稳态堆芯多物理耦合系统CSSS V1.0的研发
|
摘要
充分考虑反应堆堆芯中子学物理、热工水力、燃料等专业的相互耦合过程,将先进节块法堆芯中子学计算软件NACK V1.0、热工水力子通道软件CORTH V2.0、燃料棒性能分析软件FUPAC V1.1进行集成耦合,得到稳态堆芯多物理耦合模拟设计分析系统CSSS V1.0,可计算典型压水堆的稳态运行物理、热工、燃料等专业参数。通过NEACRP-L-335压水堆基准问题验证计算,CSSS V1.0系统的计算结果与国际基准PARCS程序总体符合较好。
The reactor core is the coupled result of the multi-physics including the neutronics, thermal-hydraulics, fuel and so on. Coupling the advanced nodal core key program NACK V1.0, core thermal-hydraulic analysis program CORTH V2.0 and fuel rod performance analysis code FUPAC V1.1, the reactor core steady simulation system CSSS V1.0 was got. CSSS V1.0 was used to simulate typical pressurized water reactor core. The calculation result of pressurized water reactor benchmark problem NEACRP-L-335 shows that CSSS V1.0 is in good agreement with benchmark program PARCS.
The reactor core is the coupled result of the multi-physics including the neutronics, thermal-hydraulics, fuel and so on. Coupling the advanced nodal core key program NACK V1.0, core thermal-hydraulic analysis program CORTH V2.0 and fuel rod performance analysis code FUPAC V1.1, the reactor core steady simulation system CSSS V1.0 was got. CSSS V1.0 was used to simulate typical pressurized water reactor core. The calculation result of pressurized water reactor benchmark problem NEACRP-L-335 shows that CSSS V1.0 is in good agreement with benchmark program PARCS.
引文
[1] IVANOV K N,MACIAN-JUAN R,IRANI A,et al.Features and performance of a coupled three-dimensional thermal-hydraulic/kinetics TRAC-PF1/NEM pressurized water reactor (PWR) analysis code[J].Annals of Nuclear Energy,1999,26:1 407-1 417.
[2] KLIEM S,KOZMENKOV Y,HOHNE T,et al.Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations[J].Progress in Nuclear Energy,2006,48:830-848.
[3] KOZMENKOV Y,KLIEM S,GRUNDMANN U,et al.Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET[J].Nuclear Engineering and Design,2007,237:1 938-1 951.
[4] 赵文博.瞬态节块格林函数方法及其与热工-水力耦合研究[D].北京:清华大学,2012.
[5] 安萍.多几何堆芯中子学先进节块法及其在物理热工耦合技术的应用研究[D].成都:中国核动力研究设计院,2014.
[6] 安萍,姚栋.六角形格林函数节块法[J].原子能科学技术,2014,48(4):667-672.AN Ping,YAO Dong.Hexagonal green function nodal method[J].Atomic Energy Science and Technology,2014,48(4):667-672(in Chinese).
[7] 刘余,谭长禄,潘俊杰,等.子通道分析软件CORTH的研发[J].核动力工程,2017,38(6):157-162.LIU Yu,TAN Changlu,PAN Junjie,et al.Development of self-reliant subchannel analysis code CORTH[J].Nuclear Power Engineering,2017,38(6):157-162(in Chinese).
[8] 涂晓兰,柴晓明,尹强,等.燃料棒性能分析软件FUPAC中燃料棒径向功率密度分布模型研发[J].核动力工程,2014,35(2):197-199.TU Xiaolan,CHAI Xiaoming,YIN Qiang,et al.Research and development of radial power density distribution model for FUPAC[J].Nuclear Power Engineering,2014,35(2):197-199(in Chinese).
[9] PARCS Workshop.Nuclear power plant simiulation with coupled neutronics/thermal-hydraulics codes[C]//PHYSOR 2010.Pittsburgh,Pennsylvania,USA:American Nuclear Society,2010.
[10] DOWNAR T,LEE D.User manual for the PARCS kinetics simulator module[R].US:School of Nuclear Engineering Purdue University,2000.
[11] DOWNAR T,LEE D,XU Y,et al.PARCS v2.6 core neutronics simulator theory manual[R].US:NRC,2007.
[12] FINNEMANN H.NEACRP 3-D LWR transient benchmark[R].Germany:[s.n.],1991.
[13] FINNEMANN H.Results of LWR core transient benchmarks[R].Germany:[s.n.],1993.
[14] FRAIKIN R.NEA-NSC PWR benchmarks review[C]∥Specialists Meeting on LWR Core Transient Benchmarks.Paris:NEA/NSC,1995.
[15] KNIGHT M P,BRYCE P.Derivation of a refined PANTHER solution to the NEACRP PWR rod-ejection transient[C]//Proceedings of the Joint International Conference on Mathematical Method Applications.Saratoga:[s.n.],1997.
[2] KLIEM S,KOZMENKOV Y,HOHNE T,et al.Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations[J].Progress in Nuclear Energy,2006,48:830-848.
[3] KOZMENKOV Y,KLIEM S,GRUNDMANN U,et al.Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET[J].Nuclear Engineering and Design,2007,237:1 938-1 951.
[4] 赵文博.瞬态节块格林函数方法及其与热工-水力耦合研究[D].北京:清华大学,2012.
[5] 安萍.多几何堆芯中子学先进节块法及其在物理热工耦合技术的应用研究[D].成都:中国核动力研究设计院,2014.
[6] 安萍,姚栋.六角形格林函数节块法[J].原子能科学技术,2014,48(4):667-672.AN Ping,YAO Dong.Hexagonal green function nodal method[J].Atomic Energy Science and Technology,2014,48(4):667-672(in Chinese).
[7] 刘余,谭长禄,潘俊杰,等.子通道分析软件CORTH的研发[J].核动力工程,2017,38(6):157-162.LIU Yu,TAN Changlu,PAN Junjie,et al.Development of self-reliant subchannel analysis code CORTH[J].Nuclear Power Engineering,2017,38(6):157-162(in Chinese).
[8] 涂晓兰,柴晓明,尹强,等.燃料棒性能分析软件FUPAC中燃料棒径向功率密度分布模型研发[J].核动力工程,2014,35(2):197-199.TU Xiaolan,CHAI Xiaoming,YIN Qiang,et al.Research and development of radial power density distribution model for FUPAC[J].Nuclear Power Engineering,2014,35(2):197-199(in Chinese).
[9] PARCS Workshop.Nuclear power plant simiulation with coupled neutronics/thermal-hydraulics codes[C]//PHYSOR 2010.Pittsburgh,Pennsylvania,USA:American Nuclear Society,2010.
[10] DOWNAR T,LEE D.User manual for the PARCS kinetics simulator module[R].US:School of Nuclear Engineering Purdue University,2000.
[11] DOWNAR T,LEE D,XU Y,et al.PARCS v2.6 core neutronics simulator theory manual[R].US:NRC,2007.
[12] FINNEMANN H.NEACRP 3-D LWR transient benchmark[R].Germany:[s.n.],1991.
[13] FINNEMANN H.Results of LWR core transient benchmarks[R].Germany:[s.n.],1993.
[14] FRAIKIN R.NEA-NSC PWR benchmarks review[C]∥Specialists Meeting on LWR Core Transient Benchmarks.Paris:NEA/NSC,1995.
[15] KNIGHT M P,BRYCE P.Derivation of a refined PANTHER solution to the NEACRP PWR rod-ejection transient[C]//Proceedings of the Joint International Conference on Mathematical Method Applications.Saratoga:[s.n.],1997.