Status of LMFBR reheat in Western Europe, 1972 : report of the United States of America LMFBR Sodium Reheat Team visit to France, Germany (FRG), Netherlands, and United Kingdom, May 22-June 6, 1972

550 Pages·1999·37.676 MB·English

by  Max Weber, U.S. Atomic Energy Commission. Division of Reactor Development and Technology

XA9949996
IAEA-TECDOC-1083
Status of liquid metal cooled
fast reactor technology
^yi^p
INTERNATIONAL ATOMIC ENERGY AGENCY ft
3 0 - 22 April 1999
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STATF ULOSIQ UID METAL COOLED FAST REACTOR TECHNOLOGY
IAEA, VIENNA, 1999
IAEA-TECDCC-1083
ISSN 1011-4289
©IAEA, 1999
Pre inIhAtt eyEbd nA Ai ustna
April 1999
FOREWORD
In 1985 the International Atomic Energy Agency published a report entitled "Status of
Liquid Metal Cooled Fast Breeder Reactors" (Technical Reports Series No. 246). It was a
general reviee whts fot atusf o fast reactor development ta that time, covering some aspectsfo
designd na operationd na reviewing experience frome ht earliest days.t I summarizedeht
programmes and plans in all countries which were pursuing the development of fast reactors.
Durin eghtp eriod 1985-1998, there have been substantial advancen si fast reactor
technology development. Chief among these has been the demonstration of reliable operation by
several prototypes dna experimental reactors,e ht reliable operationf o fuela ta high bumup (BN-
600, BN-350, Phenix, PFR BOR-60, FFTF). Some additional countries, such as China, India and
the Repubf lKoic orea have launw cfhaesentd reactor prograe mIhAtmE teAAs .meetinnogs
liquid metal cooled fast reactor (LMFR) technology,t i became evident that there have been
significant technological advances as well as changes in the economic and regulatory
environment since 1985. Therefore, the International Working Group on Fast Reactors (IWGFR)
has recommended the preparation of a new status report on fast reactor technology.
The present status report inte onpdtrso vide comprehed dnnesitavaie led informnatoion
LMFR technoloe nhgfopToyrc . sauics tical issues te hruaast eo futel ngineers, scientists,
managers, university students and professors, on the following topics: experience in construction
and operation, reactor phd ysnsaaifcest y, core structural mad tfenuraeial ltechnology, fast
reactor engineering and activities in progress on LMFR plants.
This report has been prepared with contributions from China, France, Germany, India,
Japan, the Russian Federation, the United Kingdom and the United States of America. The
responsible IAEA officer was A. Rinejski of the Division of Nuclear Power.
The IAEA expresses its appreciation to all those who have participated in the preparation
of this reference compilation and also to the Member States that have made available experts to
asd spnisaatr ticipan tteih is work.
EDITORIAL NOTE
In preparing this publication for press, staff of the IAEA have made up the pages from the original
manuscripts). The views expressed do not necessarily reflect those of the IAEA, the governments of the
nominating Member States or the nominating organizations.
Through eohtutetx t namf oeMs ember Stae rtreaes tains eatdh ey were whe shetatenw xt
compiled.
The seuo f particular designations of countrier sot erritories doet onsi mpy jlnayu dgemeehtn ybt
e lhegta pl ousbttf ale osi tsuIsuhAhcs ateEhrA ,c ,ounr ttroeierrs itof roitehse ,ir authdonritaies
instie tduhetir ltooiom nfs itation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as registered)
doet osni mpy lniyan tention oti nfringe proprietary rightr sosn, houe lbcd tio nstrun eae dsan dorsement
r roecommende pahatt irnoe toInh A otfE A.
CONTENTS
CHAPT. 1ER BACKGD ROOVNUEANRDV IEW..........................................................................................1... .
1.1. Progress within the period 1985-1998................................................................................................................ 1
1.1.1. Demonstration reactor operation.............................................................................................................1
1.1.2. Prototype reactor operation....................................................................................................................1..
1.1.3. Technical achievements...........................................................................................................................2
1.1.4. Design advances....................................................................................................................................... 3
1.2. Contenf ttosh is report.........................................................................................................................................4.
CHAPTER 2. OPERATION EXPERIENCE WITH PROTOTYPE AND
DEMONSTRATION LMFRs........................................................................................................... 5
2.1. BN-350 operating experience.............................................................................................................................. 5
2.1.1. Design features......................................................................................................................................... 5
2.1.2. Operating experience...............................................................................................................................7
2.1.3. Safety enhancement and equipment life extension................................................................................ 21
2.1.4. Review of experimental programme...................................................................................................... 22
2.2. Phenix operating experience..............................................................................................................................23
2.2.1. Design features.......................................................................................................................................23
2.2.2. Operating experience.............................................................................................................................25
2.2.3. Statutory inspection - Spring 1989........................................................................................................ 26
2.2.4. Inspections and maintenance during the test period (1991-1993)........................................................ 28
2.2.5. Negative reactivity shutdown.................................................................................................................29
2.2.6. Plant statistics........................................................................................................................................92.
2.2.7. Radiological safety................................................................................................................................. 30
2.3. PFR operational experience............................................................................................................................... 30
2.3.1. Design features.......................................................................................................................................30
2.3.2. Review of operating history................................................................................................................... 34
2.3.3. Advanced technology developments..................................................................................................... 52
2.3.4. PFR safety and licensing........................................................................................................................ 53
2.3.5. PFR and the fuel cycle........................................................................................................................... 55
2.3.6. The future............................................................................^
2.3.7. PFR in perspective.................................................................................................................................58
2.4. Super Phen1 oipx erating experience...........................................................................................................9...5.. .
2.4.1. Design features..................................................................................................................................9..5.. .
2.4.2. Operating experience............................................................................................................................. 62
2.4.3. The present situation.............................................................................................................................. 71
2.5. BN-600 operation experience...........................................................................................................................2.7
2.5.1. Design features.......................................................................................................................................72
2.5.2. Operating experience.............................................................................................................................78
2.5.3. Radiological safety................................................................................................................................. 87
2.5.4. Fire safety, sodium leaks......................................................................................................................8.8.
2.5.5. BN-600 operating safety enhancement..............................................................................................9..8..
2.5.6. Experimental programmes..................................................................................................................... 91
Bibliography....................................................................................................................................................... 92
CHAPTER 3. EXPERIENCE IN CONSTRUCTION AND PRE-OPERATION AL
TESTING OF THE PROTOTYPE LMFRs.................................................................................... 95
3.1. Construction and pre-operational testing of the LMFR SNR-300.................................................................... 95
3.1.1. Introduction............................................................................................................................................95
3.1.2. Description.............................................................................................................................................95
3.1.3. Constructiond nap re-operational testing...............................................................................................99
3.1.4. Unexpected occurrences and their remedy.......................................................................................... 106
3.1.5. Achievements................................................................................................................................9..0..1...
3.1.6. Epilogue..........................................................................................................................................5.1..1..
3.2. Construction and pre-operational testing of the prototype LMFBR "Monju"................................................ 117
3.2.1. Overviewf o project.............................................................................................................................711
3.2.2. Research and development.................................................................................................................. 118
3.2.3. Design and construction of monju....................................................................................................... 120
3.2.4. Pre-operational tests........................................................................................................................8.2..1..
References...............................................................................................................................................3...3..1... .
CHAPT. 4ER LMFR PHYSICS ..............................................................................................................5...3....1..... .
4.1. Production of nuclear data for reactor neutronics calculations....................................................................... 137
4.1.1. Measurement fo esu eht dna nuclear theory........................................................................................731
4.1.2. Analysis................................................................................................................................................138
4.1.3. Compilation.......................................................................................................................................... 138
4.1.4. Evaluation............................................................................................................................................ 138
4.1.5. Nuclear data uncertainty information.................................................................................................9.31
4.1.6. Integral measurements....................................................................................................................0.4..1..
4.1.7. Fission spectrum average....................................................................................................................04.1
4.1.8. Dosimetrycross-sections..................................................................................................................... 140
4.1.9. Nuclear data processing codes and group cross section data sets....................................................... 141
4.1.10. Validation............................................................................................................................................. 143
4.1.11. Uncertainties in reactor parameters caused by nuclear data uncertainties.......................................... 143
4.2. Energy production, radiation emission, induced radioactivity and irradiation damage.................................. 144
4.2.1. Heating................................................................................................................................................. 145
4.2.2. The radioactivity of irradiated fuel...................................................................................................... 147
4.2.3. Activatiof sont ructud crnoaalo lant materials................................................................................7.4..1..
4.2.4. Irradiation damage effects and dosimetry............................................................................................ 148
4.3. Neutronics calculation methods....................................................................................................................... 149
4.3.1. Forms of heterogeneity........................................................................................................................ 150
4.3.2. Metr htoroedfas ting heterogeneity.............................................................................................1..5...1... .
4.3.3. Control rod homogenisation................................................................................................................ 152
4.3.4. Derivation of broad group cross-sections............................................................................................ 152
4.3.5. Whole reactor calculation methods...............................................................................................2..5..1..
4.3.6. The sub-group treatment of resonance shielding................................................................................. 153
4.3.7. Perturbation theory methods......................................................................................................5...5...1... .
4.3.8. Methof daods justmf econrt oss-sect itinifo toengts ral measurements........................................5...5..1..
4.3.9. Kinetics calculations...........................................................................................................................6.51
4.4. Computer codes used in fast reactor neutronics.............................................................................................. 157
4.4.1. Codes basedn o diffusion theory.........................................................................................................7.51
4.4.2. Transport theory and codes.................................................................................................................. 158
4.4.3. Codes based on the Monte Carlo method............................................................................................ 159
4.4.4. Application of code systems to fast reactor calculations..................................................................... 161
4.5. Validaf tmiooend tdhnaotada.s .............................................................................................................4..6...1... .
4.5.1. Effective multiplication of a core at start-up....................................................................................... 165
4.5.2. Variation of reactivity with bum-up.................................................................................................... 166
4.5.3. Incinerati ffooins sion products ..................................................................................................6..6...1... .
4.5.4. Power distributions........................................................................................................................7..6..1..
4.5.5. Control rod reactivity worths............................................................................................................... 168
4.6. Reactivity coefficients...................................................................................................................................... 169
4.6.1. Doppler effects....................................................................................................................................07.1
4.6.2. Sodium voiding dnas odium density coefficients...............................................................................37.1
4.7. Shielding studies..............................................................................................................................................471
References...........................................................................................................................................5....7....1.... .
Bibliography..................................................................................................................................................... 180
CHAPTER 5. LMFR SAFETY: NEW TRENDS AND FINDINGS................................................................... 185
w treeNndn 5si.sa1 f .ety principd glneoas als...................................................................................................5.8..1..
5.1.1. LMFRs and main recent nuclear safety trends.................................................................................... 185
5.1.2. Safety fundamentals....................................................................................................................5..8...1... .
5.1.3. Risk minimization and accident analysis approaches consequences for LMFRs............................... 186
5.1.4. Localsubassemblyfaults...................................................................................................................4.9.1
5.1.5. Sodium fires......................................................................................................................................... 200
5.2. Emergency heat removal by natural and forced convection............................................................................ 205
5.2.1. Safety objectif veoems ergency heat removal systems (EHRS)..................................................5...0..2.. .
5.2.2. Main design concepts...........................................................................................................................205
5.2.3. Status of international R&D activities and future needs...................................................................... 219
5.2.4. General conclusions.............................................................................................................................220
References........................................................................................................................................................221
CHAPTER 6. SOME ASPECTS OF INSTRUMENTATION AND INSPECTION
OF MAIN LMFR COMPONENTS.............................................................................................. 225
6.1. Steam generator: designs, instrumentd apntriooatne ction..............................................................................225
6.1.1. Configuration...................................................................................................................................5..2.2.
6.1.2. Construction.........................................................................................................................................232
6.1.3. Design details.......................................................................................................................................235
6.1.4. Protection.............................................................................................................................................237
6.1.5. Sodium-water reactionn ssi team generators....................................................................................9.3.2.
6.2. Reactor core with other components: acoustic instrumentation...................................................................... 249
6.2.1. Detection of boiling.............................................................................................................................249
6.2.2. Plant monitoring...................................................................................................................................252
6.3. Ultrasonic instrumentation............................................................................................................................... 255
6.3.1. Ultrasonic imaging...............................................................................................................................256
6.3.2. Ultrasonic monitoring..........................................................................................................................256
6.3.3. Ultrasonic measurement of temperature.............................................................................................. 259
6.3.4. Ultrasonic systems...............................................................................................................................259
6.4. In-service inspection and repair.......................................................................................................................259
6.4.1. Background..........................................................................................................................................262
6.4.2. Operational in-service inspection components.................................................................................... 262
6.4.3. Rd desenevaearlcoh pn mFreiann tce..................................................................................................264
References........................................................................................................................................................270
CHAPTE. 7R CORE STRUCTURAL MATED RFUNIAEALL
TECHNOLOGY FOR HIGH BURN-UP..................................................................................... 273
7.1. Introduction......................................................................................................................................................273
7.2. Behaviour of materials for fuel pin cladding and subassembly duct (wrapper).............................................. 273
n c7il.aP2d.1d .ing materials..........................................................................................................................273
7.2.2. Wrapper-tube (duct) materials.............................................................................................................279
7.3. Irradiation performanf oocxe ide fuel elements........................................................................................0..8...2. .
7.3.1. Fuel and fission product behaviour......................................................................................................283
7.3.2. Mechanical interaction between fuel and cladding (FCMI)................................................................ 286
7.3.3. Chemical interaction between fuel and cladding (FCCI).................................................................... 286
7.3.4. Behaviour under off-normal conditions........................................................................................8..8...2. .
7.3.5. Fain lbeiped haviour............................................................................................................................290
7.3.6. Modellif noogx ide fuel performances ................................................................................................291
7.3.7. Achievable bum-up..............................................................................................................................291
7.4. Irradiation performaf mnocee tallic fuel elements..................................................................................2...9...2.. .
7.5. Irradiation performancf oae dvanced fuel elements....................................................................................4..9.2.
7.5.1. Carbide fuel elements...........................................................................................................................295
7.5.2. Nitride fuel elements............................................................................................................................299
7.5.3. Fr huoiefgl h plutonium burning....................................................................................................0..0..3..
7.6. Irradiation performance of absorber elements................................................................................................. 307
References........................................................................................................................................................313
CHAPTERS. LMFR ENGINEERING.................................................................................................................317
8.1. Sodium pumps.................................................................................................................................................. 317
8.1.1. Design basis......................................................................................................................................... 317
8.1.2. Sodium pump design features..........................................................................................................9.1.3..
8.1.3. Pump testing techniques....................................................................................................................... 336
8.1.4. Pump operating experience.................................................................................................................. 337
8.1.5. Cavitation snio dium pump impellers.............................................................................................0..4.3..
8.1.6. Structural improvements in new sodium pump designs......................................................................343
8.2. Thermal-hydraulics of the primary circuit....................................................................................................... 351
8.2.1. Introduction..........................................................................................................................................351
8.2.2. Objectives of the thermal-hydraulic studies........................................................................................ 351
8.2.3. Steady state studiet ohp ehst foo olsf o pool-type reactors.................................................................2.53
8.2.4. Transient studiet oph sfoo ols.............................................................................................................26.3
8.2.5. Studies of the cold pool of a pool-type reactor.................................................................................... 363
8.2.6. Strongback..................................................................................................................................4...6...3.. .
8.2.7. Diagrid.................................................................................................................................................364
8.2.8. Gas entrainment at the free surface...................................................................................................... 365
8.2.9. Cover gas thermal-hydraulics.............................................................................................................. 366
8.2.10. Thermal stripping studies.....................................................................................................................366
8.2.11. Conclusions..........................................................................................................................................369
8.3. Decommissioning of the experimental liquid-metal fast reactor Rapsodie..................................................... 370
8.3.1. Introduction..........................................................................................................................................370
8.3.2. Main features of the installation........................................................................................................... 370
8.3.3. Pre-decommissioningoperations (1983-1985)...................................................................................372
8.3.4. Decommissioning operations (1986-1994).........................................................................................373
8.3.5. Conclusions..........................................................................................................................................378
References........................................................................................................................................................ 381
CHAPTER 9. ACTIVITIES IN PROGRESS ON LMFR PLANTS.................................................................... 385
9.1. European fast reactor.......................................................................................................................................385
Introduction..........................................................................................................................................385
Organizational e sEthruurtco ftpuoerae n fast reactor cooperation............................................5...8...3.. .
.3. EFR programme status.........................................................................................................................386
.4. Main design features............................................................................................................................ 388
Ds&u.Rp5e pdh. toe frost ign..............................................................................................................6..0.4.
Concept validation phase achievements...........................................................................8......0......4..... ..
R pFroEe jse heucThct.tc. .ef.os..s ...............................................................................................5..1..4..
9.2. BN-1600M.......................................................................................................................................................415
9.2.1. Stages of development and design concept evolution.........................................................................415
9.2.2. Basic paramd denetesairgs n optimization..........................................................................................416
9.2.3. Reactor unit design...............................................................................................................................420
9.2.4. Reactor plant safety..............................................................................................................................429
9.3. BN-800 reactor plant........................................................................................................................................431
9.3.1. Design status........................................................................................................................................ 431
9.3.2. Reactor design basis...................................................................................................................1...3...4... .
9.3.3. Reactor design concept........................................................................................................................431
9.3.4. Main design improvements and reactor plant features........................................................................ 433
9.4. BN-600M advanced LMFR............................................................................................................................. 445
9.4.1. Design status........................................................................................................................................445
9.4.2. Design basis.........................................................................................................................................449
9.4.3. Reactor design concept and main data.................................................................................................449
9.4.4. Safety features......................................................................................................................................451
9.4.5. Improvement of economic performance.............................................................................................. 452
9.5. The demonstration fast breeder reactor...........................................................................................................453
9.5.1. Introduction..........................................................................................................................................453
9.5.2. DFBR plant design...............................................................................................................................453
9.5.3. Techd necnicoaan lomic evaluation.....................................................................................................463
9.5.4. ProspectsR BF rof commercialization.................................................................................................468
9.5.5. Conclusion.......................................................................................................................................9.6..4.
9.6. Prototype fast breeder reactor-design description........................................................................................... 469
9.6.1. Introduction.......................................................................................................................................... 469
9.6.2. Reactor assembly................................................................................................................................. 469
9.6.3. Primary sodium system................................................................................................................5...7..4.. .
9.6.4. Secondary sodium systems............................................................................................................6..7..4..
9.6.5. Fuel handling...................................................................................................................................6..7..4.
9.6.6. Decay heat removal..............................................................................................................................476
9.6.7. Materials..............................................................................................................................................67.4
9.6.8. Instrumentatiod ncna ontrol.................................................................................................................478
9.6.9. Fire protection......................................................................................................................................478
9.6.10. Containment.........................................................................................................................................479
9.6.11. Emergency control room .....................................................................................................................479
9.6.12. Balancf ope lant (BOP)........................................................................................................................479
9.6.13. Present status........................................................................................................................................479
9.7. ALMR technology development......................................................................................................................479
9.7.1. Introduction..........................................................................................................................................479
9.7.2. ALMR plant design..............................................................................................................................484
9.7.3. Integral fast reactor concept development........................................................................................... 501
9.8. BMN-170 modular fast nuclear reactor........................................................................................................... 510
9.8.1. Design goals and status........................................................................................................................510
9.8.2. Design basis...................................................................................................................................2..1..5..
9.8.3. Reactor safety features......................................................................................................................... 515
9.8.4. Radiological safety......................................................................................................................7...1...5.. .
9.8.5. Economics............................................................................................................................................517
9.9. Experimental fast reactor CEFR-25.................................................................................................................518
9.9.1. Introduction.........................................................................................................................................^^
9.9.2. Histod arncya hievements....................................................................................................................518
9.9.3. Trends of experimental reactors .......................................................................................................... 522
9.9.4. CEFR-25 design...................................................................................................................................522
9.9.5. Site work..............................................................................................................................................528
9.9.6. Nuclear licensing systen Cmi hina.................................................................................................9..2..5..
9.9.7. Safety features......................................................................................................................................529
Bibliography.....................................................................................................................................................531
CHAP. T0E1SR UMD FMUNATRAUYR E TRENDS..............................................................................3...3...5.. .
10.1. Introduction......................................................................................................................................................533
10.2. Technology trends............................................................................................................................................533
10.2.1. Fast reactorss ab urnersf op lutonium................................................................................................533
10.2.2. Fast reac stionarcs ineraf tnoours clear waste....................................................................................536
10.2.3. Design styles......................................................................................................................................538
10.2.4. Safety.................................................................................................................................................. 539
e phroTgr1a0m.3 .me.......................................................................................................................................0...4...5.. .
10.4. Conclusion........................................................................................................................................................541
Contributors to drafting and review.............................................................................................................................543
Cha1pte r
BACKGROUND AND OVERVIEW
1.1. PROGRESS WITE PHHEITNR IOD 198- 51 998
1.1.1. Demonstration reactor operation
The outstanding success of the decade has undoubtedly been the reliable operation of the
BN-600 pt lBs aabn enaoen ltpiooehn evjna Rare ariur sttasisk Iloslaina h.sin cde n19a80 ,
lifetime lon Ia1d9 .% 9t f2i aa2 7cc thfooire vea dl oad face htsf Toou8r c 3.c%.s5ea sws
achieved in spite of a number of incidents including sodium fires. The effectiveness of the
protective and remedial measures clearly demonstrates that a sodium-cooled fast reactor is
capable of sustained reliable contribution to an electricity supply system.
In Frane cShetu per Phenix plan taCt reys-Malvs isalwluec cessfully commissiondenad
operated, but not without difficulty. Three incidents marked the commissioning procedure and
caused lengthy delayn s1I. 987 ta hs seoarwde ium leak froe uhmts ed fuel storage drumn 1i, 990
there was an air leak into an auxiliary circuit which caused extensive pollution of the primary
sodium,d na alson i 1990n a exceptional snow fall causede ht collapsee ht fo roofe ht fo turbine
hall with extensive damage to the steam plant. Partly as a result of this experience the safety of
the s ptalhawonr toughly revied wmneaodd ificatioo intms pre orhevtse pono sstee condary
sodium fires were made. A public enquiry on renewal of the operating license was held and
reported s prn e r apeor1oheial9swas di9fttntdy4ia tvrnets .al fyau T,rwhthise r dae lyaybed
small leak of argon from the sealing bell surrounding one of the intermediate heat exchangers.
Corrective action was taken and in the latter part of 1995 and through 1996 power was gradually
raisee dht fot ull-power level. Later political changes called future operation into questiondna
in February 1998 the French Government finally confirmed to discontinue its operation.
1.1.2. Prototype reactor operation
In Germany construction foS NR-30s a0cw ompleted dcna ommissionins agsw uccessfully
taken to an advanced stage. A small leak in a ferritic steel sodium dump tank was found, and
ths easworem e poll euphttri oifmon ary sodiu ymmb oisture released from shielding material.
Effective technical provisions to meet the extensive criticism of the safety of the plant were
successfully put into place, but the project was eventually terminated for political reasons.
The earlier problem fsos team lea ekhet snvi aporaR ptFPo lehart fsno ntiS cotland were
successfully solved, but in 1987 there was a large leak in one of the superheaters. This gave rise
to a major review of the protection against large sodium-water reactions. In 1991 the primary
sodium was polluted by oil from one of the sodium pumps. These technical problems were
overcome successfully, and in addition in 1990 an operating license (which hitherto had not been
a legal necessity) was obtained from the safety authority after a review against the same criteria
appliedo t commercial thermal reactors. Whes awfn tii nally shut downn i 19saw9 RF4P
operating with good reliability.
ehBTN-350 plann Kit azakhsts aaonhp erate roodfv 0ey2r ea ac dransa rs,o fefu ture
operation has been made. Means of improving the protection of the plant against the effects of