Motivations for Fast Neutron Systems
Basic Principles and Consequences
In any nuclear reactor, there is both fissile material destroyed (FD)
and produced by conversion of the fertile material (FP):
The degree of conversion that occurs in a reactor is denoted by the general term of conversion ratio, CR, which is defined as CR = FP/FD. If this conversion ratio is greater than 1, it is called breeding ratio, BR.
Conditions for Breeding
A nuclear reactor can be a breeder in broad neutron energy spectrum, but adequate breeding ratios can be achieved only by selecting the appropriate fertile and fissile isotopes for that spectrum.
Starting from the fission process, if
-
...
Access this chapter
Tax calculation will be finalised at checkout
Purchases are for personal use only
References
A review of fast reactor program in Japan. In: Annual Meeting of the technical working group on fast reactors (TWG-FR), IAEA Vienna, 26–29 May 2008
ABAQUS Version 6.7 Simulia Dassault Systems, USA
Absorber materials, control rods and designs of shutdown systems for advanced LMFR. (1996). IAEA-TECDOC-884
AFNOR 1996, A3-406, RCC-MR, RB 3261.123, 2007 issue. AFCEN, Paris
AGT 1 sub-group 1 (AEA Windscale and Harwell, CEA Cadarache, KfK Karlsruhe, Interatom, Belgonucleaire) (ed) (1990) Catalogue Europeen des propriétés de l’oxyde mixte (U,Pu)O2 – fast reactor data manual 1. CEA
Anderson EP (ed) (1978) Fermi I: new age for nuclear power. ANS publication
Anzieu P, Serpantie J-P, Verwaerde D, Dufour Ph, Martin Ph (2007) A program on innovative SFR in France. In: Proceedings of ICAPP2007, paper 7398, Nice Acropolis, France, 13–18 May 2007
Arai Y, Suzuki Y, Iwai T, Ohmichi T (1992) Dependence of the thermal conductivity of (U,Pu)N on porosity and plutonium content. J Nucl Mater 195:37
Arnol’dov MN et al (2004) The permeability and solubility of hydrogen in a lead-bismuth melt of eutectic composition. High Temp 42(5):715–719
Arrêté du 15 mai 2006 relatif aux conditions de délimitation et de signalisation des zones surveillées et contrôlées et des zones spécialement réglementées ou interdites compte tenu de l’exposition aux rayonnements ionisants, ainsi qu’aux règles d’hygiène, de sécurité et d’entretien qui y sont imposes
ASME Sec III Div 1, Subsection NB-3133, Components under External Pressure, 2007
Astegiano J-C (1990) Assessment of thermal-hydraulic characteristics of primary circuit. Nucl Sci Eng 106
Astegiano JC et al (1981) EFR primary system thermal-hydraulics – status on R&D and design studies. In: Proceedings of the international conference on fast reactor and related fuel cycles, Kyoto, Japan
Athiannan K (2002) Buckling of imperfect cylindrical shell. PhD thesis, IIT Madras
Auger T et al (2008) Liquid metal embrittlement of T91 and 316L steels by heavy liquid metals: a fracture mechanics assessment. J Nucl Mater 377(1):253–260
Axisa F, Antunes J Modelling of mechanical systems. Fluid Structure Interaction Hermès
Bagley H, Harbourne B, Lennov T, Linekar G, Mignanelli M, Plitz H, Pluchery M, Rouault J (1989) Synthesis report on the understanding of failed LMFBR fuel element performance. CEA/UKAEA/KFK/PSB. Report IV 307 KL IV
Balbaud-Célérier F et al (2001) Investigation of models to predict the corrosion of steels in flowing liquid lead alloys. J Nucl Mater 289(3):227–242
Baldasari JP et al (1984) Open azimuthal thermosyphon in annular space – comparisons of experimental and numerical results. In: Liquid metal engineering and technology. BNES, London, pp 463–467
Basmajian JA et al (1972) Nuclear Technology 16:238
Benamati G et al (2002a) Mechanical and corrosion behaviour of EUROFER 97 steel exposed to Pb-17Li. J Nucl Mater 307–311, Part 2, 1391–1395
Benamati G et al (2002b) Temperature effect on the corrosion mechanism of austenitic and martensitic steels in lead-bismuth. J Nucl Mater 301(1):23–27
Blanks DM (1985) Fuel handling options for commercial fast breeder reactors. In: International conference on engineering developments in reactor refueling, England, 13–15 May 1985
Bernard H (1989) Advanced fuel fabrication. J Nucl Mater 166:105
Bertrand F, Devictor N (2006) Combination of deterministic and probabilistic safety analysis in support to the design of new nuclear reactor. In: Proceedings of the IAEA technical meeting on effective integration of deterministic and probabilistic safety analysis in plant safety management, Barcelona, Spain, 4–8 September 2006
Bettes C, Judd AM, Lewis WWJ (1994) Avoiding thermal striping damage: experimentally based design procedures for high cycle thermal fatigue. IWGFR/90. In: Specialists meeting on correlation between material properties and thermohydraulics conditions in LMFBRs, Aix-en-Provence, France, 22–24 November 1994
Betts C, Bourman C, Sheriff N (1983) Thermal striping in liquid metal cooled fast breeder reactors. In: Second international topical manufacturing on nuclear reactor thermal hydraulics, NURETH-2, Santa Barbara, California, pp 1292–1301
Billone MC, Jankus VZ, Kramer JM, Yang CI (1977) Progress in modelling carbide and nitride fuels performance in advanced LMFBRs. In: Advanced LMFBR fuels – topical meeting proceedings, 10–13 October 1977, Tucson, Arizona, USA
Billone et al (1986) Status of fuel element modeling codes for metallic fuels (pp. 5.77–5.91). In: Reliable fuels for liquid metal reactors, TUCSON conference, Arizona, September 1986, p. 5.77
Blanchet Y, Obry P, Louvet J (1981) Treatment of fluid–structure interaction with the SIRIUS computer code. In: SMIRT-6, Paris, B8/8
Blay N, Touboul F, Blanchard MT, Lebreton F (1997a) Piping seismic design criteria: experimental evaluation. In: SMIRT 14, Lyon, K15/4, pp 95–102
Blay N, Touboul F, Blanchard MT, Lebreton F, Cara S (1997b) Piping seismic design criteria: test simulations. In: SMIRT 14, Lyon, K15/5, pp 103–110
BN-350 reactor. (1969). Fuel Processing Technology 12:323–334
Bogolovskaia GP et al (2002) Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors. IAEA-TECDOC-1289, Vienna
Borstedt HU, Champeix L (1989) Corrosion in fast breeder reactors. EFC publication N°1. The Institute of Metals, Brussels, Belgium, ISBN 0-901462-73-X
Bose MRSC, Thomas G, Palaninathan R, Damodaran SP, Chellapandi P (2001) Buckling investigations on a nuclear reactor inner vessel model. In: Experimental mechanics. SAGE publications, vol 41, no 2, pp 144–150
Bour C, Sperandio M, Louvet J, Rieg C LMFBR’s core disruptive accident mechanical study of the reactor bloc of SPX-1. In: Tenth SMIRT, Anaheim, USA
Bourganel S Etudes de radioprotection dédiées au réacteur RNR-Na pour la configuration de cœur Valentin V2B. Rapport DM2S/SERMA/LPEC/RT/08-4602/A
Brehm WF et al (1987) Corrosion and fission products in primary systems of liquid metal cooled reactors in the USA. In: Proceedings of fission and corrosion product behaviour in primary circuit of LMFBRs, IWGFR/64, RFA, Karlsruhe, 5–8 May 1987; Feuerstein H, Thorley AW (eds) (1987) KfK 4279, pp 75–92
Briceno DG et al (2001) Behaviour of F82H mod. stainless steel in lead-bismuth under temperature gradient. J Nucl Mater 296:265–272
Brissonneau L et al (2008) Evaluation of alternative fluids for SFR intermediate loops. In: Proceedings of ICAPP’09, Tokyo
Brissonneau L, Simon N, Saez M, Balbaud F, Rochwerger D, Baqué F, Rodriguez G, Gerber A, Menou S, Prèle G, Capitaine A (2009) The potential use of an alternative fluid for SFR intermediate loops: selection and first design. In: Proceedings of FR’09, Kyoto, Dec 2009
Breuil E, Sperandio M, Waeckel N, Djouini C, Jullien JF (1991) How to improve the post-buckling stiffness of the LMFBR roof slab’s bottom plate. In: SMIRT 11 transactions, vol E, August 1991, Tokyo, Japan
Brochard J, Combescure A, Tomassian R, Locatelli Th (1991) Thermal buckling influence on reduction of critical buckling loads. In: SMIRT 11 transactions, vol E, August 1991, Tokyo, Japan
Buiron L (2009) (Courriel de) Données SFR 3600 pour le dimensionnement des protections. (DER/SPRC) du 11 mars 2009
Buland P (1995) Symphony experimental mockup SMIRT
Bunch WL, O’Dell LD (1969) Fission product inventory and decay heat associated with FTR fuel. BNWL-961
Cabrillat et al (1997) Benchmark on a thermal ratchetting test. Comparison of different constitutive models. Transactions of the 14th conference on structural mechanics (SMIRT 14th), LW/4, Lyon, France
Cahalan JC, Wei TYC (1990) Modeling development for the SAS4A and SASYS computer codes. In: Proceedings of international fast reactor safety meeting, Snowbird, Utah
Cahalan JE et al (1977) A preliminary users guide to version 1.0 of the SAS 3D accident analysis code. SR-239831. ANL
Cameron IG et al (1978) The computer code SEURENUK-2 for fast reactor containment studies. Comp Phys Comm 13:197
Carter JC et al (1970) SAS1A – a computer code for the analysis of fast reactor power and flow transients. ANL-7607
CEA internal reports
CEA (2003) http://www-cast3m.cea.fr/cast3m/. CAST3M – user manual
Chang YM, Gvildys J, REXCO-HEP (1975) A two-dimensional computer code for calculating the primary system response in fast reactors. ANL-75-19. Argonne National Laboratory
Chang YI, Konomura M, Lo Pinto P (2007) A case of small modular fast reactor. J Nucl Sci Technol 44(3):264–269
Chapuliot et al (2005) Hydro-mechanical analysis of thermal fatigue in a mixing tee. Nucl Eng Des 235:575–596
Chellapandi P (2000a) FUSTIN: a code for structural analysis of primary containment under CDA: mathematical modeling. PFBR/31050/DN/1016
Chellapandi P, Chetal SC, Bhoje SB (2000b) Effect of reactor internals on structural integrity of PFBR main vessel under CDA. ASME, PVP 403:161–172
Chellapandi P, Chetal SC, Raj B (Aug 2008) Investigation on buckling of FBR vessels under seismic loading with fluid structure interactions. J Nucl Eng Des 238:3208–3217
Chellapandi P, Chetal SC, Raj B Investigation of structural mechanics failure modes in FBR. In: Raj B et al (eds) Pressure vessels and piping: codes, standards, design and analysis. Narosa Publishing House, New Delhi (in press)
Chellapandi P, Suresh Kumar R, Chetal SC, Raj B (2007) Numerical and experimental simulation of large elastoplastic deformations of FBR main vessel under core disruptive accident loadings. In: IMPLAST 2007, symposium on plasticity and impact mechanics, Ruhr University, Germany, 21–24 August 2007
Chellapandi P, Velusamy K, Kannan SE, Om Pal Singh, Chetal SC, Bhoje SB (2002) Core disruptive accident analysis in prototype fast breeder reactor. In: First national conference on nuclear reactor safety, 25–27 November 2002, Mumbai, India
Ciampichetti A et al (2008) LBE-water interaction in sub-critical reactors: first experimental and modelling results. J Nucl Mater 376(3):418–423
Clement G, Drubay B (1991) Influence of cyclic loading on buckling progressive buckling. In: SMIRT 11 transactions, vol E, August 1991, Tokyo, Japan
Code de calcul CASTEM www.cast3M.fr
Combescure A (1999) Proposition of a design for creep buckling. In: LMT-Cachan/CEA-DMT-SEMT, France, transactions of the 15th conference on SMIRT-15, Seoul, Korea, 15–20 August 15 1999
Combescure A, Brochard J (1991) Recent advances on thermal buckling new results obtained at CEA. In: SMIRT 11 transactions, vol E, August 1991, Tokyo, Japan
Conceptual design of advanced fast reactors. In: Proceedings of a technical committee meeting, Kalpakkam, 3–6 October 1995. IAEA-Tecdoc-907
Core safety improvements in sodium fast reactors. In: INPRO meeting, CEA, Cadarache, 7 February 2007
Courouau JL (2004) Electrochemical oxygen sensors for on-line monitoring in lead-bismuth alloys: status of development. J Nucl Mater 335(2):254–259
Courouau JL (2007) Chemistry control and monitoring systems. In: Fazio C (ed) Handbook on lead-bismuth eutectic alloy and lead properties, materials compatibility, thermal-hydraulics and technologies, OECD, pp 129–177
Cowler MS, Hancock SL (1979) Dynamic fluid–structure analysis of shells using the PISCES-2DELK computer code. In: SMIRT-5, Berlin, B1/6
Crawford et al (2007) Fuels for sodium cooled fast reactors – US perspective. Journal of Nuclear Materials 371:202–231
Crespo LS et al (2001) Short-term static corrosion tests in lead-bismuth. J Nucl Mater 296:273–281
Defense in depth in nuclear safety – INSAG-10 (1996). A report of international. Nuclear Safety Advisory Group, IAEA, Vienna
Delangre E (2001) Fluides et solides Les éditions de polytechnique
Design of the Clinch river breeder reactor plant steam generator. (1976). Nucl Tech 28:305–314
Donea J, Fasoli-Stella P, Giuliani S, Halleux JP, Jones AV (1980) The computer code EURDYN-1M for transient dynamic fluid–structure interaction. EUR 6751. Commission of the European Communities
Dufour Ph (2007) Post accident heat removal analysis for SPX and EFR. In: CEA-IGCAR technical seminar on liquid metal fast reactor safety aspects related to severe accidents, Kalpakkam, India, February 2007
Dunn FE et al (1974) The SAS2A LMFBR accident analysis computer code. ANL-8138
Eguchi Y, Yamamoto K, Funada T et al (1994) Gas entrainment in the IHX vessel of top-entry loop type LMFBR. Nuclear Engineering and Design 146:373–381
Eptein M, Grolmes MA, Henry RE, Fauske KK (1976) Transient freezing of a flowing ceramic fuel in a steel channel. Nucl Sci Eng 61:310
Evaluated nuclear data file (2008). http://www-nds.iaea.org/exfor/endf.htm#1
Experience with liquid metal fast breeder reactor steam generators – U.S design. (1981). Nuclear Technology 55:60–87
Farrar B (1990) Use of the PHOENCIS CFD code to simulate the 3D thermal hydraulic behavior of a pool type LMFBR hot plenum. In: IMechE seminar on thermohydraulics of nuclear reactors, London, pp 95–105
Farvacque M et al (2003) CATHARE 2 v2.5: a fully validated CATHARE version for various applications. In: Proceedings of NURETH-10, Séoul, Korea, 2003
Fazio C et al (2001) Compatibility tests on steels in molten lead and lead-bismuth. J Nucl Mater 296(1–3):243–248
Fazio C (2007) handbook on lead-bismuth eutectic alloy and lead properties, materials compatibility, thermal-hydraulics and technologies, vol 1. OECD, Paris
Fink JK (2000) Zircaloy thermal conductivity. Argonne National Laboratory
Fleitman AH et al (1971) Mercury as a nuclear coolant. Nucl Eng Design 16:266–278
Foletti C et al (2008) ENEA experience in oxygen measurements. J Nucl Mater 376(3):386–391
Fontaine B (1997) Seismic analysis of LMFBR reactor cores. SYMPHONY mockup SMIRT
François G et al (2008) Sodium fast reactor concepts. In: Proceedings of ICAPP’08, Anaheim
Francois G, Azarian G (1989) SUPER PHENIX reactor block thermalhydraulic behaviour comparison between calculations and experimental results. In: Tenth international conference on structural mechanics in reactor technology, Lyon, France, pp 37–42
French steam generator experience – Phenix and beyond. (1976). Nuclear Technology 28:482–488
Fritz J (1972) The effect of liquids on the dynamic motions of immersed solids. J Eng Ind
Funada T, Yamamoto K, Eguchi Y et al (1991) Gas entrainment in the IHX vessel of top-entry loop type LMFBR. In: Proceedings of NURETH-5, Salt Lake City, pp 1399–1406
Gabor JD et al (1974) Studies and experiments on heat removal from fuel debris in sodium. In: Proceedings of fast reactor safety conference, FONF-740401, p 823
GCFS (2007) Fondamentaux de Sûreté Nucléaire pour la conception des Réacteurs de Génération IV. Safety Approach for the design and the assessment of future nuclear systems. FRENCH ADVISORY GROUP ON SAFETY, ICAPP
Gelineau O, Sperandio M (1994) Thermal fluctuation problems encountered in LMFBRs. IAEA-IWGFR/90. In: Specialistic meeting on correlation between material properties and thermohydraulics conditions in LMFBRs, Aix-en-Provence, France, 22–24 November 1994
Gelineau et al (1999) Predictive methods applied to thermal striping problems and recommendations, SMIRT 15. Paper F 0/6
GenIV GIF. Basis for the safety approach for design & assessment of generation IV nuclear systems. Report GIF/RSWG/2007/002/Rev.1
Gessi A et al (2005) Parametric experiments on Eurofer steel corrosion by Pb-17Li; LB-A-R-022. Final report on EU Task TW2-TTBC003.D3, ENEA Brasimone
Gessi A et al (2008) Corrosion experiments of steels in flowing Pb at 500 ̈C and in flowing LBE at 450°C. J Nucl Mater 376(3):269–273
Gibert RJ (1988) Vibration des structures. Eyrolles
Gluekler EL, Huang TC, Jospeh D (1979) In-vessel retention of core debris in LMFBRs. In: Proceedings of international meeting in fast reactor safety technology, Seattle
Gluekler EL et al (1982) Analysis of in-vessel core debris retention in large LMFBRS. In: Proceedings of the LMFBR safety topical meeting, Lyon
Golden GH, Tokar JV (August 1967) Thermophysical properties of sodium. ANL-7323. Argonne National Laboratory, Illinois
Goldstein S, Joly J, Vidard M (1979) Thermal analysis of the penetrations of a LMFBR. In: Fifth international conference on structural mechanics in reactor technology
Gorse-Pomonti D et al (2007) Liquid metals for nuclear applications. J Non-Crystalline Solids 353:32–40, 3600–3614
Gosset G, Simeone D, Quirion D (2000) Endommagement du carbure de bore sous irradiation neutronique : évaluation en diffraction X. J Phys Fr IV(Pr), 10–55
Gosset J, Gicquel R, Lecomte M, Queiros-Conde D (2005) Optimal design of the structure and settings of nuclear HTR thermodynamic cycles. Int J Thermal Sci 44(12):1169–1179
Glasbrenner H et al (2001) Corrosion investigations of steels in flowing lead at 400°C and 550°C. J Nucl Mater 296(1–3):237–242
Graham J (1971) Fast reactor safety. Academic, New York
Graveleau JL, Louvet P (1979) Calculation of fluid–structure interaction for reactor safety with the CASSIOPEE code. In: Proceedings of the fifth international conference on structural mechanics in reactor technology, Paper B1/7, Berlin, Germany
Gregory JN et al (1956) The static corrosion of nickel and other materials in molten caustic soda, AERE C/M 272, Harwell, Berks
Guérin Y, Rouault J (1986) In-pile behaviour of carbide fuel elements designed for low doubling time. In: Proceedings of the international conference on reliable fuels for LMR, 7–11 September 1986, Tucson, Arizona, USA
Guidez J, Gognet G (1990) Simulation by water-test of the argon entrainment in the sodium breeder. In: Proceedings of second international symposium on gas transfer at water surfaces, Minneapolis
Harbur et al (1970) Studies on U–Pu–Zr alloy system for fast breeder applications. Report TID-4500
Harish R, Sathiyasheela T, Stinivasan GS, Om Pal Singh (1999) KALDIS: a computer code system for core disruptive accident analysis of fast reactors. IGC-208
Harlow FH, Welch JE (1965) Numerical calculation of time dependent viscous incompressible flow of fluid with free surfaces. Physics of Fluids 8:2182–2188
Haubensack D, Thevenot C, Dumaz P (2004) The COPERNIC/CYCLOP computer tool: pre-conceptual design of generation 4 nuclear systems. In: 2nd international topical meeting high temperature reactor technology, HTR 2004, Beijing, China, 22–24 September 2004
Heuvel HJ, Höller P, Donner P (1985) Absorber material cladding chemical interaction in vented FBR absorber pins. J Nucl Mat 130:517
Hiroshi Akiyama. ‘Seismic Resistance of FBR Components Influenced by Buckling, Kajma Institute publishing, Japan, 1997
Hirt CW, Nichols BD (1981) Volume of fluid (VOF) method for the dynamics of free boundaries. Journal of Computational Physics 39: 201–225
Hu LW, Kazimi MS (2006) LES benchmark study of high cycle temperature fluctuations caused by thermal striping in a mixing tee. J Heat Fluid Flow
IAEA annual reports: status of fast reactor programmes
IAEA (1996) Absorber materials, control rods and designs of shutdown systems for advanced liquid metal fast reactors. IAEA-TECDOC-884
IAEA (1996) Defense-in-depth in nuclear safety. IAEA safety series no. 75. INSAG-10
IAEA (April 1999) Status of liquid metal cooled fast reactor technology. IAEA-TECDOC-1083
IAEA (2000) Safety of nuclear power plants: design. IAEA safety standards series NS-R-1
IAEA (2005) Assessment of defence in depth for nuclear power plants. IAEA safety report series no. 46
IAEA (2007) Proposal for a technology – neutral safety approach for new reactor designs. IAEA-TECDOC-1570
IAEA INSAG10 Defence in depth in nuclear safety
Igari et al (1993) Proposal of a new estimation method for the thermal ratchetting of a cylinder subjected to a moving temperature distribution. Nucl Eng Des 139:261–267
Igari et al (2002) Inelastic analysis of new thermal ratchetting due to a moving temperature front. Int J Plas 18:1191–1217
INEEL (2001) RELAP5-3D{ ©} code manual, vol I: code structure, system models, and solution methods. Report INEEL-EXT-98-00834. Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho
International symposium on fast breeder reactors: experience and future trends, Lyon, France, 22–26 July 1985
Iritani Y et al (1991) Development of advanced numerical simulation of thermal stratification by highly-accurate numerical method and experiments. In: Proceedings of the international conference on fast reactor and related fuel cycles, Kyoto, Japan
Iwai T, Nakajima K, Arai Y, Suzuki Y (1996) Fission gas release of uranium–plutonium mixed nitride and carbide fuels. IAEA-TECDOC-970. Studies on fuels with low fission gas release
Iwasaki T, Konashi K, (2009) Development of hydride absorber for fast reactor, Application of hafnium hydride to control rod of large fast reactor. J Nucl Sci Tech 46(8):874-882
Jackson JF, Nicholson RB (1972) VENUS-II; a LMFBR disassembly program. ANL-7951
Jasserand F Etudes de radioprotection dédiées au réacteur RNR-Na pour la configuration de cœur Valentin V2B. Rapport DM2S/SERMA/LPEC/RT/09-4694/A
Justin F, Natta M, Orzoni G (1985) Safety criteria for the future LMFBRs in France and main safety issues for the Rapide 1500 project. In: Proceedings of the international topical meeting on fast reactor safety, Knoxville
Kaguchi H et al (1999) Strain limits for structural integrity assessment of fast reactors under CDA. In: Proceedings of ICONE-7, Tokyo,Japan
Kasahara, Lejeail (2002) Frequency response approach for thermal fatigue induced by random fluctuation of fluid temperature. ASME, PVP
Kasahara et al (2002) Structural response function approach for evaluation of thermal striping phenomena. Nucl Eng Des 212:281–292
Kazimi MS, Tsai SS, Gasser RD (1977) Post accident fuel relocation and heat removal in LMFBR. BNL-50570
Kikuchi K et al (2004) Lead-bismuth eutectic compatibility with materials in the concept of spallation target for ADS. JSME Int J Ser B Fluids Thermal Eng 47(2):332–339
Kobus H (1991) Introduction to air–water flows. In: Wood IR (ed) Air entrainment in free surface flows. IAHR hydraulic structures design manual. A.A. Balkema, Rotterdam
Kondo M et al (2005) Metallurgical study on erosion and corrosion behaviors of steels exposed to liquid lead-bismuth flow. J Nucl Mater 343(1–3):349–359
Kondo M et al (2006a) Corrosion resistance of Si- and Al-rich steels in flowing lead-bismuth. J Nucl Mater 356(1–3):203–212
Kondo M et al (2006b) Study on control of oxygen concentration in lead-bismuth flow using lead oxide particles. J Nucl Mater 357(1–3): 97–104
Konys J et al (2004) Oxygen measurements in stagnant lead-bismuth eutectic using electrochemical sensors. J Nucl Mater 335(2): 249–253
Kotake S et al (1993) Application of the PSA method to decay heat removal systems in large scale FBR design. In: Proceedings of the IAEA specialists’ meeting on evaluation of decay heat removal by natural convection, Oarai, Japan
Latgé PhD (October 1981) Study of the Na2O crystallization in liquid sodium. Application to the sodium purification in SFR
Latge S, Sellier (1993) Oxidation of zirconium–titanium alloys in liquid sodium: validation of a hot trap, determination of the kinetics. In: Borgstedt HU (ed) Material behaviour and physical chemistry in liquid metal systems, vol 2. Plenum Press, New York
Laxman D et al (2004) Free level fluctuations study in 1 ∕ 4 scale reactor assembly model of PFBR. In: NUTHOS-6, Nara, Japan, 4–8 October
Le RC, Kayer G (1979) An internal core catcher for a pool type LMFBR and connected studies. In: Proceedings of international meeting on fast reactor safety technology, Seattle
Lee YK, Hugot FX Validation/qualification du code Monte Carlo TRIPOLI-4.4 en neutronique-criticité: résultats. Rapport SERMA/LEPP/RT/05-3613/A
Lee YK, Hugot FX Validation/qualification du code Monte Carlo TRIPOLI-4.4 en protection: résultats. Rapport SERMA/LEPP/RT/05-3621/B
Lejeail, Kasahara (2005) Thermal fatigue evaluation of cylinders and plates submitted to fluid temperature fluctuations. Int J Fat 27:768–772
Lenoir G et al (1981) Thermal hydraulics of the annular spaces in roof slab penetrations at liquid sodium cooled fast breeder reactor. In: Sixth international conference on structural mechanics in reactor technology, Paris, pp 104–118
Libman J (1996) Elements of nuclear safety. Les Editions de Physiques
Liquid metal cooled reactors: experience in design and operation. (2007). IAEA TECDOC – 1569
Liquid metal fast breeder reactor steam generator design and experience in UK. (1986). Nuclear Energy 6:355–360
Louvet J, Hamon P, Smith BL, Zucchini A (1987) MARA 10: an integral model experiment in support of LMFBR containment analyses. In: Ninth SMIRT, vol E
Marth W (1988) Sodium – still the best coolant for fast breeder reactor. In: Proceedings of LIMET, SFEN, AvignonSFEN
Martin Ph, Rouault J, Serpantie J-P, Verwaerde D French program towards a GEN IV sodium cooled fast reactor. In: ENC 2007, Brussels, Belgium, 16–20 Sept 2007
Martin P, Pelletier M, Every D, Buckthorpe D (eds) (2008) French and United Kingdom experience of high-burnup mixed-oxide fuel in sodium-cooled fast breeder reactors. Nucl Tech 161:35
Maruyama T, Onose S, Kaito T, Oriuchi H (1997) Effect of fast neutron irradiation on the properties of boron carbide pellet. J Nucl Sci Tech 34(10):1006
Maruyama T, Onose S (1999) Fabrication and properties of boron carbide/copper cermet, J Nucl Sci Tech 36(4):380
Matsuura S, Nakamura H (1997) Shear bending buckling analysis of fast breeder reactor main vessel. In: Seismic resistance of FBR components influenced by buckling. Kajma Institute, Japan
Matzke Hj (1986) Science of advanced LMFBR fuels. North Holland Physics Publishing, Holland
Mejane H, Durin M (1982) Natural convection in an open annular slot. In: Proceedings of seventh international heat transfer conference, Meichen, Germany
Menant B, Villand M (1994) Thermal fluctuations induced in conducting wall by mixing sodium jets: an application of TRIO-VF using large eddy simulation modeling. In: IAEA specialist meeting on correlation between material properties and thermohydraulics conditions in liquid metal-cooled fast reactors (LMFRs), Aix-en-Provence, France, 22–24 November 1994
Mineral commodity summaries 2007 US Geological Survey, Ed. Washington, 2007
Miyahara S et al (2006) Reaction, transport and settling behavior of lead-bismuth eutectic in flowing liquid sodium. In: Proceedings of 14th international conference of nuclear engineering (ICONE), ASME, Miami
Moriya S et al (1988) Thermal striping in coaxial jets of sodium, water and air. In: Proceedings of the fourth international conference on liquid metal engineering and technology, Avignon, France
Müller G et al (2002) Results of steel corrosion tests in flowing liquid Pb/Bi at 420-600 ̈C after 2000 h. J Nucl Mater 301(1):40–46
Müller G et al (2003) Control of oxygen concentration in liquid lead and lead-bismuth. J Nucl Mater 321(2–3):256–262
Muramatsu T (1993a) Intensity and frequency evaluation of sodium temperature fluctuation related to thermal striping phenomena based on numerical methods. In: Proceedings of the fifth international symposium on refined flow modelling and turbulence measurements, Paris, France, 1993
Muramatsu T (1993b) Frequency evaluation of temperature fluctuations related to thermal stripping phenomena using a direct numerical simulation code DINUS-3. In: Proceedings of ASME PVP conference, Colorado, vol 253
Muramatsu T (1994a) Development of thermohydraulics computer programs for thermal striping phenomena. In: IAEA specialist meeting on correlation between material properties and thermohydraulics conditions in liquid metal-cooled fast reactors (LMFRs), Aix-en-Provence, France, 22–24 November 1994
Muramatsu T (1994b) Investigation on the reduction measures of coolant temperature fluctuations based on numerical methods in LMFR designs. In: IAEA specialist meeting on correlation between material properties and thermohydraulics conditions in liquid metal-cooled fast reactors (LMFRs), Aix-en-Provence, France, 22–24 November 1994
Muramatsu T (1998a) Computer code developments and their validations for the thermal striping phenomena in PNC. In: Third research coordination meeting (RCM) of the IAEA coordinated research programme (CRP) on harmonization and validation of fast reactor thermomechanical and thermohydraulic codes and relations using experimental data, France
Muramatsu T (1998b) Numerical analysis of non-stationary thermally response characteristics for a fluid–structure interaction system. In: ASME/JSME pressure vessels and piping conference, USA, 1998
Muramatsu T, Ninokata H (1991) Intensity evaluation of the temperature fluctuations related to thermal striping phenomena using the algebraic stress turbulence model. In: Proceedings of ANS winter meeting, San Francisco, pp 156–162
Nam HO et al (2008) Dissolved oxygen control and monitoring implementation in the liquid lead-bismuth eutectic loop: HELIOS. J Nucl Mater 376(3):381–385
Noraiki Takahashi et al (1989) Study of an advanced fuel handling system. Nucl Technol 86
Noden JN (1972) A general equation for the solubility of O2 in liquid. Sodium British Report RB/B/N 2500
Nonaka N, Sato I (1992) Improvement of evaluation method for initiating-phase energetics based on CABRI-1 in-pile experiments. Nucl Tech 98:54
Ogata A, Yokoo A (1999) Development and validation of ALFUS: an irradiation behaviour analysis code for metallic fast reactor fuels. Nucl Tech 128:113–123
Ohshime H et al (1994) Current status of studies on temperature fluctuation phenomena in LMFBRs. IAEA-IWGFR/90. In: Specialists meeting on correlation between material properties and thermohydraulics conditioning in LMFBRs, Aix-en-Provence, France, 22–24 November 1994
Olander DR (1976) Fundamental aspects of nuclear reactor fuel elements. ERDA Technical Information Center, Oak Ridge
Old CF (1980) Liquid metal embrittlement of nuclear materials. J Nucl Mater 92:2–25
Om Pal Singh, Harish R (1998) Results of transient calculations up to onset of boiling of a comparative calculation for unprotected loss of flow accident in BN-800 type reactor with near zero void reactivity coefficient. In: IAEA/EC consultancy meeting on the comparative calculations for severe accident in BN-800 reactor, Obninsk, Russia, 2–16 June 1998
Operating experience on fast reactor. (1986). Nucl Ene 2:73–84
Ostensen RW, Henry RE, Jackson JF, Goldfuss GT, Gunther WH, Parker NE (1974) Fuel flow and freezing in the upper subassembly structure following an LMFBR disassembly. Trans Amer Nucl Soc 18:214
Pacheco JE (2002) Final test and evaluation results for the solar two project, SAND 2002-0120, Sandia National Laboratories, Albuquerque
Pahl et al (1988) Experimental studies of U–Pu–Zr fast reactor fuel pins in EBR-II. Conf 8009202-2
Papazoglou IA, Aneziris ON (2003) Master logic diagram: method for hazard and initiating event identification in progress plants. Journal of Hazardous Material A97:11–30
Pascal P Nouveau traité de chimie minérale, vol II – Lithium. Sodium Edition Masson et Cie
Petiot P, Seiler JM, CEN G (Grenoble) (1984) Physical properties of sodium: a contribution to the estimation of critical coordinates. High Temperatures High Pressures 16:289–293
Popiel CO, Trass O (1991) Visualization of a free and impinging round jet. Experimental Thermal and Fluid Science; Elsevier Science, Elsevier, Amesterdam
Porten DR et al (1981) Internal fuel motion as inherent shutdown mechanism. In: Proceedings of topical meeting on reactor safety aspects of fuel behaviour, Sun Valley, Idaho
Presentation by AEA Technology – UK (1994) Thermal striping benchmark exercise: thermal hydraulic analysis of the T-junction. In: Proceedings of specialists meeting on correlations between material properties and thermohydraulics conditions in LMFR, Aix-en-Provence, France, 22–24 November 1994
Proceeding of international conference on fast reactors and related fuel cycles (FR’91), Kyoto, 28 October–1 November 1991
Raj B (2007) Assessment of Indian INS. In: INPRO meeting, CEA, Cadarache, 7 February2007
Rasmussen NC (1975) Reactor safety study: an assessment of accident risks in US commercial nuclear power plants. WASH 1400-NUREG-75/014. US DOE
RCC-MR (2007a) Règles de Conception et de Construction des Matériels Mécaniques des Installations Nucléaires. AFCEN editions
RCC-MR (2007b) RB 3261.1112, out of the creep range/RB 3262.1112, in the creep range. AFCEN, Paris
RCC-MR (2007c) RB 3261.114, out of the creep range/RB 3262.1131, in the creep range, 2007 issue. AFCEN, Paris
RCC-MR (2007d) RB3261.115, out of the creep range/RB 3262.114, in the creep range, 2007 issue. AFCEN, Paris
RCC-MR (2007e) Section I, Design and construction rules for mechanical components of nuclear installations. RCC-MR, Subsection Z, Appendix A12, AFCEN, Paris
RCC-MR (2007f) Section I, Design and construction rules for mechanical components of nuclear installations. RCC-MR, Subsection Z, Appendix A7, AFCEN, Paris
RCC-MR (2007g) Section I, Design and construction rules for mechanical components of nuclear installations. RCC-MR, Subsection Z, Appendix A3, AFCEN, Paris
RCC-MR (2007h) Section I, Design and construction rules for mechanical components of nuclear installations. RCC-MR, Subsection B, AFCEN, Paris
RCC-MR (2007i) RB3261. 123. AFCEN, Paris
RCC-MR(2002) Design and construction rules for mechanical components for FBR nuclear islands, Section I, Subsection B, Class 1 Components, AFCEN, 2002.
Roche JL, Nouailhas D (1989) A unified constitutive model for cyclic viscoplasticity and its applications to various stainless steels. ASME J Eng Mater Technol 111:424–430
Ronchi C, Campana M, Coquerelle M, Van de Laar J (1984) Reactor performance of MC, MN, MCN and MCO: results of the comparative irradiation experiments. European Applied Research–Nuclear Science Technology, Gocar, vol 6, nos 1 and 2, p. 323
Roubin P et al (1988) Thermal-hydraulic study of LMFBR hot pool with internal storage. An experimental and computational approach. In: Proceedings of the fourth international conference on liquid metal engineering and technology, Avignon, France
Roux S, Elie D (1988) Comparison between measurement and computational analysis on open azimutal thermosyphons in annular space of the Lilliput model. In: Liquid metal engineering and technology, pp 410_1–410_12
RSWG (2008) Basis for the safety approach for design & assessment of generation IV nuclear systems. November 24, 2008
Saez M et al (2007) Thermal criteria to compare fast reactors coolants for the intermediate loop. In: Proceedings of GLOBAL 07 – advanced nuclear fuel cycles and systems, ANS, Boise
Safety of nuclear power plants: design (2000). IAEA safety standards series, NS-R-1. IAEA
Sathiyasheela T, Harish R, Om Pal Singh (1998) A comparative study of ULOF for BN-800 reactor with non-zero sodium void coefficient. RPD/SAS-101
Scherbakov SI (1994) Computational investigations of thermohydraulic aspects of two convergent flows at different temperatures mixed in the Tee junction area. In: Proceedings of specialists meeting on correlations between material properties and thermohydraulics conditions in LMFR, Aix-en-Provence, France, 22–24 November 1994
Section III, Div 1, Subsection NH, Class 1 Components in elevated temperature service, 2001
Seran JL, Levy V, Dubuisson P, Gilbon D, Maillard A, Fissolo A, Touron H, Cauvin R, Chalony A, Le Boulbin E (1992) Behavior under neutron irradiation of the 15-15Ti and EM10 steels used as standard materials of the Phénix fuel subassembly. In: Proceedings of 15th international symposium: effects of radiation on materials, p. 1209. ASTInternational
Shuh PY, CC Scott (December 2005) Attenuation measurements of sound and performance of ultrasonic transducers in 600°F liquid Na. APDA-180
Siegel R, Norris RH (1957) Transactions of ASME 79(I):663–673
Sienicki J, Moisseytsev A, Cho D, Momozaki Y, Kilsdonk D, Haglund R, Reed C, Farmer M (2007) Supercritical carbon dioxide Brayton cycle energy conversion for sodium-cooled fast reactors/advanced burner reactors. In: GLOBAL ’07 – advanced nuclear fuel cycles and systems, Boise, 9–13 Sept 2007
Sigrist JF, Broc D (2007) Dynamic analysis of a tube bundle with fluid structure interaction modelling using a homogenization method. Computer Methods in Applied Mechanics and Engineering
Silver et al Three-dimensional finite-element analysis of the cellular convection phenomena in the Clinch River Breeder Reactor Plant prototype pump
Simeone D, Deschanels X, Berthier B, Tessier C, (1997) Experimental evidence of lithium migration out of an irradiated boron carbide material. J Nucl Mat 245:27
Simon N, Latgé C, Gicquel L (2007) Investigation of sodium – carbon dioxide interactions with calorimetric studies. In: Proceedings of ICAPP2007, paper 7547, Nice Acropolis, France, 13-18 May 2007
Smith MR (1990) Techniques for the investigation of scaling criteria for gas entrainment mechanisms in liquid metal cooled fast reactors. GEC Journal of Research 8: 49–56
Sobolev VA, Kuzavkov NG (1994) Identification of places with fluid temperatures in BN 600 reactor and reactor systems. IAEA-IWGFR/90. In: Specialists meeting on correlation between material properties and thermohydraulics conditions in LMFBRs, Aix-en-Provence, France, 22–24 November 1994
Srinivasan GS (2002) Kinetic studies of reactivity oscillations under seismic conditions. PFBR/01117/DN/1019. IGCAR report
Status of French nuclear program. In: Meeting of the technical working group on fast reactors (TWG-FR), 26–29 May 2008
Status of liquid metal cooled fast reactors technology. (1999). IAEA-TECDOC-1083
Stevenson MG et al (1974) Current status and experimental basis of the SAS LMFBR accident analysis code system. In: Proceedings of international conference on fast reactor safety, Beverley Hills, California
Subbotin VI et al (2002) Liquid metal coolants for nuclear power. Atomic Energy 92(1): 29–40
Surle F et al (1993) Comparison between sodium stratification tests on the CORMORAN model and TRIO-VF computation. In: Proceedings of the sixth international topical meeting on nuclear reactor thermal hydraulics, Grenoble, France
Tenchine D et al (1990) Sodium thermal-hydraulics in the pool LMFBR primary vessel. Nuclear Engineering and Design 124
The BN-800 reactor – a new stage in fast reactor development. IAEA/SM/284. pp 209–216
The nuclear fuel of pressurized water reactors and fast reactors – design and behaviour. (1999). In: Bailly H, Menessier D, Prunier C (eds) Collection du Commissariat à l’Energie Atomique. Lavoisier
Thomson TJ, Beekerly JG (1964) The technology of nuclear rector safety, vol 1. MIT Press
Timo DP (1954) Free convection in narrow vertical sodium annuli. Knolls Atomic Power Laboratory Report: 1082. Schenectady
Tobita Y et al (1999) Evaluation of CDA energetics in the prototype LMFBR with latest knowledge and tools. In: Proceedings of ICONE-7, Tokyo
Toda et al (1990) Natural convection in a vertical narrow annular gap. In: Proceedings of ninth international heat transfer conference, Jerusalem, Israel, vol 2, pp 199–204
Tokuhiro A, Kimura N (1999) An experimental investigation on thermal striping mixing phenomena of a vertical non-buoyant jet with two adjacent buoyant jets as measured by ultrasound Doppler velocimetry. J Nucl Eng Des
Touboul F, Blay N, Lacire MH (1999) Experimental, analytical, and regulatory evaluation of seismic behavior of piping systems. J Pres Ves Tech 121(November):388–392
Touboul F, Blay N, Sollogoub P, Chapuliot S (2006) Enhanced seismic criteria for piping. Nucl Eng Des 236(1):1–9. Janvier
Tourasse M, Boidron M, Pasquet B (1992) Fission product behaviour in Phenix fuel pins at high burn up. J Nucl Mater 187:122
Tsilanizara A, Huyhn TD, Jouanne C, Luneville L Guide d’utilisation du logiciel Darwin/Pepin2 version 2.2. Rapport DM2S/SERMA/LPEC/RT/07-4168/A
Vaidyanathan G An overview of activities pertaining to heat and mass transfer in reactor cover gas of FBTR. In: Proceedings of specialists meeting on hat and mass transfer in cover gas, IWGFR-57, Harewell, England
Vanavaramban S (2008) Solidus and liquidus temperatures for metallic fuels. CG
Velusamy K et al (2005) Investigations of thermal striping in primary circuit of prototype fast breeder reactor. In: Proceedings of ICONE 13, Beijing, China
Wada et al (1995) Mechanism-based evaluation of thermal ratchetting due to moving temperature distribution. ASME PVP 313(2):471–480
Wakamatsu M et al (1995) Attenuation of temperature fluctuations in thermal striping. Journal of Nuclear Science and Technology 32(8)
Wakamatsu M, Nei H, Hashiguchi K (1995) Attenuation of temperature fluctuations in thermal striping. Journal of Nuclear Science and Technology 32(8):752–762
Walker RA et al (1970) The solubilities of bismuth and tellurium in liquid sodium. J Nucl Mater 34:165–173
Waltar AE, Reynolds AB (1981) Fast breeder reactors. Pergamon Press
Walter AE, Reynolds AB (1982) Fast breeder reactor. Pergamon, UK
Walters et al (1984) Performance of metallic fuels and blankets in liquid metal breeder reactors. Nuclear Technology 65:179–231
Walters, Reynolds Fast breeder reactors. Pergamon
Wang CY, ICECO (December 1975) An implicit Eulerian method for calculating fluid transient in fast reactor containment. ANL-75-81
Weeks JR (1971) Lead, bismuth, tin and their alloys as nuclear coolants. Nucl Eng Design 15:363
Weisenburger A et al (2008) T91 cladding tubes with and without modified FeCrAlY coatings exposed in LBE at different flow, stress and temperature conditions. J Nucl Mater 376(3): 274–281
Wider HU et al (1982) Status and validation of the SAS4A accident analysis code system. In: Proceedings of LMFBR safety topical meeting, ANS, France
Winterton RHS (1972) Cover-gas bubbles in recirculating sodium primary coolant. Nuclear Engineering and Design 22:262–271
Wittingham (1976) An equilibrium and kinetic study of the liquid sodium hydrogen reaction. Journal of Nuclear Materials 60:119
Working Group Report – A Technology Roadmap Generation. (December 2002). GIF-007-00
Yamakawa M (1983) Thermal hydraulic analysis in LMFBR plenum. In: International conference on numerical methods in nuclear engineering, Montreal, Canada
Yamakawa et al (1986) Analysis of natural convection in narrow annular gaps of LMFBR. Journal of Nuclear Science and Technology 23(5): 451–460
Yamano H et al (2008) Development of 3-D CDA analysis code: SIMMER-IV and its application to reactor case. Nuclear Engineering and Design 238:66
Yevick JG (1966) Fast reactor technology. Plant design. MIT Press, USA
Yokoo et al (August 2000) A design study on the FBR metal fuel and core for commercial applications. Nucl Sci Tech 37(8): 636–645
Zimmermann H (1982) Investigation of swelling of U–Pu mixed carbide. J Nucl Mater 105:56
Zuppiroli L, Lesueur D (1989) Modelling the swelling and microcracking of boron carbide under neutron irradiation. Phil Mag 60(5):539
Author information
Authors and Affiliations
Editor information
Editors and Affiliations
Rights and permissions
Copyright information
© 2010 Springer Science+Business Media, LLC
About this entry
Cite this entry
Rouault, J. et al. (2010). Sodium Fast Reactor Design: Fuels, Neutronics, Thermal-Hydraulics, Structural Mechanics and Safety. In: Cacuci, D.G. (eds) Handbook of Nuclear Engineering. Springer, Boston, MA. https://doi.org/10.1007/978-0-387-98149-9_21
Download citation
DOI: https://doi.org/10.1007/978-0-387-98149-9_21
Publisher Name: Springer, Boston, MA
Print ISBN: 978-0-387-98130-7
Online ISBN: 978-0-387-98149-9
eBook Packages: EngineeringReference Module Computer Science and Engineering