 218 0.36 0.36±0.01

CR-1 25.1 4 CrB2 250 158 0.24 0.27 ±0.01

TABLE 4.2. EFFECTIVENESS OF CONTROL RODS OF DIFFERENT DESIGN

CS rod Absorber Numbef o r Radius of Effectiveness, AEin CS location, cm O/o/ -%-A« rods

RR-2 B4C 7 0 1.70 + 0.02 RR-2 B4C 1 0 2.20 ±0.02 CR-2 B4C 4 23.4 0.36±0.01 CR-2 Eu2O3 7 23.4 0.20±0.01 CR-trap Eu2O3 1-AE 23.4 0.26±0.01 ZrH2 1-ME CR-1 CrB2+(Ta,W) 4 25.1 0.27±01

199 5. The Main Results of Spent Control Rods Investigation Since the BOR-60 reactor has been put in operation (1969), more than 80 control rod differenf so t design were successivel rod0 2 y s d inserte weran t i e laten di r analyzed in the material hot cells in detail (Tab.5.1). During operatio controe th f no lBOR-6e rodth n si 0 reacto caso n rf failur eo e of the main function actions was observed due to decreasing of their operational reliability. In this case no marked loss of physical effectiveness as well as break of mobilit guide th n eyi cartridges were observe failuro rodse t th e f .do e du Ther e took places three cases of premature withdrawal of the control rods from the reactor zone becaus cartridgee th f rodS eo C sf werso e failured. rode Th s . containinCR g boron carbide (thermally recovered with carbod nan magnesium), chromium and tantalum diborides and europium oxide as well as the trap-rod (europim oxid zirconiud ean m hydride) were investigated. RR-2. The rods of three design were investigated: - seve sealeE nA d wit collectors hga ; - seven AE vented with sodium sub-layer; - one AE vented with sodium sub-layer.

rode Th s . composeRR-1 SR containind seale.e an th E f do A g boron carbide (thermally recovered with carbon and magnesium) as well as the SR after prolonged storage in the cooling pond and the SR containing refabricated boron carbide were investigated maie Th . n result presentee sar Tab.5.2n di .

TABLE 5.1. MAIN CHARACTERISTIC INVESTIGATEF SO D CONTROL RODS

Rod • Design type Absorber Filling Operation conditions No. in gap Oper time, Power Neutr Max absor 1CB bumup, effday production fluence, tempC ° , %, aver MWh l(fcm2 RR-2 1057A-E 07A B,C air 215 205320 E>2 4 0 630 2.6 RR-2 10S7-E D7A B4C sodium 2572 241550 45 440 26 RR-2 1057-E D7A B4C sodium s * 551987 87 440 77 RR-2 130E iA 9 B4C sodium 540 522000 103 540 7.2 CR-2 106SA-0 B.C air 215 205320 27 500 45 CR-1 1057A-0 CrBj air 306 295046 36 500 77 CR-2 1057A-0 B^C magn air 271700 36 29 CR-2 1065D-O Eu203 helium 348 337906 47 630 CR-2 1065D-O Eu203 helium 729 787000 105 630 CR-2 1847 Eu203 helium 542 520367 69 ZrH, sodium CR-2 2121 EuA helium 601121 64 630 ZrH, sodium 463 RR-1 1057A-0 B,C air 306 293457 48 500 23/42 RR-1 1020 B

200 TABLE 5.2. RESULT INVESTIGATIOND RO F SO S

Rod Design Testing conditions Results

thermallC 4 B % y80 R C 205320 MWh Wrapper and AE withhout defects. recovered with 2.7xl(fn/cm2 Unchanged diameters of AE. carbon 1-10% gas release under E>0 irradiation pelletC intacte B4 . sar . max.burnup - 5.2% sealeE 4A d with B4C pellet swelling ( %) 0.9-1.5 - gas collector botto, 0.4-0.AE f m 9o - center , 4.1-4.2 - top. Mechanical propertie claddingf so mose :th t increase of strength and decrease of bottoE , plasticitA 20 e mt a th t ya 300, 600 °C (21% plasticity at °C)0 600°an2 . t a % d4 thermall80C %B„ y 271700 MWh WrappewithouE A d tan r defects. recovered with 3.6x10" Increas diameteE A f eo 0.5-1%y rb . magne- E>0 25% gas release under irradiation. sium max.burnup 2.9% intacte mose ar Th C 4 t. parB f o t sealeE A 4 d with Great spread under swellinf go gas collector pellets.( %) 1-3.5 - top; 3.5-8.5'- bottom. Mechanical propertiee sar similar to those of the first.

CrB2+TaB2 - 80% 29345h 7MW Wrapper and AE without defects. 4 AE sealed with S.oxlO^O Unchanges d diameteGa . AE f o r gas collector max.burnup 2.6% releas t irradiatioea n- 77% . Swelling of pellets ( %) 1.6 - top; 2. 0center- ; 3. 0bottom- .

Eu2O3 33790h 6MW WrappewithouE A d tran defects. sealeE 7A d 4.7xl022 E>0 Unchanged diamete claddingsf o r . 10.95-11.0 diametem 9m pelletf o r s I.D.)(11.m m 2 . Unchanged mechanical properties.

Eu203 78700h 0MW WrappewithouE A d tran defects. 7 AE sealed 10.5xl00 2E> 2 Increas wrappef eo r diametey rb 1.5%. Increas claddinE A f eo g diamete Pellet. 1% sy rb diameter s- 10.95 -11.09 mm. Unchanged mechanical properties.

201 TABLE 5.2. (cont.)

CR-trap 520367zMWh Block of moderator without - Eu E 2OA 3 1 sealed 6.9x10 E>0 defects. Unchanged diameter of 1 ME - ZrH2 cladding. Hydride- ) block% ( s- vented 0. 5- 1.3 . Unchanged hydrogen content. Faiiured absorber block. Increase of outer cladding diameter by 1.3%. Eu2O3 swelling ( %) 2.8 - top; 4.1 - bottom. Mechanical testin t 350 ga 650°d an ° - zero plasticity. CR-trap 601121 MWh AE and ME without defects. Z2 1 AE - Eu2O3 sealed 6.4xlO E>0 Unchanged diamete . 39.52ME f -ro 39.65 mm AE diameter (initial value - 39.5 mm) RR-2 205320 MWh Wrapper without defects- E A . 7 AE sealed with 4.2xlOME>0 residual bendin g20-3- . 0mm gas collector burnup - 3.1% Cross destruction of one peripheral B4C 80%. AE. 1-2% increase of cladding diameter in nondestructive AE and 4% - in destructive regions. Four sealed AE, 1-10% gas release. Tight-fitting pellets to cladding. Swellin pelletsf 4.3-4-) go % - .9 ( : bottom; 4.2-4.6 - center; 4.0-5.1 - top. Change of mechanical propertie claddingf E so A l al r sfo zones; the more fluence, the more changes. RR-2 7 AE vented with 55200h 0MW Wrappe withouE A d tran defects. Na SJxlO^EX) Increase of AE cladding diameter sub-layer max.burnup 9.0% by 1.4% at the bottom. Increase of pellet diametep to e 3-4y th rb t %a cented e d 5-6%an th an t r a bottom. Mechanical properties of cladding more :th e strengthe th , more fluence and the low irradiation temperature. 3-4% relative elongation for all sections 1-2d t an 350-55% a ° 20 . t 0a °C Compatibility: B^C-ciadding interactio centee th f n i ro periphera - 0.15mm E lA . venteE 1A d with 51800h 0MW Wrapper and AE without defects. sub-layea N r E>0 Increas claddinf eo g diametey rb 10.3xl022 1.9% at the bottom. B„C pellet burnu p7.2- % top- - swellin2 6 ;2. 3. ) % g( center bottom- 6 6. ; . Mechanical properties wer ehige holth h t da level.

202 TABLE 5.2. (cont.)

1 2 3 4 RR-1 sealeE 7A d wits hga 293457 MWh Wrapper withoug t sa defects m m 5 . collector 4.8xl022E>0 of AE. Unchanged diameters of B4C thermally average burnup - claddings. Sealed AE. 1-10% gas recovered with 2.3% release -1.22 0. . % increasf eo carbon cladding diameter. sealeE 7A d wits hga 28270h 0MW Wrappe withouE A d rtan defects. collector 5.6xl(f(E>0) Unchanged cladding diameter. 10- B4C thermally 20.5% gas release. 2.5-6% (bottom) recovered with increas pellef eo t diameter. magnesium SR 7 sealed AE 974276 MWh Wrapper without defects. Severe 1.2xlOTE>0 bending of peripheral AE. Increase max. burnup -4.3% of cladding diameter (bottom) up 1-2%o t releass Ga . e 18-24%. Destructed pellets of peripheral AE. 4.5-4.8% (bottom) swellinf go pellet. %) s( SR 7 sealed AE 51630h 0MW Wrapper and AE without defects. 5xl022 1.5% max. increase of gas collector 1-2% average diameterE A d destructio% an 40 . n burnup of B4C pellets. Diameter of B4C 18 -year storage in pellets did not exceed 11.0 mm/ pond Mechanical propertie claddingf so : increased strength, high plasticity. SR 7 sealeE dA 54955h 0MW Wrappe withouE A d rtan defects. 22 refabricated B4C 4.2xI0 E>0.1 39.5-39.68 mm (39.5 mm initial) diamete wrapperf ro . Unchanged diameter of claddings. Shape and integrity of boron carbide pellets were kep . 5.8-7.7up t releases %ga . 10.85-11.01 mm diameter of pellets. Mechanical properties of cladding: 2.9-6.5% plasticity, 975- ultimata 1525MP e strength within 20-60. 0°C

203 PRODUCTION OF GAMMA-SOURCES, BASED ON EUROPIUM OXID FASEN I T REACTORS

V.D. RISOVANY, A.V. ZAKHAROV, E.P. KLOCHKOV, T.M. GUSEVA Research Institut f Atomieo c Reactors, Dimitrovgrad

V.B. PONOMARENKO, V.M.CHERNYSHOV Moscow Polymetal Factory, Moscow

Russian Federation

Abstract

papee Th r present basie th s c result f europiuso m oxide operating propertiess it , physical efficiency, radiation resistance and compatibility with stainless steel; it considers structuree th characteristicd san y-sourcef so s mad f europiueo m oxide irradiatef do nuclear reactors controls absorbing cores sets.

INTRODUCTION

Conditions of fast reactors operation are characterized by the hard neutron spectrum, high neutron fluences and operating temperature, corrosive medium which makes a number of requirements for controls absorbing materials, the main of such requirements are: 1. High neutron absorbing efficienc storags it procese d th n yean i operationf so . 2. Resistanc radiatioo et n damages tha revealee tar changn d i forme th f eo , sizes, structure, physico-mechanical properties in the process of reactor irradiation. 3. Compatibility with structural materials. In the aggregate of these requirements europium oxide is one of the most promising materials whic confirmes hi greay db t positive experienc usags it t f eo eno BN-reactore onlth n yi s controls therman i t bu , l neutron reactor same wells th sa t e .A lase timth t yearn ei s there appeare dtendenca reduction yo usags it f Russiann o ei n nuclear reactors. It is caused first of all by the enhancing requirements for ecology and safe operation of nuclear reactors. Under reactor irradiation in europium oxide accumulate e europiudar m radionuclides 152Eu, 154Eu, 153Eu, which arises serious problems with storage, reprocessing and transportation of radioactive wastes and also can cause catastrophic consequences in case of damage of the nuclear reactor controls and core. Still, complete refusal against usag europiuf eo m oxid materiald ean s it n so basis with accoun theif o t r unique propertie t expedientno s si . Their high activity after reactor irradiatio effectivele b n nca y use powerfudn i facilitie- y l s instea appliee th f do d 60Co and 137Cs based sources. presene Th t paper present basie sth c result operatinn so g propertie europiuf so m oxide, its physical efficiency, radiation resistance, compatibility with stainless steel, and considers the structures and characteristics of ^-sources made of europium oxide irradiated in the set of nuclear reactors controls absorbing cores.

205 1. Nuclear and physico-mechanical properties Properties of europium oxide are described in detail in refs. [1-3]. Table 1 presents the most important characteristics from the operation viewpoint. Eu oxide wit monoclinie hth c structur highee th s reha nuclear densitys it o s , usage in this modification as an absorbing material is most preferable. The high structure stabilit meltind yan g temperatur oxidu E f eo (2330°C [2]) shoul notede d b disadvantage .Th e include thermaw lo s sit l conduction (2-3 W/m°C)

TABLE 1. BASIC PHICIAL CHARACTERISTICS OF EUROPIUM SESQUIOXIDE [2]

Properties Monoclinic Cubic structure structure

1. Melting point, °C ~ 2300

2. Density> g/cma 7,921 7,287

3. Absorber cation desity , atom/cm3 2,72*10 • 2,49*10ZZ

4. Thermal conductivity coefficient. V/m.cK t a 25*- C 3,42 t 100a - 0 °C 2,38

5. Thermal expansion coefficient,

- from 30 to 840 °C 10,5*10~* 35*1-, 10 fro06 ~ m 0 to 1000°C

6. Young's modulus, N/mm2- 1,33*10*

7. Poisson's ratio 0,25

8. Transformation rate t 110a - 0 C several days hou1 < r - at 130 C ° 0

206 which sets restriction productior sfo large th f en o diamete r absorbing core becausf eo its great heating under reactor irradiation. possesseu E s unique nuclear properties bein bese gth t representativ vien ei f wo this index among all the known neutron absorbers. Natural Eu consists of two isotopes 151 I53d EEuan u wit mase hth s content 47. 52.d 8 [2]an 2% . Their thermal neutron- absorption cross-sections make up 9,000 and 420 barns [2], respectively. It is notable that europium effectively absorbs fast neutrons too, inferior onl isotopo yt e s 10i Bt I . this fact that substantiates oxidusagu E f fasn eei o t reactors controls. During neutrons capture ther e producear e d daughter isotopes 155Eu, 152Eu, 154Eu, 156Eu, 157Eu, the fission products of which are gadolinium isotopes (Tab. 2 /2/). All these isotope alse sar o characterize higy db h neutron-absorption cross-sections. Therefore, physical efficienc f neutroo y n absorptio t oxidpracticallu E no y s ei nb y decreased under long-term reactor irradiation.

TABLE 2. SCHEME OF NUCLEAR TRANSFORMATION AND EUROPIUM NUCLIDES DISINTEGRATION [2]

Izotop Thermal neutron absor- Hal f- lif e Desintegration ption cross section products

Eu 9000 barn stable n (47,8%)

5000 barn z" 13,54 years 152, 'Eu -> Gd 1 » 153,Eu 420 barn stable (52,2%) n 1500 barn »- 8,59 years 154 Eu > Gd n 14000 barn .- 1,7 years 15SLEu n r 15,4 days Eu n ~ 15,4 hours 157 Eu

207 . Radiatio2 n resistance Numerous investigations showed that the size stability of Eu oxide under reactor irradiation doe t practicallsno y depen spectrue th n do neutrod man n fluences i t bu , determine irradiatioe th y db n temperature alone (Tab.3) [4]samplee .Th s retain their shape and integrity, swelling does not exceed 1%, if the following conditions are satisfied: - maximal irradiation temperature should not exceed 1600-1700 °C; - temperature gradient should be less than 45°C/mm.

TABL . ESIZ3 E STABILIT EUROPIUF YO M SESQUIOXIDE SPECIMENS UNDER IRRADIATION

Irradiation conditions.

fluence Temperature, °C id/d, Statf eo Ü102'' cnf* mm spec meni t (E>0,lMeV sure )th -n o at the cen- AT, face tre °C/mm

0,22 890-930 1350-930 45 0 1 , 2cracks

0,39 760-800 1210-1250 55 - destruction

1,40 920-960 1380-1420 55 0,5 cracks

2,72 1100-14000 1800-1850 80 4-5 destruction dissociation

70,0 390-440 660-700 60 0,6 cracks

80,0 630-680 840-890 45 1,0 craks

120,0 630-680 840-890 45 1,0 cracks

Figs2 presen1, . t cross-section d structurean su oxidE f eo s after reactor irradiation. The number of cracks increases with growth of the temperature gradient. The structure of Eu oxide up to 1600°C does not practically differ from the initial one. Increas irradiatioe th f eo n temperatur 1800-1850°o et increaso t e d Cth le f eo pellets diameter up to 4-5%. Empty voids were formed in their centre. Three zones are distinctl ystructure seeth n ni e (Fig.2). Closecentre th o rt e ther colume ear n grains

208 a b

Figure 1. Cross section of Eu^Os pellets after irradiation: a = tmax ft; 900°C; F = I^IO25 cnf* tma= b x1400*C« 1,4*10= F ; AI cm"* c = tmax * 1850°C; F = 2,72*10at

209 Bode r Centere

Figure 2. Eu^Os structure after irradiation (tma 1850"Cx= ; F=2,72*10*' n/cm*) x 800

turning into equiaxed large size grains and, finally, besid pellete eth s surface there ear small grains. Suc hstata explaines ei masy db s transfeproceedino t u e E du e r th f go oxide dissociation, particularly, according to reaction EiijOj -» 2EuO + Ot [1]. This proces s alsi s o characteristi f non-irradiateco oxidu dE e samplee s th heate o t p du temperature above 1700°C (Fig.3). Positio porositf no thin yi s cas differents ei . Under irradiation it is maximal in the samples centre, where the strongest heating is observed. whild An e general heatin electrie th n gi c furnac uniformls i t i e y distributed ovee rth whole volume. According to X-ray investigations irradiation leads to Eu oxide amorphisation when grain boundaries disappear. High radiation resistance of Eu oxide at the irradiation temperature up to 1600- 1700° accounteCs i peculiaritiey b r dfo crystalline th f so e monoclinic structure, that is characterize presency db large th f eeo quantit stoichiometrif yo c vacanciesn ca t I . be referred to the "porous" structures class, where the focusing atomic lenses of the closest environmen compacy an r fo t t packed tren destroyede dar , which suppresses focusin d canalizatioan g e primarth f o n y radiation defects [5]. Thus, mosf o t internodal atoms generated under reactor irradiation r frocannofa m t "theirge t " vacancy at a great distance and form stable radiation defects.

. Compatibilit3 y with stainless steels Eu oxide does not interact with austenitic class steels up to 900-1000°C. The main components of stainless steels are iron, chromium, nickel, molybdenum, manganese importans i t I . t that full energ oxidu E f greates yei o r than for these elements oxides [2] freo .S e energ reactioe th f yo n 2Me+Eu203-^Me2O3+2Eu (where Me-Cr, Mn, Fe, Mo, Ni) will be positive and proceeding of these reactions is not expected.

210 « " ' • .c--. " -*-'••' ."- '" "Z?;'•

«^ •",'*' ^- -• '"'. ''.'', .' '-*.-' ' " "* '**{.-• -•'f

v •".""-*-•<-;.' • '..''

i. V-ft v*?- ;^^%1^KFj^:^ * *t * '•

Figure 3. Structure ofEu^Os a t = 1560°C, ^r = l h. , x 100 b t = 1700°C, t = l h. , x 200 c,d t = 1800°C, t = l h. , x 200

211 Samples Eu203 + Me (where Me = Ni, Cr, W, Nb, Mo, Ta, Re, Cu) were produced for reactor tests. Irradiation was conducted in special capsules in the BOR- 60 reacto heliun ri m atmospher t 400-800°ea faso t tp Cneutronu s fluence 2-1022cm~2 (E > 0.1 MeV). The samples retained integrity, the volume changed by a factor of 2- 3%, interaction traces were absent (Fig.4) which confirms high compatibility of Eu oxide with these metals.

Figure 4. Typical structure of Eu^Os + Me (where Me=Mo,Re,Ta, W, Cu, Cr, Ni) after irradiation (t-400-800 C, F=2*10,2 2n/cnT ), X200

4. Eu oxide-containing controls serviceability

Typical structure f absorbinso g elements wit oxidu hE t fasea t nuclear reactors are displaye Fign di . .5 Post-reactor material science investigations showed thae maith t n factor, restricting the operating time of these structures, is low radiation resistance of structural materials achievinn O . e fasgth t neutrons fluenc 102 1. e1 2 20. cm" > E 2( MeV) swelling of the wrapper and claddings up to 10-12% occurred, while all the Eu oxide pellets retained their sizes, shape, integrity (Fig.6). There were no traces of pellets-claddings interaction (OCrlSNilOTi, Crl5Nil6Mo3Nb). Any noticeable reduction of physical efficiency of neutron absorption by controls was not detected. If the structural materials were replaced by more radiation resistant ones, the Eu oxide-containing controls operatin increasedge b tim n eca .

212 /i\\\\\\\\\\\\\ \\\\ \\\x\\\\\\\\\\\\\\\

,* r3 ^. -/! ~-'i **• < 5: — A \ \ \ \ \X \\\ \ \ \ \ \ \\\\\\ \ \\\ \\\\ \ \\

\\\\\\ \ \\\v\ \s \ \ \\\\\\\ \ \\A\ \ \ \ \ \ rrr . * *v II.-. , • •»« *• *v* • « .. '. •** (%ï ü r *" OJ* » •*"<*"' * ***. ~+ t • v *« * ^^ * "••\ ;.-. - •*• . M'.;.- ' :^ ; ' s'J »vT.; .*. •»* :;-:* * • * LÜJ *iiM* * < * * \\\\\\ \\x\\\\\\\ \ M \ \ \ X\

Figur . Typicae5 l desigcontroe th f no l element with Eu2O3 of the control system rods of the reactors BN-60 BOR-60d 0an .

5. Spectral characteristics of Eu radionuclides Main characteristics of some Eu radioniclides compared to 60Co are given in Table 4. As the Table shows, 152Eu and 154Eu have spectral characteristics that are rather attractive for ^-sources. 45% of probability of ^-quanta yield per 1 decay have energe th ydecaf o 0.866...1.2 154r % 152yr Efo uprobabilit65 fo Eud an V , 1Me y have the energy 0.72...1.0 MeV. In terms of the values for ^-constant K and radioactivity equivalent 1 mCi/1 mgR radionuclidee ath s 152 154fold E2 uEde an uinferioar mose th to r t commonl y used industrial ^-source 60Co, but they considerably exceed the analogous values for I37Cs, that is also widely used as K-sources at the industrial facilities. The advantages of 152Eu and 154Eu radionuclides over 60Co include the larger half-life period and decrease the costs due toi reducing the expenses for their power production (larger neutron-absorption cross-section, irradiation in the controls set).

213 X4 xl

Figure 6. Pellets of Eu203 after irradiation , 2xl0(900°C= 1 t = 23 n/cmF , 2)

TABL . CHARACTERISTICE4 NUCLIDF SO , l54EuEu 2 , 15 E155 Eu, 152.., Parameter Eu Eu J55Eu '«Co

Life - time,Tj£ , year 13,54 8,59 1,7 5,27

Full if - constant, 6,35 6,70 0,33 13,2 Kj» , R/h

Equivalent of iMKu isotope 0,60 0,74 0,01 1,54 per 1 mg Ra

Averade energ , yIvfe V 0,709 0,745 0,1 1,252

. Design6 characteristicd san f p-sourceso s

RIAC ASS tR (Dimitrovgrad) there were develope firse dth t standard ^-sources u oxide-containinE froe mth g absorbing element se SM-spenth n i 2t reactor. Preliminarily prepared absorbin gdiameten i elements m witd m wale an 1 rhth 4. l, thickness 0.3 mm, are located in the cladding, llxl mm in diameter (Fig.7). The elements are 75-400 mm long. The exposure dose rate (EDR) of the Eu oxide based sources at a distance of l m is equal to 0.2...0.6 r/s which is comparable to the 60Co based sources. Fig. 8 demonstrates the ^-source design with an insert made of Eu oxide or Eu2O3+Co composition. The inserts are pre-irradiated as an absorbing core of the controls. Characteristics of various source modifications are presented in Tab.5 214 ^,vv\ \\x\\\\ \ \\\\ \ --\xx\\ \ \x\x \ . \ x \ \ ^ : : : : : < > : à 's-:-ï':-'.ï- - .' ': .\:-ï-'-ï- •:•'•'•'. 'ïï /S. .ï. ^v^'-v- ^•'i.^;\'^! -^»:^^'----:^- - ^':v::'i-| /^> I ^'''^^\-'\'<-:--.--.'-:C'-::.-':\-':\'-/-:^'-.-^:;\"^ ^ 0 / / \ '// >T //^ t-\ f^^ •-*'•;- '»••..-, • • •»'.-• • " • . ' •*.".•".'• " •T-**---".".:",- ,r* • " * *-•,'•**•., > • • •»/'* yf** v* . •-.'"'f A : : : u // / ::"•.•.IV . :.,-<:-'o v./-:- .-/-.? : . - .-'..v^v:^:>'.:-:-î-^--«.-- -:-. -.vî'--.-.i?.- - --- « ':~>-;-.| // lAu.i'i-^%0:7.:^";^'-^:V'--;-^v;::x'siv^^v>/V^^^^ x y. • / ^k X X X • X XXXX\\X\X \X \XXXXX X\XX\\XX\X'\TV "eT"

Figure 7. Gamma-source design based on the control elements with which wer reactoee th worke n i rf SM-dof 2

Two variants of active core: first variant

Co+Eu 03(5-90%)

1,5 cobalt bung Figur . eGamma-sourc8 e design base europiun do cobaltd man . 1 - active core Eu, Co+Eu 2 - inner hermetic capsule 3 - external hermetic cover.

215 a

Figur Externa. e9 l"rod-trap"wite vieth f wo h Eu^O) s(a and cross section ) aftes(b r exploitation ni the reactor BOR-60

216 TABLE 5. BASIC CHARACTERISTICS OF GAMMA-SOURCES

Exposure dose po- External Active core wegammf o r - radi- sizes, mm sizes m ,from m 1 mt a n io t â Source the warking sur- Type and face of a source. number

Diamet. Height Diamet. Height. Nominal Error, D H D H powers .R/ %

GSEC-7-2 0,155 20 GSEC-7-3 11,0 81 , 0 8,0 75,0 0,387 30 GSEC-7-4 0,775 30

GSEC-7A-1 0,155 20 GSEC-7A-2 12,8 86,0 10,0 80,0 0,387 30 GSEC-7A-3 0,485 20 GSEC-7A-4 0,775 30

GSEC-12A-1 0,155 20 GSEC-12A-2 11,0 99,0 8,0 94,0 0,387 30 GSEC-12A-3 0,485 20 GSEC-12A-4 0,775 30

Conclusion Great positive experience has been gained on long-term operation of Eu oxide- containing control fase th t n nucleasi r reactors BN-60 BOR-60d 0an oxidu E .s eha high neutron absorption efficiency which doe t practicallsno y change under reactor irradiation. At the irradiation temperature up to 1600-1700 °C it retains its shape, integrity, swelling achievin o t doe t exceep u sno fase % d1 gth t neutrons fluenc2 1. e 1022 cm"2 (E > 0.1 MeV). Eu oxide does not interact with austenitic steels up to the irradiation temperature 900-1000 C. The operating time of Eu oxide-containing controls in the BN-reactors is limited to degradation of the properties of stainless steel elements. High-active, long-live radionuclideu dE s Is2Eu, b4Eu, 155 accumulateEe uar n di Eu oxide under reactor irradiation which makes it possible to use them as ^-sources. First design f sourceso s base controln do s spen nuclean i t r reactors were produced. Two-purpose application of absorbing cores, made of Eu oxide and materials on its

217 basis, allows reductio expensef no controlr sfo s operation, solutio problee th f no f mo radioactive wastes utilization, production of cheaper ^-sources compared to 60Co and 137Cs ones with comparable operating characteristics.

REFERENCES 1. Murgatroyd R.A., Kelly B.T., Technolog assessmend yan neutrof to n absorbing material / sAtom/ . Energy Rev. , 3-74l , 1977N ., 15 , 2. Paul C.S., Europium as a nuclear control material// Nucl.TechnoL, Vol.39, 1978, P.84-94. . Bejiae3 B P.A apH .. CsoöcTBa OKHCJIOB eBporraa , AroMH3ÄaTM. . , 1974, c.52 . 4 PHcoBaHHHH B.fl. PaanaiiHOHHbie noBpexaeHHa OKcnaa eBporraa// rperbeo MeacoxpacjieBOo KOH

218 REPROCESSIN IRRADIATEE TH F GO D BORON CARBIDE ENRICHE BORON-1Y DB 0 ISOTOPREUSE S IT TH D EN I EAN CONTRO FASLE RODTH T F BREEDESO R REACTORS

V.D. RISOVANY, A.V. ZAKHAROV, E.P. KLOCHKOV, A.G. OSIPENKO, N.S. KOSULIN, G.I. MIKHAILICHENKO Research Institute of Atomic Reactors, Dimitrovgrad, Russian Federation

Abstract

The present paper discusses the development of technology for reprocessing of irradiated boron carbide, which provides complete removal of radionuclides from irradiated 0 1 materials. This technology allow enriche repeatee B sth f o de dwitus h B 4fasCn i t reactors.

As a neutron absorber in the control rods of the fast breeder reactors the boron carbid boron-1f o e% enriche80 0o t isotop p du useds ei . e onlThisth ys , i material having the sufficient physical absorbing efficiency in the fast neutron spectrum to perform the function of the emergency reactor shutdown. Due to the enrichment this material is very expensive. On the other side, according to its in-reactor operating condition safete sth mose y th rodelevatee time e tth parsar th n ef i o td positiod nan the absorbing pin is subject to the neutron flux only in its lower part that is adjacent coree examinatioe th Th . severaf no l safety rods discharged fro BN-60e mth 0 reactor demonstrated tha maximue th t m burnu boron-1e th f po 0 isotop lowee th n rei parf o t the pin didn't exceed 5%. In the upper part of the safety rods the burnup of boron-10 t noticedno s wa . n pi e inth Thu in-reactoe sth r operatio safete th f nyo rod finishes si d when they still have a great physical efficiency for neutron absorbtion. The reason for this lies in the radiation damage of the steel structural parts and first of all those located in the lower part of the pin and those operating in the reactor core (Fig.l). The SSC RIAR has performed the examination of 4 safety rods spent in the BN-600 reactor from 100 to 330 effective days. It was pointed out that the lower extension part of the absorbing core, discharged afteday0 operationf 33 rso substatiaa d ha , l bend causenone th -y db uniform material swelling (Fig.2). Fracture toughnes materiae th f so thin i l s part (steel 09Crl8NilOTi) is very low (less than 0.3 MJ/m2) and when testing on the impact mashine the samples are subject to brittle failure. But at the same time the ductility of the cladding material of the absorbing pin (steel 08Crl6Nil5Mo3Nb) was decreased up to 3% at 500°C. This testifies that the service life potential of the safety rods of this design is exhausted prolonTo . servicgits e lifsubstantiaethe l modernizatio requirednis . First of all it is necessary to replace the structural materials for the more radiation resistant materials. This procedure is being developed and during the last campaigns the new safety rods of more radiation resistant steels have been tried in the BN-600 reactor. Considerable benefit to the extension of the safety rod service life can be updates i d achieve ro thay suc e design de t i doe i twa t th hav th h a f df i e no s n o eth lower extension part radiatioe th , n damag whicf eo h limit service sth safete e th lif f yeo rods. These measure somo t n es ca exten t prolon service gth e lif safetf eo y rod t thesbu y are not capable to resolve the problem of the rational use of the expensive enriched boron carbide.

219 T——r

«N

b designreactorFIG.a- SR 1. of BOR-60, reactorb- BN-600.

220 a-

FIG. 2. Lower extension parts ofSRs ofBN-600 after exploitation a- 331 efpd, b- 146 ejpd

This problem can be resolved when the pin is reused for manufacturing the new scram rods and this feasibility was tried in the BOR-60 reactor of the SSC RIAR. The safety rods were withdrawn from the storage pool where they had been stored yearsfo8 1 materiare .Th l science examination demonstrated tha propertiee tth e th f so most part of the pellets (about 40%) of the enriched boron carbide from the absorbirc pins were quite satisfactor thed yan y were feasibl manufacturinr efo rodsw ne . e th f go The only problem we faced at the early stage of this manufacturing process was the presence of the induced pellet ^-activity. Differernt methods were used for decontaminatio e activitth t cellt reduces f pelletho no y Bu wa . a n sdi onl 8-1y yb 0 times .e operation Thuth l al s fabricatiod sro relatee th o nt d were carrien i t ou d schielded boxes. The finished rod also had y- activity but with certain precaution measure suitabls wa t puttinsi r efo t intg i reactore oth . This rod has been successfully operated for 416 effective days in the BOR-60 reacto afted ran discharges r thawa t i t t havexamined y no d an e an d t cells di ho t I .n d i noticeable damage absorbine th f so claddingn gpi , wrapper tube, welds, etcpellete .Th s in the majority of the absorbing pins were in good state, had no swelling and could be use fabricatior dfo similae th f no r safety rod. Thionls sywa partial solutio probleme th f no . Therefor RIAC SS R e continueeth d searching for options to use the enriched boron carbide from the spent safety rods. This resulte e developmenth n i d e flowsheeth f o e borot th tf no involvin e us e gth carbide pellets discharged from the spent safety rods (Fig.3).

The first option boroe provideth simple f no th r e carbidsefo us e pellets having the sufficient quality for fabrication of the new safety rods as it was done for the BOR- 60. Depending on the state of the rod the quantity of such pellets may range from 10 to 15%.

The second option provide mechanicae th r sfo l reprocessin pare f subth o tf go - standard pellet theid san r fragments. The millee yar powdeo dt r whic thes hi n pressed and sintered into new pellets. This option enables reprocessing practically of the whole core provided tha inducee th t d p-activit vert no y s y i high , otherwis complicatn ca t ei e furthemanufacturind an d r ro operation e th f go s wit. hit

221 Worke safetf dof d yro

Primary parting

Final parting , extracB f to

oui L-iJig <-u pej. j-euo * r i r itioned Non- conditioned Fragments Lets of BjjC little-active pellets of pellets

Disactivation of pellets

Grinding fine of B C Obtaining of pellets to powder, 4sift boric acid from boron carbide

Obtaining of C from boric 4 acid * Powder control C -B 4 powder Pressur pelletf eo , s sintering

Control of pellets , Absorbing ele- passport procedure pellets ment cladding

Control Making of absorbing elements, quility control Complement

Control Rod assembling, control

Prepared dro

FIG. 3. Technological scheme of the complex remarking of worked off fast neutron reactor safety rodsenrichedand boron carbide.

222 The third option provides for full chemical reprocessing of the core. This reprocessing technolog bees yha n deepl RIARC y boroe studieSS e .Th th n t dcarbida e through several stage transformes si d int boroe oth n acid whic thes hi n synthesized int boroe oth n carbide powde sintered an r pelletsafete w th d ne r int ye sfo rodsoth . thin I s cas materiae eth l gets completel radioactive th f o d yri e impuritiee th d san finished product has no ^-activity. During the investigations of the BN-600 safety rods it was established that the main ^-activity source were the technological microimpurities of Eu. The radiation intensity of the long-lived Eu and Eu isotopes was by several orders greater than that of other impurities (Fig.4). As this takes place, the correlation of the radiation intensity of these isotopes from burnup of pelletn i B practicalls 10 i s y absent. practicalls Ii t y impossibl o removet e radioactiveth impuritieu E e s providing decontaminatio differeny nb t solutions procese chemicae th th n I f . so l reprocessing the impurities are completely removed and as a result of this the ^-activity is lowered naturae th uo pt l level. RIAC engineerings SS R ha e Th possibilitie reprocessinr sfo enrichee th l al f gdo boron carbide built up in the storage pools. This makes it possible to empty the storages from spent rods, utilize wastes and to provide safety rods for such reactors as BOR-6 d BN-600an 0 e preliminarTh . y works suggest that this technologs i y economically efficient.

1500

152 154._ Eu Eu

100 200 300 400 500 DISTANCE FROM BOTTOM OF ELEMENT, mm

FIG. Gamma-radiation4. intensity isotopsofEu absorbingin elements of a lower link ofSR.

223 TABLE 1. MECHANICAL TEST RESULTS OF ANNUAL SPECIMENS OF ABSORBING ELEMENTS CLADING OF SR AFTER 101 efpd AND 331 efpd

Rod Link Absorb. Test. Ultimate 0,2%Yield Total Uniform and element temp. strength strength elongat elongat efpd region T°C MPa MPa % %

20 670 440 16,2 15,3 top middle 500 540 340 9,5 8,8 T-08 600 490 300 10,2 9,8

101 efpd 20 750 480 15,4 14,2 tort bo n bottom 500 580 380 9,8 9,2 600 550 360 8,8 8,2

20 670 470 11,7 10,7 top middle 500 280 260 5,2 3,1 T-02 500 570 410 9,2 7,6

331 efpd 20 880 700 9,0 8,0 bottom bottom 500 650 540 4,4 3,1 600 600 510 4,6 '"/ P,

224 EXPERIENCE IN PRODUCTION OF ARTICLES FROM BORON CARBIDE FOR FAST REACTOR CONTROL RODS

I.A. BAIRAMASHVILI Institute of Stable Isotopes Tbilisi, Georgia

Abstract

The paper presents a multi-stage process, including separation of boron - 10 and boron - 11, stable isotopes, production powder and the various configuration articles from enriched baron carbide.

INTRODUCTION Since 1964 our institute has intensively been conducting 10 works connected with B4C powder production by various methods. In 1989 the institute started the production of compact articles from boron carbide powder with various configurations producey db us to complete the absorbing rods for the control and protection systems /CPS nuclear /fo r power reactor fasn so t neutrons,(BR-10, BOR-60, BN-350 and BN-600). 10 Articles production fro mcomplea Bs 4 i Cmulti-stagd xan e process including the separation of B10 and B11 stable isotopes. e technologica e stepth th f o sl Al l proces e accompaniear s y b d respective control. The scientifi productiod an c n experience gained allowo t s draw some conclusion e advisabilitth n o s f researco y d an h development work fulfilment in two main directions: intensification of the technological processes and improvement of the economic index of production.

10 PRODUCTIOF BNO 4 C POWDERS Currently boron carbid maie th n s ei absorbe r material used S elementiCP n r fasfo st reactors. This positio s beeha n n strengthened as compared to 1983 with the development of nuclear power and relatively constant need in boron carbide containing a highly enriched boron-10 stable isotope. In this connection it is tim o analyst e s productione methodit th e r fo s t shoulI . e b d taken into consideration that boron isotopes separation technology includes the production of an elemental material as amorphous powder. Together with this other compound producee sar d as well which contain highly enriched boron-10 and boron-11 /the gas BF3, the salt KBF4, boric acid H3B03, etc./. In all the cases e initiath BFs i 3 l material thereby leavin a certaig n nuancn o e the technological proces elementaf o s l boron extractios it r o n remaking into other boron-containing compositions. I.G.Gverdtsiteli Institut f Stablo e e Isotope n 196i s 4 started to perform works on the production of powder and articles IO from B4C. The work at the very beginning was carried out in two 10 directions -the production of B4C articles by open hot pressing of an elemental boron-10 and carbon mixture as well as the 10 productiof o Bn4 C powdea metho r y magnesio-thermab rfo d l 10 reductio e sal th f potassiu o tf o n m tetrafluoroborate /K BFn 4i / the presence of carbon which is subsequently hot pressed.

225 10 e firsIth n t case e procesth , f o s B4C synthesins wa g combined with the process of the article formation. However the shortcomings of such a technology were soon revealed - the articles n instablformea d ha d e chemical content, differen i d non-uniform distribution along the boron, boron carbide and carbon samples section, high porosity, a tendency to crack- forming, the complexity of the formation process at high temperatures, etc. option A magnesio-thermaf no l metho basie madth s sdn wa eo of literature data on the possibility of boron carbide powder production with the nearest stoichiometric composition. The blend thermicit s undoubtedlwa y y allowed for. Accordine th o t g thermodynamic calculations it was necessary to perform reduction at elevated temperature r normafo se th l f carryino t ou g magnesiothermal process e briquetteTh .e reduceth f o sd blend undergo milling and chemical treatment to wash the products of the reduction reaction off the boron carbide. The basic technological characteristic wels sa othes la r matters concerning magnesiothermal boron carbide powder produced were determined in Technical conditions "Boron carbide labelled with boron-1d 0an boron-11 isotopes". 10 Furthe y b rusin e articleth g s from magnesiothermal B4C the control rods with a ring geometry for the reactor BN-350 / SR-TC- numbern i numbern 1 i 4- /C werT , e fabricate d testean d d during 250-350 effective days e d SR-TTh successfull. ro C y combine e functionth d a temperatur f o s e compensaton a d an d ro r emergency shutdown rod. The operation tests passed without any criticism on the construction. 10 When H3 B03 e instituteappearee th r_angth f o n ei d s ' :0 products a carbotherma, l methor B4fo Cd powder productios nwa developed blene methoA .th dr preparatiofo d solubilite th y nb y metho alss dwa o developed blene .Th d represent friablsa e powder- like product which subsequently undergoes carbothermal reduction ituba n e electric Raman furnace wit graphita h e heater. The work beine sar g conductin usiny gb spontaneouga s high- temperature synthesis /SHTS/ for the production of elemental 10 boron-10, powder and articles from B4C, other boron-containing compounds and composites on their basis. This perspective direction of our work is aimed at the development of the energy- saving SHTS technolog e productioth r fo yf boron-containin o n g materials. 10 Despit e e methodvarietth th e r f o fo B4ysC powder production adopter institutou t a d e currentlew y fabricate th e product e synthesiy meanth b s f o s s from elements. However some 10 alterations have been made. B4C powder synthesi carries si t dou separately with its subsequent formation under vacuum. This technology was borrowed from Moscow Plant of Polymetals. The synthesi preliminarila f so y briquetted blend /10B+C performes /i d in a vacuum induction furnace. After cooling the briquettes are milled up to the necessary dispersity, and the powder produced is t pressingho read r fo y .

10 PRODUCTIOF BO N4 C ARTICLES 10 B4C powders manufactured by us according to the above technological techniques undergo forming on the plants for vacuum hot pressing. These plants enable to achieve the necessary temperature for forming simultaneously applying press and overall effor 40to t plante p . Th tu heatee sar d wit hele graphithf th po e

226 resistance heaters consuming considerable amounts of electric power. Graphit moldina uses i es a d g tool. Allowin hige th h r fo g prices of this material and their great consumption it is urgent to prolon e life-timth g graphita f o e e molding tool. The most compactabl e powdeth s ri e from magnesiothermal boron carbide particlee , th probabl o t e s du yshape . The hot pressed articles are subjected to mechanical treatmen basie th cf o tproductione stage e whicth on f o s hi t .I is provided by using a diamond tool. Thermal treatment is also used for the stress relief and workability improvement. In fig.l 10 a pattern for the production of B4C powders and articles is presented. In 198e institutth 9 e starte e productioth d f articleo n s from boron carbide powder to complete the absorbing rods for fast reactor contro d protectioan l n systems manufacture y Moscob d w Plant of Polymetals.

SOME PROPERTIES

10 Thus there are various methods for B4C powder production at our institute which differ in the chemical content, dispersity particlee th , s shape, etc. froe anotheron m . In table 1 the characteristic chemical content of boron carbide powders produce variuoy db s method presentede ar s . BTotal, ctota c d free determinean ar le a chemica y b d l analysis. The remaining impurities were determined with the help of an emission-spectral analysis. The B4C phase formation is evaluate usiny b d g X-ray structural investigations.

10 TABL CHEMICA. E1 L CONTENF TBO 4 C POWDERS MA S%

Sample Btotai Ctotai Cfr« Mg Si Fe Ni AI Ça Cu Cr ™B designation enrichment %at

M agt n a em sio r he t

80/1-H 77.60 29.9 0.6 0.030 0.016 0.46 tr. tr. tr. tr. natural

82/1 77.70 22.6 0.5 0.026 0.110. tr 0.23 . tr 0.0 6. tr 84.5

83/1 77.75 22.9 0.8 0.038 0.020 0.23 tr. tr. 0.07 0.006 - 61.7

C a rbo thermal

8/K 76.0 23.4 0.6 0.040 0.018 0.25 0.12 0.0 . 3tr 0.02 0.15 natural

56/1 76.22 22.3 0.5 0.035 0.030 0.35 0.03 0.05 tr. 0.001 0.03 82.6

8/K-56-1 76.04 22.4 0.7 0.035 0.04 0.40 0.04 0.08 0.03 0.002 0.08 60.0

S y nth e size à

CK5-234 76.98 22.5 0.35 0.01 2. tr 0.0 0.01. tr - 18 . 0.0tr 4 natural

CK5-235 76.46 21.6 0.45 0.00 . 3tr 0.0 0.01. tr . 3tr 0 0.0 - 1 59.7

CK5-270 . 76.8 21.8 0.25 . tr 0.10.00 . tr 38 . tr 0.010 2 0. 0.6 80.0

227 By means of the SHTS we managed to produce elemental boron with the base material's content 98-99% mas with impurities: K- 0.3% mas and Mg-0.1% mas, boron carbide powder with the base material's content 98-99% mas at Btotal 77% mas and ctotal 22% mas. The powder from magnesiothermal boron carbide differs in the most profitable shape-spherical. The average dispersity of the powder when using such a method for the production is l.Oum. The carbothermal boron carbide powder is characterized by a needle-like /elongated/ shape whil e synthesizea th e s ha e on d faceted shape near to equiaxial. In both cases the average dispersity of the powders increases more than up to 55um. NEARESE WORKTH N I ST FUTURE The scientific and production experience gained on the

manufactur powdef o e d articlean r s froBCm allow o comt so t e

4 the conclusion on advisabilit10 y of fulfilling our future work in the two basic directions - intensification of the technological processes and improvement of the economical index of the production. To execute the first direction it is necessary to mechanize and automatize the production process for improvement of the quality and reliability of the products. Particularly: control of 10 B4C molding powder granulometric compositione ; th workin f o t gou technological condition pressint ho f so g under vacuu thermad man l treatment with the purpose of producing an equilibrium structure in the articles free from internal stresses; the study on the mechanis f appearanco m d behaviouan e f theso r e stresses during the formatio d thermaan n l treatmen e articles,etc.th f o t n a ; increas outpue th mechanizatio d n ei tan certaif no n technological operations /formin e articlesth f o g , fabricatio a moldin f o n g tool, assemblin moldsf o g , etc./. To execut secone eth d directio e followinnth g probleme sar of great importance: the development and introduction into the production permitting its functionation at relatively small expence powef so r resources moldin,a g tooothed lan r auxilliary materials n thiI . s connectio f high-frequenco a widne us e y facilities, deepening of the investigationson the SHTS, etc. are of great interest e SHT Th s distinguishe.i S e th s wela dy b l possibilit f boroo y n carbide alloying with titanium, zirconium and aluminiu necessaryf i m . OUR PERSPECTIVE knows Ii t n that boron carbide possesses unique properties for application in various fields of the industry and especially for nuclear engineering. It seems quite reslistuc to use this materia n separati l e construction f nucleao s r reactors which diffe n functioi r n e froanotheon m r usine attractivth g e properties of 10B and nB stable isotopes. As an example such important part f faso s t reactor namee b s heat-releasinn a dca s g elements, control and emergency shutdown rods, a side water wall /a reflector/, reactoa stee r fo l r body, etc. 10 Apparentl usage th y f Beo contron 4 i C l elements doe t settlsno e the matter. Widenin deepenind gan knowledge gth borof eo n carbide will promote the soonest and most effective solution of these problems.

228 10 We should try to attain to produce B4C free from impurities,molding powder witn a optimah l granulometric composition with maximally uniaxial particle size. The rich experience with various métallurgie fields shows that we n achievca e striking propertie f materialo s y meanb s f o s alloying. In particular the examples of favourable influence of alloying on serviceability under operating conditions of fast 10 reactors are known relative to B4C articles. There have been such results. speciaA l futurr placou n ei work occupie problee sth f mo 10 regeneration /refabrication workee mattee th f Th B f 4. dof /o C r is attractive, but it is connected with certain difficulties. We suppose thabasie th t c difficult worseninye th lie n i s g of the refabricated boron carbide quality, as 10B nucleus burn-up leads to 10B /nB nuclei ratio disturbance, a chemi-cal content alteratio n particula-i n n increasa r carbon i e n content, i.ea . change in the B/C ratio and therefore in stoichiometry. To achieve practically the necessary dispersity both in dimension and'i e shapth n e seems doubtful e residuaTh . l y~activite th f o y 10 refabricated material for its remaking into a high-quality B4C article /in a chemical content, grain size, density, etc./ seems very serious. Economic profi s alsi t o controversial because th e addition of the corresponding amount of fresh 10B into the fabricated materia s necessaryi l e y~th ace tivitf th I . f o y 10 samples of the refabricated B4C is acceptable we are ready to participat n suci e a worh k wite purposth h f high-qualito e y articles production. Technologic pattern of powders and articles 10 IO production from B and B4C

Separatiof no

Isotopes

Hydrausfeof Syrtfhesfeof Reduction of 10 salt BF+KF KBF4+Mg+C 1 K °BF4

Evaporation Electrolysis Chemicha! of molten salts treatment and 1 °KBF4+KF refining. 1800ÛC

Reduction of Productfcn Hot pressing

of elemental B4C powder boron 2150«: 10R'amorph

Hot pressing Synthesis 1850°C Diamond 10 1D B4C powder 4 B+C treatment,

'°B4C powder ameaSng

Diamond Hot pressing 1°B,C articles 1 treatment. of °B4C anneaäng powder 2150°C

10 B4C articles Diamond treatment, annealing

10 84C artistes

229 DIVING-BEL DOUBLE-VENTED LAN CONTROC 4 DN B PI D LRO LI SHIKUN, Z. CHANGSHAN, XU YINGXIAN CIAE, China

Presented VoznesenskiR. by

Abstract

In this paper a conceptual design of vented device for absorber pins with sodium anbondins dga describeds gi relateA . d analytical calculation experimentad san l programme are presented.

I . Introduction

For fast reactors, up to the present only B4C is extensively used as an actual and potential neutron absorber of the control rod pins. But the awkward problems

accompanied with the large yields of helium and lithium tn due to the burnup of boron-10 (3.72 STD cc helium per 1020 atomic burnup of boron-10) are the significant B..C pellet swellin d heliuan g m releas n ordeI eo attait r n high burnu d an long-lifp e pine ,th B

Vente. I d Phenomenalized Model

In our conceptual structure (see section IV) , the helium released into the bell when control rod pin is in work can only be discharged by the vented pore in individual bubbles eventually The phenomena 1ized model governing this process may be described as follows.

231 1) Surface Tension—Capillarity Phenomena1183 The Neccessary Conditior nfo The Initial Bubble Formation Accordin e phenomenth o t g f surfaco a e tension s weli t i l , know a bubbln r thafo et with radiu e e sameliquith th sr ,n i (o Td dropleI e steam)ti n e i forcestth , ' balance equation at the interface is

4 « rï(Pl-p2) =- ) .( rn 4cr f dr.

(1)

where PI, p2 are inner pressure of bubble (or droplet) and the outer pressure of liquid (or steam) respectively; ° is the surface tension of the liquid. For a circluar capillary with radius r; (or diameter di) , the free interface in static state betwee o immiscibltw n e fluid media withi e capillarth n y (suc s a heliuh m and sodium) is a spherical meniscus with curvature radius TI/COS 9 , and the balance equation relativ e e forceinterfacth th o t t a se e becomes

(P.- Pi) {n/cos 6)a 2= — —— - ( /cos6)2 s co / » r d( ri

- Pi = cos /it (2)

where pB. fi are gas and sodium static pressures, 6 is the wet angle of the liquid to the wal n capillaryi l n awkwara , d delicatan d e parameter, whic s veri h y difficule b o t t measure d usedan d . If the meniscus is not a spherical, this equation is written as

PB- PI= (I/ îCOaO ) S8 (2')

with TI , r2 being the two principal radii of curvature of the meniscus. When vent condition is reached , the bubble is formed suddenly on the vented end sruface of the

capillary with initial diameter db>d n generali i , comparinSo . n gd (2)ca an eqs e , w ) .(1 understand easily thae neccessarth t y condition (the minimum pressure differencer fo ) the bubble formatios i n

PB- p>>2 a / r, (3)

This conditio s i t independennangl we 6 obviouslye e th f o t . 2) The Bubble Growth—— Dynamic Balance of The Forces For this phas f bubblo e e growth e phenomenalizeth , d model show n Fig.i n s i used1 .

232 a) Kinetic Energy (Expansive Work) K and Power K' of The Bubble I ne definitioth ligh f o t f kinetio n c energd an y f momentuo w la e m th conservatio s wel a. s nnotina l e th g 191 conceptio f virturo n e mass (M=l. 5M0 when db»di,

M=l. 688M0 when db«di that can only be encountered while the bubble jet is in very high frequence), the following equation e writteb n , ca s the n1 froyg Fi m are

Fig1 Boubl. e Growth e Pori'Th n e iF^bd'bP ) 32 / 2* 3 ( = K (4) Immersing in The Sodium Flow

2 K' = (3 * / l«) p .F.dgd'b b+l. 5d'b } (5)

and

(po) Kfl-F'= ) ( /2 a- P. 2(d;)]d?db (6)

where P, is the density of sodium; d' =—(d ), d =——j(d ); F is the modification for b dt b b dt b t

non-fully free motion of the bubble, F<1; p is the pressure on the outer interface L 0 2 1 1 bubblee oth f 2 (dF ;b )= (3/4 ) (db) - in's {ta5 . (d( n[0 b ) -')]}{!- (1/3) (dS) in' s -'5 ta. n[0

(tdb) -Mllwith db = db/d0, and d0>df . e havw , e ) (6 Fro d m an eqs ) (5 .

1 2 . 372 P .FJl- Futdb))- (dbdb+l. 5d'b ) (7)

e pressurth s i b ewhilp withi b =po+, e b pressurp e ed bubblth th / r n 4 o (o n eo innee r interfac f bubblo e e with pressure field withi e bubblth n e being neglected). e RelationshiTh b) r Dynamipfo c Balanc e ForceTh f seo At laste dynamith , c balance equatio e forceth f s i o ns

2 p«=pi + 0. 372 P ,F! [1- F2(db)]-' (dbd;-f 1. 5d'b ) +4 o /d„+E A p (8)

where 2Ap is the total pressure loss for the gas flow, and the following dynamic s satisfiedi rela n tio :

(9) PK>Pb>Po>Pl

Equations (8) and (9) are in accordance with equation (3). 3) Bubble Detachmen — Critica— t l Bouble Diameter d^ Here the diffusion loss is neglected. Then a dynamic bubble just previous to detach the wall encounter a svariet f forceo y s including surface tension force, bouyancy force, drag force e previou, th e wakdra f th o e go t sforc bubblee du e , momentum fluf o x

233 s phasethga e , added mass inertial force' e treatmenTh 91 f balanco t e relatio r thesfo n e forces is more complication, but in our interest range, merely the balance of the former three terms is appr innate iy used, i e

2 [l-F2(

d;2 (d;t F dc= )bu b= c/d J substitute s ; i wher0 (d 2 F 9 e 81(m/sd r bg= ,s fo di d )C , 2 viscous drag coefficient equa o 24/Rt l 8 4/Ree1 0 (Re, 4

v 6 e velocitar d v viscosit (Re>500an yd an = vd f v sodiue o R by c ), / m respectively, F3 1 (d; 0 5(d c"Mtan[ )) = b = 0 OBin" ((d^)' } s MdJ definitei bc s a criticad l bubble diamete r detachmenfo r t from wall surface

IU . Analysis of Hermetical Characteristics ——— Model of Helium Leakage / Sodium Permeability.

The vital function of double vented device is hermetizat ion of the pore through a sealing sodiu e mbel th level n i whill e helium sourc s i absene t Heliu / msodiu m reversal leakag s causei e y solub d b / diffusibi11t11 y 1t f heliuo e y th n i sodium d an m hydraulic fluctuation of sodium coolant 1) Solubilit Heliuf o y Hign i m h Temperature Sodium Ther e lot ar f eexperimentao s d theoreticaan l le th research r e Henry'fo th w n o sla s helium in sodium, but only the typical three are cited here e ExpressionTh ) a Henry'f so Determinew sla Experimeny b d t o presentt p U »e mosth , t accepted expressio r f Henryso heliufo n w n i sodiula 'm m

x lnKH = 4 247- 6523T (11)

2 I in which KH[at - f r (N m" ) ]=CO/PH«, c0 is the solubility of helium in sodium, pS« is partial pressure of helium in gas phase , T=605K-850K f date o othe Th t a • se rcome s fro ms expresse i [12 d ]an s a d

logK„[mo 3595T- 4 l3 '2 ' mol' - ' m ]= at 1 (12)

wher 573= T e - 873K e agreemenTh t (particullar t a higy h temperature e tablse ) ,1 betweee e th n determinations of these two sets of data is supprisingly good in view of the disparate techniques adopted ) Phenomenalizeb d Theoretical Model Henry's law can also be derived from statistical thermodynamics It is well known that the vital procedure for the method of thermodynamics is always to find the 112 131 Hemholtz free energy F first To find KH makes no exception One of the results is

234 K„= (Vf/RT)exp(AF^f/RT) (13)

By the definition of free energy and further by the assumption that the entire process of dissolution consist of three individual steps: digging hole, diving gas atom into the hole, and raising the gas atom to oscitation as liquid atom in the solution, we have

= U«r. c- ,. .-TS..

and

KH= (Vr/RT)exp[- (U.c. <=+U8f. , U« /RT+lf«+ ) .. . ./RJ (14)

in which Vf is molar volume of sodium in solution; Ar^r is the molar Helmholtz free energ f heliuo y n solutioni m e reversiblth , e molar work require o introduct d e atomth e s

s inte solutioga th of o f concentratioo n n c o e small0s e solutio th b (cs i 0 , n ca n if\ff regarded as pure sodium); Uef, <= is the molar internal energy required for forming a e largesth s atom s ga valus i it e ,s t plu i th d e e esizliqui an th th f •hols o en f i o ed n magnitudei e on e accompanieth , d entrop s knowa i very s a yn little plut i s valut bu e s***

is very difficult to calculate, so is neglected; .U « is the molar inner energy gr required to raise the 'quasi-rigid sphere' introduced to necessary potential ( i. e.

introduced to the hole) , its value is minus and intermediate in magnitude; UK. « is the molar internal oscillatio s atoga n me energafteth f o ry introduced inte holeth o ,

its value is plus and intermediate in magnitude; ASg = S. „ is the molar entropy of f e oscillation of the gas in solution, the only entropy being considered, it is a ^** f*^ **•* S** f*J f** significant plus value. Vf, Uer. c> Uef. „, UK. ., SB. ., hence A F'Kf are temperature -dependent. This mode s mori l e easily understood t thaverye n yothers't no t 1bu '", precise in its parameters, there are a certain margins to be improved. e quantitTh y comparison between these three model s showi s n tabli n. 1 e

Table 1 : The quantity comparison of these three models

T (K) 623 673 723 773 823 873

eq.Ül) 1.96XKT8 4. 27X10'8 8. 35XKT8 1.50XKT7 2. 50X10-7 3. 93X1Q-7 KH=CO/P£. eq.(12) 7.75XKT9 2. 08X1(T8 4. 87X1Q-8 1.02XJO-7 1.96X1Q-7 3.48XKT7 ) tm a ( , pH eq.(14) 2. 48X1CT8 1.50XKT7 7. 30X1Q-7

2) Diffusion Coefficient of Helium in High Temperature Sodium

235 There are several different argumentst12' 14> 1S1 on the diffusibi l ity in liquid solutionse rigorouth d t generalizean bu ,s d mode s i difficull o t findt . Here o typicatw l model e quotedar s . Classicae aTh ) l A HydrauliMode — — l c Model With potential flow as its prerequisite, thi s model is first derived by Stokes- Einstein for the movement of a not too small spherical particle through the continuum (liquid) . - Stoke — Einsteis n Formationtlo] Whe a spherican l particle move n fluii sy b unid t force e originath , l irregular Brownian motion becomes an oriented one named as mobility. This mobility is expressed as

Ui = l/AnrT) (15)

in which e A=6particlth s , ri e fluieth radiuss di viscosityI T , . - Use in The Regular Diffusion Field of Solutions When diffusion law is applied to the field of chemical potentia1 / concentrât ion of the solution, the diffusion coefficient D in one dimensional X can be derived as

n = n ° + kTinc

u = u . {- d n / dX) = - u i (kT / c) (dc / dX)

) (ddX T , k u e / - = e= u) i (dD dX e/ - J=

At last we obtain

D=kT/A*m (16)

where A is approximately constant (A=1.4—6 dependent on the relative size of the particle to the continuum's one and the slip degree or the chemical activity ( or fugacity) for the diffusion, etct153); r is particle radius; TI is viscosity of solvent. k=1. 381X10"23J/K, M. (and n °) is chamical potential of the particle in solution. Eqs. (15) and (16) (that we may call as the law of first-order power of temperature for diffusion) are extensively used in many fields with fine accuracy. Some data of diffus ion/self-diffu r sodiumfo n sio , lead sodiuman d / lead system ,e usefub whic y ma lh in the technology of fast reactors, are availablell*' 173, we can find that the eq. (17) is applicabl n engineerini e g sens e self-diffusioeth ever fo n f sodiuo n n whice i m th h pair potential is broad and shallow. But for the component with intense volatility (the intensely spontaneous and irreversible process) in the solution such s heliua n i sodiumm s diffusioit , n (the particle velocity s i evidentl) o s fasty , that this model of potential flow should not be used. b) The Relationship From [12] n experimentaa Thi s i s l relationship havin o n paralleg l anywher d perhapan e s

236 agreeable to the thermodynamic principle, so it is the one that can only be used as an expedient at present, and is approximately expressed as

logD=3 3051T- .77 1 (17)

Compar d analyeizan e e eqs. (12) thr ough (18) as we II the diffusions of the components of solid d solutionsan n s ca e on , learn a good deal of interest instruction e analogs th suc s a hy betwee e vocancth n y theory118' e holf solidth o ed theoran s t a y liquids e irreversiblth , / e others «dissolve A/v = 70cm reversible relation between dissolution and diffusion, etc. Fig. 2 shows the significant h (diss8 ) . difference between the processes 0 100 200 400 s (degau) of dissolution and degassifica- Fig. 2 Comparison between Solution and Degassification tion of the helium / sodium in He/Na and Ar / Na Systems syste s wele Argona m th l / sodium system imitated from the report [12j ) Hydrauli3 c Fluctuatio Sodiuf o n m Coolant116-19-20-211 The hydraulic fluctuatuon (V or p) within a channel of the core assembly, which in turn is feedback to the pore and cause sodium flow in it, may be caused by the periodic suction of a venduri flow, by the wake of a vortex, by the impulse of an impingin e causeb n tur y i y t pumpin(suc b ma dnt je g je h g puls r o vortee x formatio/ n shedding), or by the motion of contro 1-rod-se If, etc and their secondary effects. The periodic suction of the venduri flow comes from random displacement or vibration of the control rod or rod pin. The impulse of the vortex formation / shedding and wake is the characteristics of the turblent flow, and the maximal one with AV equal to the

velocity of the largest edding and f equaling to V0/l (V0 is the average velocity of the coolant sodium flow 1 correspond, e characteristith o t s c dimensio f thio n s edding). e pumpinTh g pulse comes froe rotatioth m f pumo n p blades wit X rpm f equan hn o , t l being the number of blades. | Letting V = V0 + Va, p = p0- -pi, the Navier- Stokes equation and the continuity equation can be written into two counterparts, one presents the average flow, the other presents the fluctuation (perturbation) flow. The average flow is

(Vograd) V0= -grad(p0/ p ,) + v A V0 divV0= (18)

237 e anfluctuatioth d n flos i w (dV(V+ ) dt grad)V/ (V+ grad)V = -grad{ VA /p,v + ) 0a a 1 0 plx 1 > (19)

The boundary conditio s tha i a nV tvanis d V n 0an fixe o h d solid surface. Evidently, e resolveb t (19no . eq n d ) n ca i becaus . Va.p, , e they have complex e spectrumsth d an , method of noise analysis or measurement must be used. In our question, what we most concern is the effective amplitude and the frequence can be regardless. This effective amplitude V* is expressed as

fV= [V-V= s 2 -Vf]a / V o= 1/2 (20)

i. e. is the standard deviation or rms of the V0. Then the velocity u s. in the pore is resolved from following equation:

~-v.^7 2.6 67V . 0 ?= (21)

s dynamiwheri v e c viscosit f sodiumo ys capillari L , y length.. 4) Helium loss Rate in Plenum (Bell) by Diffusion and Hydraulic Fluctuation e founb Thiy n resolvinb dca s e continuitth g y equatio e diffusioth f o n n component of the sol u t ion.Thi s equation is (her the reversal diffusion of argon being neglected

conservately for the KH and D of argon in sodium being much less than those of helium in sodium)

j f v P i c dv = - f P i eu df- f Jdf dt t

-div) u Jic P d( - iv = d t

where PI is sodium density in at./cc , J is diffusion flux at. /(cm- s)

r questionou n I , following condition e satisfiedar s : 2 d • Incompressiv ediv(p+ i fluidP whicn — i , u—- h )=0, t t dt

• J= - P !Û [grade + kTgradT + kpgradp] = - P iDgradc, • One dimension and only steady state being considered,

t u •s=con d havin J directio an e . st th g e presumeb n d conservatively, d lettinan g D(c)=const. Then, this equatios solution it e writteb d n an nca sn easil: as y

238 u t c = 0

(22)

X=0, C=C0; X = L„rr (L«rr

Ü-exp[= c t (L,-u rr/D]) - x L.rr/Dexp( - e u }[l - ] ^ (23)

2 1 1 (Het cm"• [a ) s"• )] 0 c t u i -P (24)

Q= * d2J/4=AJ [at (He) - s'1] (25)

in whice s porelengti th L hf . o h e seeb n Ica easilt y that when D-*-0, Q-» P iUtAc- e los0th , s rate cause y b purd e

hydraulic fluctuation; whe «u .n -»-O , Q-* DP iAc 0/ L„ e losfr th , s rate cause y b purd e diffusion of c.

I. VConceptua l Design

1) Design Parameters

a conceptua s A l design owinI l ,fo g parameter e conar s s idered laddin:c g diameters D0/Dj.

= 15/13(mm), diamete f BCo r pellet D=12. 3mm, pellet , averagstaccm 1 5 k e burnuf ï o p 4

4Xl0cm", volumetric swelling rat f BCo e r 101.5cnTpe % Bburnu, operatiof o p n 10

3 21 3 10

temperatur21 e 800K, minimal shutdown temperature 500K, pressure 0.2MPl4MPa. 0 / ,a pins pitch 16~16. 2mm (with wire wraps as spacer) . 2) Vent Calculation In this area, only the detached bubble diameter is concerned.So, the calculation is carrie t accedin ou e (11). dresult Th eq . o e t listegar se dissolution tablTh i d . 2 e s i n

Table 2 Evaluation of equilibrium vent in full power

Ur6 . n 13 d i= di=48. 1 urn di = 0. 1mm di=0. 2mm di =0. 3nmo v (m/s) m m c db f s-m 1n c db f s'1 dbe mm f s'1 c midb B f s'm 1m c db f s'1

0 >1. 16 0.023 >1. 76 0.057 >2. 25 0.024 58 . >2 0.01 >3. 25 0.008

0. 10 >1. 03 0.3 >1.65 0.07 >2. 14 0.03 >2. 72 0.013 >3. 13 0.009

0.25 >0. 74 0.8 >1. 28 0. 15 >1. 72 0.06 >2. 28 0.022 >2. 68 0.015

0.50 2 4 . >0 4.3 >0. 77 0.67 >1.1 0.24 >1.51 0.075 >1.83 0.052

0.75 / / >0. 53 2.2 >0. 76 0.75 >1. 07 0.19 >1. 28 0.17

1.00 / / >0. 39 5.4 >0. 58 1.7 >0. 81 0.45 >1.0 0.30

1.25 / / >0. 32 9.7 >0.46 3.2 >0. 62 0.92 >0. 81 0.61

239 conservately neglected in the calculation of bubble diameter for this process is relatively slower than the total process of bubble formation-growth-detachment in genera e maximaTh l l bubble diameter e than subchanne th e ca tcontroth d f ro o ll accomodate but can not be in contact with neighbour pins is determined by the pitch of c mus db e les b te thsth etha , pin 3 25mmSo n d sthi an ,e desig s th siz s i ne limitation 3) Evaluation of Hermetical Characteristics Using equations (23e relevan th o (25t ) d an )t parameters e hermeticath , l function (the helium loss rat y diffub e / hydrauli n sio c fluctuation while sourc s i absente n ca ) e executeb e resultTh d e listear s n tabli d 3 e

e evaluatioTablTh 3 e f heliuo n m loss rate"

di = 0.03cm A= 7. 07X 10~ cm L=0.1cm L,ff

PS. at ml"12 62X1010 2 33X1019 2 15X1019 2 OOXlO19 1 875X1019 1 811X 1019

ml"• Pt ia .1 10X 230 22 2 245X1022 2 20X10 2102X 216 22 2 11X102 102X 08 222

1 8 8 8 7 7 7 - mol l mo "c0 1 03X10" 3 87X10" 8 44X1CT 1.65X 10" 2 97X10" 3 93X10"

u t cm - s" r 2 2 2 2 2 2

t a Qs" 1 3.45X 1011 1 40X1012 4 50X101 102X 38 113 4 10X10 1103X 10 173

1 s" • t a Q' 3. 25X1011 1 23X1012 2 63X1012 5 03X10 1102X 190 2 8 1 20X 1013

Q' at • s"1 4. OOX1010 5 70X1011 . OOXlO3 12 1. 10X1013 3 65X 10JS 6 52X 1013

t day • ml'1 / 197 55.5 16 8 5 29 2 95

y -ml"da t* 1 / / 94 7 45. 9 24 4 17 5

- ml" y da 1 " t / / 84 21 5 94 3 21

Q , t corresponds to diffus ion / hydrau1ic effect, Q' , t' corresponds to pure hydraulic effect, Q" , t' corresponds ot pure diffusion effect

From this table one can see that to prelect the helium from leakage in high sodium tempratur n awkwara s i e d matte n thiI rs case, QccA/L.rr e morth ,e effective methoo t d

decrease leakage is the reduction of the d4 4) Conceptual Design of The Diving-Bell e sketchTh e b y n structurca 3 e DBDVt i th g Fros show fi i f B o , men i n thi g Fi s see s plenunga thae mth t consist o organitw f o s c parts wit a h capillary link between them, the filters in the bell is used to protect the particle from escape and to calm e sodiuth me plenuleveth r protecn fo i ml e capillarth t y from sodium sneakinn i g Before commissioning e plenuth ,s fillewa m d wit e hvente th argo n advanci dnd an e pors pluggewa e d with eutectic alloy.After commissioning e eutectith , c alloy melt out automatically and the device is put into operation, For a successful and

240 dutiful design e e capillarth th , f o y plenum must be protected from the cap i l lary surreptitious intrusio y sodiub n f inneo m r or outer be 11.Therefore, for the volume Vi of the inner bell, the effective accumulated swelling of the B inne1 1 e b r sodium sneak-in) during the designed a N -leve l duration with no source present, a design holddown spring margin Voa for the helium volume o shrinkagt maxima e / du le minima l p ga a N temperatur epressur/ e differences between JIMS B4C pellet e regimeth f rato d s reloaan e d operation must be stipulated Fig. 3 The Sketchy DBDVD Structure Basing on these requirements, for our conceptual design e followinth , g relations e organith f o c plenum components shoule b d satisfied:

V,+V0

V,+Vol-t-V0i (26) \ 35V> ! (27)

Here the referential values are arbitrarily taken: Vi = 5cc, V01 = 4cc, V02 = 7.5cc, Vt 16 5cc The Sketchy structure of DBDVD is shown in Fig 3

V . The Hermetical Test Rig

As mentioned above, the referential detail information or theoretical / experimental e DBDV modeth e nowhere publi e ar f Dfounth b e o l n o th i ct d e f literatureso l al , arguments in this paper can not but be based merely on the expedient scientific principles Obviously, there are lots of works to be done to convert this conceptual design e inte reabaddlth th ol t oneybu ,neccessar y affai e t a presenth r s i t demonstration of the hermetical characteristics of the vented pore, especially the diffusion coefficient D of the helium in sodium solution at our particular conditions s i designe e Nowg tesd th ri ,fabricatean t d ds i show n i Figg Thinri s4 . diagranimatically. Here onle maith y n point e illustratear s s followsa d :

241 Test objects: Verifying the hermetical characteristics of the vented pore, hence the conceptual design, some referential results be pinned hopes on; Test arrangements: T=230~600' t stead(a C y state) with 50~100°~ C 06 . 0 steps = p , 0. iMPa; Main technologic points: e treatmentTh • leac inesf n1 o s s/ degassifica d sealinan n tio e g systeth degre f o m e satisfying test requirements; e purificatioTh • f sodium/Ao n r satisfying test requirements (0<10ppm < lOppmH , , initially); e sodiuth f o m tank/Ap T/ • r tank adjustable; • Several pairs of thermocouple to measure the sodium level in the helium can with a siren to alarm the sodium lever. The shortcoming of this design which will affect the applicability of the test result d whican s h shoul e takeb d n int t o imitatno accoun e n sodiuth ca s tha i et t i t m colant flowr thifo , st possesseye imitatio t no t s presenta di n . Amplifier output P<>wer supply stop valve manome ter

pressure transducer

.Fig.4 The Diagrammatical Rig Scheme for The Hermetical Test

This paper and the arguments are prepared by Li shikun, the design of the rig and test is entrusted to Zhou changshan with Xu yingxian participation. We would like to aknowledge Wang Xiaoron s elaborathi r fo g e edir u zaixianJ he t ing/pri r d fo an gg in nt . Ï e l b ta calculâ f o n o ti

242 REFERENCES

1 CEA-CONF-6655 r 198Ap , 3 2 CEFR Primary Desig B Chinese(I n , )199 5 3 IAEA Report IWGFR/48, OBNINSK, 7-10, June, 1983 4 HEDL-SA-1995-FP, 1980 5 C W Hollenberg, L J Jackson and J A Basmajan, Nuclear Technology, vol 49, June 1980, p93-101 H Kashihar, Fasd l an t a M M ReactoOakdet l e a a t 6 e nr d CorFuean e l Structure Behaviour, Procedings (1990, Inverness) / org by BNES London p33-39, 41-45 7 British Nuclear Energy Society Fast Reactor Power Station, Proceedinge th f o s international conference, 11-14, War, 1974, London 8 £tî^, fcÏÏ^ un Chinese) , 1955 9 S C Chuang & V W Goldschmidt, J of Basic Engg , Dec 1970, p705-711 10 E Veleekis et al , ANL-7802, 1971 ü H Borgstel11 d Mathewsan d , Applied Chemistr e AllkalTh f o y i Metals, 19878 h c , 12 KFK-116 r EURTNR-785o 6 , 1970 13 Klaus Thormeier, Nucl Engg and Des 14(1970), p69-82 14 Sylva Z , nLiqui Beer d e ,d Metal ( chemistrs d physican y , 197 ) s 2 p250-l, 514-525 15 Tyrrei, Diffusion and Heat Flow in Liuquids, 1961 16 L D Landaul and E M Lifshitz 2nd ed Translated by J B Sykes and W H Retd, Fluid Mechanics {course of theoretical physics vol 6), p58-61, et al 17 Natrium (in Russian) , 1992 18 Donald R Olander, Fundamental Aspects of Nuclear Reactor Fuel Elements , 1976 pl-81l a t e , 19 L S Torig S, J Weismaa, Thermal Analysis of Pressurized Water Reactors, 2nd ed 1979, p209-216l a t e , 20 Joseph A Thie, Power Reacotr Noise, 1981, p!7-20, et al 21 Robert P Benedict, Fundamentals of Pipe Flow, 1980

243 LIS PARTICIPANTF TO S

Alexandrov, I. Experimental Machine Building Bureau (OKBM) Burnakovski Proyez5 d1 603 603 Nizhny Novgorod - 74, Russian Federation

Baldov. A , MAEC, NPP BN-350 466200 Aktau Kazakhstan

Bairamashvili. I , Institut f Stableo e Isotopes 21, Kavtaradze str. 380086, Tbilisi, Georgia

Babu. R , Indira Gandhi Centr Atomir eFo c Research Departmen f Atomio t c Energy Kalpakkam 603 102, Tamil Nadu, India

Bukshau Y , Institut Physicf eo Powed san r Engineering, IPPE Bondarenkl . oSq Obninsk, Kaluga Region 249020 Russian Federation

Chernyshov. V , Moscow Factory Polymetals Kashirshoy Strass9 4 e- Moscow 115409, Russian Federation

Edelmann, N. Forschungszentrum Karlsruhe Postfach 3640 76021 Karlsruhe, Germany

Favet, D. FRAMATOME Direction NOVATOME 10 rue Juliette Récamier, BP 3087 69456 Lyon Cede6 x0 France

Ivanov, A. Institute of Physics and Power Engineering, IPPE 1 Bondarenko Sq., Obninsk Kaluga Region, 249020 Russian Federation

245 Kaito, T. O-arai-Engineering Center Power Reactor and Nuclear Fuel Development Corporation (PNC) 4002, Narita-cho, O-arai-machi Higashi-ibaraki-gun, Ibaraki-ken 311-13 Japan

Kryger, B. CEA, Centre d'Etude Saclae sd y AMT/SEMI/LEMA 91191 Gif-sur-Yvette Cedex France

Matveev, V. Institut f Physiceo Powed san r Engineering, IPPE l Bondarenko Sq., Obninsk, Kaluga Region 249020 Russian Federation

Okada, K. Mitsubishi Heavy Industries, Ltd. Advanced Reactor Engineering Department 3-1 Minatomirai 3-chôme Nishi-ku, Yokohama 220-84 Japan

Poplavski, V. Institute of Physics and Power Engineering, IPPE Bondarenk , Obninsl . oSq k Kaluga Region 249020 Russian Federation

Ponomarenko, V. Moscow Factory Polymetals Kahshirskoy Strass9 4 e- Moscow 115409 Russian Federation

Rudenko. V , Institute of Physics and Power Engineering, IPPE Bondarenkl . oSq Obninsk, Kaluga Region 249020 Russian Federation

Rogov, V. Experimental machine Building Bureau (OKBM) Burnakovski Proyez5 d1 603 603 Nizhny Novgorod - 74 Russian Federation

Roslyakov, V. NPP BN-600 Beloyarsk 624 051 Russian Federation

246 Rinejski, A. IAEA, Division of Nuclear Power (Scientific Secretary) Wagramerstrasse - 15 0 10 P.Ox Bo . A-1400 Vienna, Austria

Risovany. V , V.I. Lenin Research Institute of Atomic Reactors (RIAR) Dimitrovgrad, 43 35 10 Russian Federation

Shabalin. A , Experimental Machine Building Bureau (OKBM) Burnakovski Proyezd 15 Nizhn603 360 y Novgoro4 7 d- Russian Federation

Tarasikov. V , Institut f Physiceo Powed san r Engineering, IPPE Bondarenko Sq. 1, Obninsk Kaluga Region 249020 Russian Federation

Truffert, J. CEA/DRN/DEX/SDC Bat 315, CE-Cadarache 13108 Saint Pau Durancs le l e France

Voznesenski. R , Institut f Physiceo Powed san r Engineering, IPPE Bondarenko Sq. 1, Obninsk Kaluga Region 249020 Russian Federation

Zakharov. A , V.I. Lenin Research Institut f Atomieo c Reactors (RIAR) Dimitrovgrad, 433510 Russian Federation

247