Numerical investigation of SUPERFACT-1 experiment by a MOX fuel pin thermo-mechanical model
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Preparations are currently underway to develop ISAA-Na code, a sodium-cooled fast reactors (SFRs) specific version of an integrated severe accident analysis code. This version aims to evaluate both protected and unprotected accidents in SFRs. Given the crucial role of fuel pin thermo-mechanical behavior in SFR safety analysis, an advanced model for MOX(Mixed OXide) fuel pin thermo-mechanical performance under operation conditions has been developed and validated. This validation involved simulating the SUPERFACT-1 fast reactor irradiation experiment using ISAA-Na. The resulting calculation data were compared with experimental measurements and the outcomes of other codes dedicated to fuel pin performance analysis (TRANSURANUS, MACROS, andGERMINAL) to assess the fuel pin thermo-mechanical model. The overall code assessment reveals that the model achieves satisfactory performance in simulating fuel pin behavior, with predicted errors falling within an acceptable range.
Abstract
Preparations are currently underway to develop ISAA-Na code, a sodium-cooled fast reactors (SFRs) specific version of an integrated severe accident analysis code. This version aims to evaluate both protected and unprotected accidents in SFRs. Given the crucial role of fuel pin thermo-mechanical behavior in SFR safety analysis, an advanced model for MOX(Mixed OXide) fuel pin thermo-mechanical performance under operation conditions has been developed and validated. This validation involved simulating the SUPERFACT-1 fast reactor irradiation experiment using ISAA-Na. The resulting calculation data were compared with experimental measurements and the outcomes of other codes dedicated to fuel pin performance analysis (TRANSURANUS, MACROS, andGERMINAL) to assess the fuel pin thermo-mechanical model. The overall code assessment reveals that the model achieves satisfactory performance in simulating fuel pin behavior, with predicted errors falling within an acceptable range.
참고문헌 (36)
- Xu Mi, "Safety properties of China experimental fast reactor", Chin. J. Nucl. Eng. Sci, (2), pp.116~126 (2011)
- Xu Mi, "Fast reactor technology R&D activities in China", Nucl. Eng. Technol, 39(3), pp.187~192 (2007)
- Foundational Development of an Advanced Nuclear Reactor Integrated Safety Code, "K. Clarno", Sandia National Laboratories (SNL), ~, pp.No. SAND2010-0878 (2010)
- A. M. Tentner, "SAS4A LMFBR Whole Core Accident Analysis Code (No. CONF-850410", ~ (1985)
- S. Kondo, "SIMMER-III : an advanced computer program for LMFBR severe accident analysis", ANP’92 International Conference on Design and Safety of Advanced Nuclear Power Plants (1992)
- Y. Tobita, "Development of SAME-2: A Connection Code between SAS4A and SIMMER-3", ~ (2002)
- G. Bandini, "ASTEC-Na code : thermal-hydraulic model validation and benchmarking with other codes", Ann. Nucl. Energy, 119, pp.427~439 (2018)[DOI]
- S. Perez-Martin, "Development and assessment of ASTEC-Na fuel pin thermo-mechanical models performed in the European JASMIN project", Ann. Nucl. Energy, 119, pp.454~473 (2018)[DOI]
- V. Matuzas, "Extension of ASTEC-Na capabilities for simulating reactivity effects in sodium cooled fast reactor", Ann. Nucl. Energy, 119, pp.440~453 (2018)[DOI]
- L. S. Lebel, "New developments in modeling sodium spray fires with ASTEC-Na", Ann. Nucl. Energy, 119, pp.23~35 (2018)[DOI]
- N. A. Mosunova, "The EUCLID", V1 integrated code for safety assessment of liquid metal cooled fast reactors. Part 1 : basic models, 2018, pp.65(5), 304~316
- V. M. Alipchenkov, "The EUCLID", V1 integrated code for safety assessment of liquid metal cooled fast reactors. Part 2 : validation and verification, 2018, pp.65(9), 627~640
- Final Report on LDRD Project: Coupling Strategies for Multi-Physics Applications, "R. Hooper", Sandia National Laboratories, ~, pp.SAND2007-7146 (2007)
- P. Gao, "Development of mechanistic cladding rupture model for severe accident analysis and application in PHEBUS FPT3 experiment", Nucl. Eng. Technol (2021)[DOI]
- S. Tang, "Research on the performance enhancement of accident tolerant fuel for eliminating core melting based on ISAA", Ann. Nucl. Energy, 186, pp.109747~ (2023)[DOI]
- J. B. Ainscough, "Isothermal grain growth kinetics in sintered UO2 pellets", J. Nucl. Mater, 49(2), pp.117~128 (1973)[DOI]
- A. Magni, "Nuclear Power Plant Design and Analysis Codes", pp.161~205 (2021), Woodhead Publishing
- B. Michel, "Nuclear Power Plant Design and Analysis Codes", pp.207~233 (2021), Woodhead Publishing
- S. E. Lemehov, "MACROS benchmark calculations and analysis of fission gas release in MOX with high content of plutonium", Prog. Nucl. Energy, 57, pp.117~124 (2012)[DOI]
- Nuclear Systems Materials Handbook, "M. F. Marchbanks", Hanford Engineering Development Lab, ~, pp.No. HEDL-SA–871 (1976)
- Fission-energy Release for 16 Fissioning Nuclides, "R. Sher", Stanford Univ, ~, pp.No. EPRINP-1771 (1981)
- The SAS4A, "SASSYS-1 Safety Analysis Code System", T.H. Fanning, 2017, No. ANL[DOI]
- M. Bober, "Advances in Nuclear Science and Technology", pp.121~179 (1973), Academic Press
- C. F. Clement, "The movement of lenticular pores in UO2 nuclear fuel elements", J. Nucl. Mater, 68(1), pp.63~68 (1977)
- W. J. Lackey, "Porosity and actinide redistribution during irradiation of (U, Pu)O2", Nucl. Technol, 16(1), pp.120~142 (1972)[DOI]
- LIFE-I: A FORTRAN-IV COMPUTER CODE FOR THE PREDICTION OF FAST-REACTOR FUEL-ELEMENT BEHAVIOR, "V. Z. Jankus", Argonne National Lab, ~, pp.No. ANL-7736 (1970)
- Joint CEA, "JRC Synthesis Report of the Experiment SUPERFACT 1", J. F. Babelot, 1999, Report JRC-ITU-TN-99[DOI]
- Some Specific Aspects of Homogeneous Am and Np Based Fuels Transmutation through the Outcomes of the Superfact Experiment in PHENIX Fast Reactor, "C. Prunier", CEA Centre d’Etudes de Cadarache, ~, pp.No. CEA-CONF–11506 (1993)
- L. Luzzi, "Assessment of three European fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment", Nucl. Eng. Technol, 53(10), pp.3367~3378 (2021)[DOI]
- L. Luzzi, "Assessment of INSPYRE-extended fuel performance codes against the SUPERFACT1 fast reactor irradiation experiment", Nucl. Eng. Technol, 55(3), pp.884~894 (2023)[DOI]
- A. H. Squillacote, "The Paraview Guide" (2007), Kitware
- K. Lassmann, "Transuranus Handbook", Document Number Version, 1, ~ (1975)
- A. Gallais-During, "Outcomes of the PELGRIMM project on Am-bearing fuel in pelletized and spherepac forms", J. Nucl. Mater, 512, pp.214~226 (2018)[DOI]
- M. Lainet, "GERMINAL, a fuel performance code of the PLEIADES platform to simulate the in-pile behaviour of mixed oxide fuel pins for sodium-cooled fast reactors", J. Nucl. Mater, 516, pp.30~53 (2019)[DOI]
- G. Pastore, "Physics-based modelling of fission gas swelling and release in UO2 applied to integral fuel rod analysis", Nucl. Eng. Des, 256, pp.75~86 (2013)[DOI]
- G. Kussmaul, "Science and Technology of Fast Reactor Safety", pp.1~103 (1987), Thomas Telford Publishing
