High-fidelity numerical investigation on structural integrity of SFR fuel cladding during design basis events
2
조회
0
다운로드
0
인용
0
참고문헌
키워드
초록
A high-fidelity numerical analysis methodology was proposed for evaluating the fuel rod cladding integrity of a Prototype Gen IV Sodium Fast Reactor (PGSFR) during normal operation and Design basis events (DBEs). The MARS-LMR code, system transient safety analysis code, was applied to analyze the DBEs. The results of the MARS-LMR code were used as boundary condition for a 3D computational fluid dynamics (CFD) analysis. The peak temperatures considering HCFs satisfied the cladding temperature limit. The temperature and pressure distributions were calculated by ANSYS CFX code, and applied to structural analysis. Structural analysis was performed using ANSYS Mechanical code. The seismic reactivity insertion SSE accident among DBEs had the highest peak cladding temperature and the maximum stress, as the value of 87 MPa. The fuel cladding had over 40 % safety margin, and the strain was below the strain limit. Deformation behavior was elucidated for providing relative coordinate data on each active fuel rod center. Bending deformation resulted in a flower shape, and bowing bundle did not interact with the duct of fuel assemblies. Fuel rod maximum expansion was generated with highest stress. Therefore, it was concluded that the fuel rod cladding of the PGSFR has sufficient structural safety margin during DBEs.
Abstract
A high-fidelity numerical analysis methodology was proposed for evaluating the fuel rod cladding integrity of a Prototype Gen IV Sodium Fast Reactor (PGSFR) during normal operation and Design basis events (DBEs). The MARS-LMR code, system transient safety analysis code, was applied to analyze the DBEs. The results of the MARS-LMR code were used as boundary condition for a 3D computational fluid dynamics (CFD) analysis. The peak temperatures considering HCFs satisfied the cladding temperature limit. The temperature and pressure distributions were calculated by ANSYS CFX code, and applied to structural analysis. Structural analysis was performed using ANSYS Mechanical code. The seismic reactivity insertion SSE accident among DBEs had the highest peak cladding temperature and the maximum stress, as the value of 87 MPa. The fuel cladding had over 40 % safety margin, and the strain was below the strain limit. Deformation behavior was elucidated for providing relative coordinate data on each active fuel rod center. Bending deformation resulted in a flower shape, and bowing bundle did not interact with the duct of fuel assemblies. Fuel rod maximum expansion was generated with highest stress. Therefore, it was concluded that the fuel rod cladding of the PGSFR has sufficient structural safety margin during DBEs.
참고문헌 (57)
- K. L. Lee, "A preliminary safety analysis for the prototype gen IV sodium-cooled fast reactor", Nucl. Eng. Technol, 48, pp.1071~1082 (2016)
- D. Hahn, "Conceptual design of the sodium cooled fast reactor KALIMER-600", Nucl. Eng. Technol, 39, pp.193~206 (2007)
- D. Hahn, "Advanced SFR design concepts and R&D activities", Nucl. Eng. Technol, 41, pp.427~446 (2009)
- System Description of Prototype Gen-IV Sodium-cooled Fast Reactor, , Korea Atomic Energy Research Institute, ~, pp.SFR-000-SP-403e001 (2015)
- Preliminary Safety Information Document of PGSFR, , Korea Atomic Energy Research Institute, ~ (2015)
- J. Yoo, "Overall system description and safety characteristics of prototype Gen IV sodium cooled fast reactor in Korea", Nucl. Eng. Technol, 48, pp.1059~1070 (2016)
- T. W. Kim, "Overview of PGSFR project", Am. Nucl. Soc, 114(1), pp.698~699 (2016)
- Hyung-Kyu Kim, "Mechanical design of a sodium cooled fast reactor fuel assembly in Korea : normal operation condition", Nucl. Eng. Des, 346, pp.67~74 (2019)
- Analysis of Heat Transfer and Thermal Stress on New Shape of Hanaro Fuel, "Yun Chang Lee", Transaction of the Korean Nuclear society, ~ (2004)
- Study on the Standard Establishment for the Integrity Assessment of Nuclear Fuel Cladding Materials, , KAERI, ~
- Du Peng, "Thermal-hydraulics analysis of flow blockage events for fuel assembly n a sodium-cooled fast reactor", Int. J. Heat Mass Tran, 138, pp.496~507 (2019)
- NSSS Design and Validation of Prototype GEN-IV Sodium Cooled Fast Reactor, , KAERI, ~ (2018)
- D. L. Porter, "Fuel performance design basis for the versatile test reactor", American Nuclear Society, 196, pp.110~122 (2021)
- K. H. Yoon, "Component design and accident analysis fuel assembly for prototype GEN-IV sodium-cooled fast reactor", Nucl. Eng. Des, 340, pp.133~145 (2018)
- H. S. Yoon, "Two- and three-dimensional analysis comparison of nozzles due to internal pressure, thermal load and external load", J. Comput. Struct. Eng. Inst. Korea, 28(3), pp.283~291 (2015)
- K. S. Kim, "Thermal Stress Analysis for the Plate Type Fuel Assembly", Transactions of the Korean Nuclear Society Spring Meeting (2010)
- Cheong Uk Kim, "Stress analysis of pressure vessels in nuclear power plants(Part II : stress analysis of tapered cylinder in the shell-head junction", J. Mech. Sci. Technol, 16(2), pp.202~209 (1976)
- Jeong, Yeong Shin, "Thermal-hydraulic Effect of Pattern of Wire-Wrap Spacer in 19-pin Rod Bundel for SFR Fueal Assembly", Korean Nuclear Society Spring Meeting (2016)
- J. -J. Jeong, "Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1", Ann. Nucelar Energy, 26, pp.1611~1642 (1999)
- Development of MARS-LMR and Steady-State Calculation for KALIMER-600, "K.S. Ha", Korea Atomic Energy Research Institute, ~, pp.Technical Report No. KAERI (2007), TR-3418[DOI]
- A Soft-Sphere Model for Liquid Metal, "David A. Young", ~, pp.UCRL-53252 (1977)
- SSC-K User’s Manual, "Y. M. Kwon", ~, pp.KAERI (2000), TR-1619[DOI]
- N. E. Todreas, "NUCLEAR SYSTEM I", MIT, Hemisphere Publishing Co
- "ANSYS CFX-Solver Theory Guide, ANSYS CFX Release 15.0", 2013, ~
- "ANSYS CFX-Solver Modeling Guide,ANSYS CFX Release 15.0", ~ (2013)
- J. Smagorinsky, "General circulation experiments with the primitive equations, I the basic experiment", Mon. Weather Rev, 91, pp.99~164 (1963)
- Turbulence Modeling Validation Testing and Development, "J.E. Bardina", NASA, ~, pp.NASA TM-110446 (1997)
- D. C. Wilcox, "Re-assessment of the scale-determining equation for advanced turbulence models", AIAA J, 26(11), pp.1299~1310 (1988)
- F. R. Menter, "Two-equation eddy-viscosity turbulence models for engineering applications", AIAA J, 32(8), pp.1598~1605 (1994)
- Dimensions of PGSFR Fuel Assembly, "S. R. Choi", KAERI, ~, pp.SFR-IOC-R (2014), M-14-003
- "ANSYS Mechanical APDL Manual, Theory Reference (Chapter 15.7 : Spectrum Analysis), Rev. 15.0", ANSYS Inc, ~
- ANSYS Workbench-Mechanical Introduction 12.0 Chapter 4. Static Structural Analysis, , ~
- Jae-ho jeong, "Comparative safety analysis with MARS-LMR and SAS4A", SASSYS-1 codes for PGSFR, 2019, ~
- Ji-Youn kim, "Safety analysis of one pump seizure accident for 2017 PGSFR related to effect of pump seizure time and loss of off-site power", Transaction of the Korean Nuclear Society Autumn Meeting (2017)
- Jae-Ho Jeong, "RANS based CFD methodology for a real scale 217-pin wire-wrapped fuel assembly of KAERI PGSFR", Nucl. Eng. Des, 313, pp.470~485 (2017)
- Min seop song, "Numerical investigation on vortex behavior n wire-wrapped fuel assembly for a sodium fast reactor", Nucealr Engineering and Technology, 51, pp.665~675 (2019)
- Jae-Ho Jeong, "Thermal-hydraulic effect of wire spacer in a wire-wrapped fuel bundles for SFR", Nucl. Eng. Des, 320, pp.28~43 (2017)
- Donny Hartanto, "A comparative physics study for an innovative sodium-cooled fast reactor (iSFR)", Int. J. Energy Res, pp.151~162 (2018)
- J. D. Hales, "BISON theory manual the equations behind nuclear fuel analysis", BISON Release, 1(3) (2016)
- Design Report of Core T, "H Design", Sun-Rak Choi, 2016, SFR-120-DR-462-001
- Report on Evaluation of Design Limiting Factors for Core of a Pooltype Reactor, "Sun-Rak Choi", ~, pp.SFR-120-DR-486-015 (2016)
- PGSFR Fuel Design Methodology for Steady-State Normal Operation, "Choen jeon sik", ~, pp.SFR-160-FP-462-001 (2014)
- Uncertainty Evaluation Report for the Design of a Liquid Metal Fast Reactor Core, "Sun-Rak Choi", ~, pp.SFR-120-DR-486-017 (2016)
- F. R. Menter, "Two-equation eddy-viscosity turbulence models for engineering applications", AIAA J, 32, pp.1598~1605 (1994)
- General safety design criteria for a liquid metal reactor nuclear power plant, , ~, pp.ANSI (1989), ANS 54.1
- "Safety Review Guidelines for Light Water Reactors", Korea institute of nuclear safety, ~ (2009)
- CRBRP Core Assemblies Hot Channel Factors Preliminary Analysis, "Y. A. Friedland", Applied Technology, ~, pp.CRBRP-ARP-0050 (1980)
- Assessment of FFTF Hot Channel Factors, "J. Muraoka", ~, pp.HEDL-TI-75226 (1976)
- Hot Channel Factors for Thermal Analysis, "Y. I. Chang", Argonne National Laboratory, ~, pp.ANL-KAERI-SFR-13-15 (2013)
- Evaluation and Selection of Hot Channel(Peaking)Factors for Research Reactor Applications, "W. L. Woodruff", Argonne National Laboratory, ~ (1997)
- Fuel Assembly Mechanical Design Report, "H. K. Kim", KAERI, ~, pp.SFR-170-FP-475-020 (2018)
- "Solid Mechanics II, C1.1 Membrane Stress Equation", ~
- N. Noda, "Thermal Stresses" (2003), Taylor & Francis
- Study on the Standard Establishment for the Integrity Assessment of Nuclear Fuel Cladding Materials, "S. -S. Kang", ~, pp.Technical Report KAERI (2008), CM-1034[DOI]
- Nuclear Reactor Analysis, "J. J. Duderstadt", ~
- X. Ye, "Full-scale finite element analysis of deformation and contact of a wire-wrapped fuel bundle subject to realistic thermal and irradiation conditions", Nucl. Eng. Des, 364, pp.110676~ (2020)
- P. Du, "Thermal-hydraulics analysis of flow blockage events for fuel assembly in a sodium-cooled fast reactor", Int. J. Heat Mass Tran, 138, pp.496~507 (2019)