基于Dragon与Donjon程序的液态熔盐实验堆临界计算与分析
|
摘要
基于组件输运程序Dragon与堆芯节块法程序Donjon,对包含有上下熔盐腔室、控制棒、实验孔道与中子源孔道的液态熔盐实验堆堆芯进行了计算与分析,给出了液态熔盐实验堆不同组件的等效均匀化模型。根据液态熔盐实验堆特性将中子能群划分为5种少群能群结构,基于所划分的每一种少群能群结构,对单根控制棒与不同控制棒组插入堆芯后的有效增殖因数和控制棒价值进行了计算分析。结果表明,7群能群结构具有更好的计算结果。基于7群能群结构开展了堆芯径向与纵向功率分布,以及控制棒拔出后堆芯的温度反应性系数计算分析,其计算结果与MCNP5计算结果相近,证明了模型等效的合理性以及Dragon和Donjon程序对液态熔盐实验堆的适用性。
Based on the lattice code Dragon and nodal code Donjon for reactor core, the liquid molten salt experimental reactor(LMSER), which contains the top and bottom molten salt chambers, control rod, experimental hole and neutron source channel, was analyzed in this study. Equivalent homogenization models of different assemblies of LMSER were given. The neutron energy group of LMSER was divided into 5 kinds of broad energy group structures according to the characteristics of LMSER. For each kind of broad energy group structure, the effective multiplication factor and control rod worth with single control rod and different control rod groups inserted into the core were investigated. The results show that more precise calculation results are obtained by taking the 7-group structure. Thereafter, the radial and axial core power distributions, and the temperature reactivity coefficient without control rod insertion were calculated based on the 7-group structure. The calculation result agrees well with that calculated by MCNP5, which therefore proves the reasonability of equivalent homogenization model and the feasibility of Dragon and Donjon codes for LMSER calculation.
Based on the lattice code Dragon and nodal code Donjon for reactor core, the liquid molten salt experimental reactor(LMSER), which contains the top and bottom molten salt chambers, control rod, experimental hole and neutron source channel, was analyzed in this study. Equivalent homogenization models of different assemblies of LMSER were given. The neutron energy group of LMSER was divided into 5 kinds of broad energy group structures according to the characteristics of LMSER. For each kind of broad energy group structure, the effective multiplication factor and control rod worth with single control rod and different control rod groups inserted into the core were investigated. The results show that more precise calculation results are obtained by taking the 7-group structure. Thereafter, the radial and axial core power distributions, and the temperature reactivity coefficient without control rod insertion were calculated based on the 7-group structure. The calculation result agrees well with that calculated by MCNP5, which therefore proves the reasonability of equivalent homogenization model and the feasibility of Dragon and Donjon codes for LMSER calculation.
引文
[1] MERLE L,HEUER D,ALLIBERT M,et al.Optimization and simplification of the concept of non-moderated thorium molten salt reactor[C]//Proceedings of the International Conference on the Physics of Reactors.[S.l.]:[s.n.],2008.
[2] MERLE L,HEUER D,ALLIBERT M,et al.Minimizing the fissile inventory of the molten salt fast reactor[C]//Proceedings of the International Conference Advances in Nuclear Fuel Management Ⅳ.[S.l.]:[s.n.],2009.
[3] MERLE L,HEUER D,ALLIBERT M,et al.Optimizing the burning efficiency and the deployment capacities of the molten salt fast reactor[C]//Proceedings of the International Conference Global.[S.l.]:[s.n.],2009.
[4] MATHIEU L,HEUER D,MERLE-LUCOTTE E,et al.Possible configurations for the thorium molten salt reactor and advantages of the fast nonmoderated version[J].Nuclear Science and Engineering,2009,161(1):78-89.
[5] IGNATIEV V,SCHIKORR M,FEYNBERG O,et al.Calculation of safety related parameters for Na,Li,Be/F MOSART concept[C]//The 3rd RCM of IAEA CRP on Studies of Innovative Reactor Technology Options for Effective Incineration of Radioactive Waste,Domain Ⅳ.India:Indira Gandhi Centre for Atomic Research (IGCAR),2007.
[6] HAUBENREICH P N,ENGEL J R.Experience with the molten salt reactor experiment[J].Nuclear Applications and Technology,1970,8(2):118-136.
[7] ROBERTSON R C.Conceptual design study of a single-fluid molten-salt breeder reactor,Technical Report ORNL-4541[R].[S.l.]:[s.n.],1971.
[8] 江绵恒,徐洪杰,戴志敏.未来先进核裂变能——TMSR核能系统[J].中国科学院院刊,2012,27(3):366-374.JIANG Mianheng,XU Hongjie,DAI Zhimin.Advanced fission energy program—TMSR nuclear energy system[J].Bulletin of Chinese Academy of Sciences,2012,27(3):366-374(in Chinese).
[9] X-5 Monte Carlo Team.MCNP:A general Monte Carlo N-particle transport code,Volume Ⅰ:User guide[R].New Mexico:Los Alamos National Library,2003.
[10] SCALE:A modular code system for performing standardized computer analyses for licensing evaluations,Version6.1,ORNL/TM-2005/39[R].USA:ORNL,2009.
[11] WANG K,LI Z,SHE D,et al.RMC:A Monte Carlo code for reactor core analysis[J].Annals of Nuclear Energy,2015,82:121-129.
[12] ROY R,MARLEAU G,TAJMOUATI J,et al.Modelling of CANDU reactivity control devices with the lattice code DRAGON[J].Annual of Nuclear Energy,1994,21(2):115-132.
[13] DAHMANI M,MARLEAU G,TELLIER R L.Modeling reactivity devices for advanced CANDU reactors using the code DRAGON[J].Annals of Nuclear Energy,2008,35(5):804-812.
[14] 成红武,谢仲生,章宗耀.围板和反射层的等效均匀化常数的计算[J].核科学与工程,1996,16(2):97-103.CHENG Hongwu,XIE Zhongsheng,ZHANG Zongyao.Calculation of equivalent homogenized constants of the baffle and reflector[J].Chinese Journal of Nuclear Science and Engineering,1996,16(2):97-103(in Chinese).
[15] 曹良志,谢仲生,吴宏春,等.先进的压水堆燃料管理计算方法研究及软件研制[J].西安交通大学学报,2003,37(5):519-522.CAO Liangzhi,XIE Zhongsheng,WU Hongchun,et al.Research on advanced pressurized water reactor fuel management calculation method and program development[J].Journal of Xi’an Jiaotong University,2003,37(5):519-522(in Chinese).
[16] 马云帆,吴宏春,曹良志.压水堆嬗变长寿命裂变产物的弹棒事故安全性研究[J].原子能科学技术,2017,51(3):439-443.MA Yunfan,WU Hongchun,CAO Liangzhi.Study of rod ejection accident safety of PWR for long-lived fission product transmutation[J].Atomic Energy Science and Technology,2017,51(3):439-443(in Chinese).
[17] 郭炯,李富.不连续因子计算强吸收体区域的改进[J].原子能科学技术,2013,47(增刊):33-37.GUO Jiong,LI Fu.Improvement of discontinuous factor for strong absorber region[J].Atomic Energy Science and Technology,2013,47(Suppl.):33-37(in Chinese).
[18] 周旭华,李富,王登营,等.不连续因子应用于高温气冷堆三维扩散计算[J].原子能科学技术,2009,43(3):237-241.ZHOU Xuhua,LI Fu,WANG Dengying,et al.Discontinuity factor applied in three-dimensional diffusion calculation for high-temperature gas-cooled reactor[J].Atomic Energy Science and Technology,2009,43(3):237-241(in Chinese).
[19] MARLEAU G,HEBERT A,ROY R.A user’s guide for DRAGON Version 5,Report IGE-335[R].Ecole Polytechnique de Montreal:Institut de Genie Nucleaire,2017.
[20] HéBERT A.DRAGON5:Designing computational schemes dedicated to fission nuclear reactors for space[C]//International Conference on Nuclear and Emerging Technologies for Space.Albuquerque,N M:[s.n.],2013.
[21] HEBERT A,SEKKI D,CHAMBON R.A user’s guide for DONJON,Version5,IGE-344[R].Ecole Polytechnique de Montreal:Institut de Genie Nucleaire,2017.
[22] HéBERT A.TRIVAC:A modular diffusion code for fuel management and design applications[J].Nuclear Journal of Canada,1987,1:325-331.
[23] RICHARD S.Assembly homogenization techniques for core calculations[J].Progress in Nuclear Energy,2009,51(1):14-31.
[24] REITSMA F,STRYDOM G,de HAAS J B M,et al.The PBMR steady-state and coupled kinetics core thermal-hydraulics benchmark test problems[J].Nuclear Engineering and Design,2006,236(5-6):657-668.
[25] HAN J S.Sensitivity study on the energy group structure for high temperature reactor analysis[M].USA:The Pennsylvania State University,2008.
[26] ZOU C Y,CAI X Z,JIANG D Z,et al.Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor[J].Nuclear Engineering and Design,2015,281:114-120.
[27] CUI D Y,LI X X,XIA S P,et al.Possible scenarios for the transition to thorium fuel cycle in molten salt reactor by using enriched uranium[J].Progress in Nuclear Energy,2018,104:75-84.
[2] MERLE L,HEUER D,ALLIBERT M,et al.Minimizing the fissile inventory of the molten salt fast reactor[C]//Proceedings of the International Conference Advances in Nuclear Fuel Management Ⅳ.[S.l.]:[s.n.],2009.
[3] MERLE L,HEUER D,ALLIBERT M,et al.Optimizing the burning efficiency and the deployment capacities of the molten salt fast reactor[C]//Proceedings of the International Conference Global.[S.l.]:[s.n.],2009.
[4] MATHIEU L,HEUER D,MERLE-LUCOTTE E,et al.Possible configurations for the thorium molten salt reactor and advantages of the fast nonmoderated version[J].Nuclear Science and Engineering,2009,161(1):78-89.
[5] IGNATIEV V,SCHIKORR M,FEYNBERG O,et al.Calculation of safety related parameters for Na,Li,Be/F MOSART concept[C]//The 3rd RCM of IAEA CRP on Studies of Innovative Reactor Technology Options for Effective Incineration of Radioactive Waste,Domain Ⅳ.India:Indira Gandhi Centre for Atomic Research (IGCAR),2007.
[6] HAUBENREICH P N,ENGEL J R.Experience with the molten salt reactor experiment[J].Nuclear Applications and Technology,1970,8(2):118-136.
[7] ROBERTSON R C.Conceptual design study of a single-fluid molten-salt breeder reactor,Technical Report ORNL-4541[R].[S.l.]:[s.n.],1971.
[8] 江绵恒,徐洪杰,戴志敏.未来先进核裂变能——TMSR核能系统[J].中国科学院院刊,2012,27(3):366-374.JIANG Mianheng,XU Hongjie,DAI Zhimin.Advanced fission energy program—TMSR nuclear energy system[J].Bulletin of Chinese Academy of Sciences,2012,27(3):366-374(in Chinese).
[9] X-5 Monte Carlo Team.MCNP:A general Monte Carlo N-particle transport code,Volume Ⅰ:User guide[R].New Mexico:Los Alamos National Library,2003.
[10] SCALE:A modular code system for performing standardized computer analyses for licensing evaluations,Version6.1,ORNL/TM-2005/39[R].USA:ORNL,2009.
[11] WANG K,LI Z,SHE D,et al.RMC:A Monte Carlo code for reactor core analysis[J].Annals of Nuclear Energy,2015,82:121-129.
[12] ROY R,MARLEAU G,TAJMOUATI J,et al.Modelling of CANDU reactivity control devices with the lattice code DRAGON[J].Annual of Nuclear Energy,1994,21(2):115-132.
[13] DAHMANI M,MARLEAU G,TELLIER R L.Modeling reactivity devices for advanced CANDU reactors using the code DRAGON[J].Annals of Nuclear Energy,2008,35(5):804-812.
[14] 成红武,谢仲生,章宗耀.围板和反射层的等效均匀化常数的计算[J].核科学与工程,1996,16(2):97-103.CHENG Hongwu,XIE Zhongsheng,ZHANG Zongyao.Calculation of equivalent homogenized constants of the baffle and reflector[J].Chinese Journal of Nuclear Science and Engineering,1996,16(2):97-103(in Chinese).
[15] 曹良志,谢仲生,吴宏春,等.先进的压水堆燃料管理计算方法研究及软件研制[J].西安交通大学学报,2003,37(5):519-522.CAO Liangzhi,XIE Zhongsheng,WU Hongchun,et al.Research on advanced pressurized water reactor fuel management calculation method and program development[J].Journal of Xi’an Jiaotong University,2003,37(5):519-522(in Chinese).
[16] 马云帆,吴宏春,曹良志.压水堆嬗变长寿命裂变产物的弹棒事故安全性研究[J].原子能科学技术,2017,51(3):439-443.MA Yunfan,WU Hongchun,CAO Liangzhi.Study of rod ejection accident safety of PWR for long-lived fission product transmutation[J].Atomic Energy Science and Technology,2017,51(3):439-443(in Chinese).
[17] 郭炯,李富.不连续因子计算强吸收体区域的改进[J].原子能科学技术,2013,47(增刊):33-37.GUO Jiong,LI Fu.Improvement of discontinuous factor for strong absorber region[J].Atomic Energy Science and Technology,2013,47(Suppl.):33-37(in Chinese).
[18] 周旭华,李富,王登营,等.不连续因子应用于高温气冷堆三维扩散计算[J].原子能科学技术,2009,43(3):237-241.ZHOU Xuhua,LI Fu,WANG Dengying,et al.Discontinuity factor applied in three-dimensional diffusion calculation for high-temperature gas-cooled reactor[J].Atomic Energy Science and Technology,2009,43(3):237-241(in Chinese).
[19] MARLEAU G,HEBERT A,ROY R.A user’s guide for DRAGON Version 5,Report IGE-335[R].Ecole Polytechnique de Montreal:Institut de Genie Nucleaire,2017.
[20] HéBERT A.DRAGON5:Designing computational schemes dedicated to fission nuclear reactors for space[C]//International Conference on Nuclear and Emerging Technologies for Space.Albuquerque,N M:[s.n.],2013.
[21] HEBERT A,SEKKI D,CHAMBON R.A user’s guide for DONJON,Version5,IGE-344[R].Ecole Polytechnique de Montreal:Institut de Genie Nucleaire,2017.
[22] HéBERT A.TRIVAC:A modular diffusion code for fuel management and design applications[J].Nuclear Journal of Canada,1987,1:325-331.
[23] RICHARD S.Assembly homogenization techniques for core calculations[J].Progress in Nuclear Energy,2009,51(1):14-31.
[24] REITSMA F,STRYDOM G,de HAAS J B M,et al.The PBMR steady-state and coupled kinetics core thermal-hydraulics benchmark test problems[J].Nuclear Engineering and Design,2006,236(5-6):657-668.
[25] HAN J S.Sensitivity study on the energy group structure for high temperature reactor analysis[M].USA:The Pennsylvania State University,2008.
[26] ZOU C Y,CAI X Z,JIANG D Z,et al.Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor[J].Nuclear Engineering and Design,2015,281:114-120.
[27] CUI D Y,LI X X,XIA S P,et al.Possible scenarios for the transition to thorium fuel cycle in molten salt reactor by using enriched uranium[J].Progress in Nuclear Energy,2018,104:75-84.