MCNP在中子输运问题计算中的影响因素分析
|
摘要
随着MCNP在核科学和工程方面日益广泛的应用，MCNP计算中的影响因素分析近来受到越来越多的关注，例如在计算反应堆的物理状态发生微小变化所引起的反应性变化及反应堆压力容器快中子通量的校核计算方面。正是基于此目的，本文分析了MCNP计算结果的影响因素。
通过了解程序中有关中子作用过程的物理模型和MCNP的使用方法，建立了两个典型的基准题的计算模型：MOX燃料试验研究堆的临界计算和反应堆压力容器通量计算，提出建模过程中一些技巧，进行了计算结果的影响因素分析：
在MOX燃料试验研究堆的临界计算中，计算了实验堆的有效增殖系数和相对功率分布。通过更改初始裂变源(SRCTP)的方式，证明以不同的随机数序列重复计算问题，得到的计算结果总存在一定的涨落，并随着总迭代次数的增加逐渐减小；紧接着在统计涨落足够小的基础上，讨论了数据库的差异对计算结果的影响；指出采用S(α，β)热散射处理是得到合适的有效增殖系数计算值的必须条件。
在压水堆压力容器通量计算中，进一步证实了计算结果的统计涨落和数据库对计算结果的影响，讨论了不同方差减小技巧下的计算结果间存在差异的原因，并分析了程序版本不同对计算结果的影响。最后通过和基准题中压力容器内表面快中子通量的相对分布比较，表明了计算过程的正确性。
With MCNP applied increasingly and broadly in the aspects of the nuclear science and engineering, analyses of the influencing factors in the computed results using MCNP have recently received more attention from areas such as computations of reactivity changes caused by the minor changes of the reactor state and benchmark computations of the neutron flux of the reactor pressure vessel. To serve these needs, analyses of the influencing factors in the computed results using MCNP are present in the thesis.
After comprehending the physics of neutron interaction being the essence of MCNP code and the method about how to use MCNP, this paper sets up computation model basing on two typical benchmark problems: the criticality computation of the MOX fuel research reactor and the flux computation of the reactor pressure vessel. During the process, there are the operative skills advanced in the paper. Finally, the most important work that analyzes the influencing factors in the computed results has been finished.
The criticality computation of the MOX fuel research reactors firstly calculates the effective multiplication factors and the relative power distributions. And secondly it proves that the computed results rerunning the problems with a different random number sequence always have certain statistical fluctuation that reduces gradually along with the number of history increasing by the way of changing the initial fission source (SRCTP) from a previous MCNP criticality calculation. As following, under the condition of the statistical fluctuation weakly enough, the difference between the computed results using different data libraries is discussed, it pointed out that using S(α,β) thermal scatting treatment is the essential condition to obtain the appropriate effective multiplication factors.
In the neutron flux computation of the reactor pressure vessel, the statistical fluctuation existing in computed results and the influence of the data libraries being on the computed result are further confirmed. There is the discussion about the reasons of difference between the computed results using different variance reduction techniques and the influence of procedure edition on the computed results. Finally, it indicates that the computation process is correctly according to the fast flux relative distribution of the pressure vessel internal surface presenting in the benchmark problem.
通过了解程序中有关中子作用过程的物理模型和MCNP的使用方法，建立了两个典型的基准题的计算模型：MOX燃料试验研究堆的临界计算和反应堆压力容器通量计算，提出建模过程中一些技巧，进行了计算结果的影响因素分析：
在MOX燃料试验研究堆的临界计算中，计算了实验堆的有效增殖系数和相对功率分布。通过更改初始裂变源(SRCTP)的方式，证明以不同的随机数序列重复计算问题，得到的计算结果总存在一定的涨落，并随着总迭代次数的增加逐渐减小；紧接着在统计涨落足够小的基础上，讨论了数据库的差异对计算结果的影响；指出采用S(α，β)热散射处理是得到合适的有效增殖系数计算值的必须条件。
在压水堆压力容器通量计算中，进一步证实了计算结果的统计涨落和数据库对计算结果的影响，讨论了不同方差减小技巧下的计算结果间存在差异的原因，并分析了程序版本不同对计算结果的影响。最后通过和基准题中压力容器内表面快中子通量的相对分布比较，表明了计算过程的正确性。
With MCNP applied increasingly and broadly in the aspects of the nuclear science and engineering, analyses of the influencing factors in the computed results using MCNP have recently received more attention from areas such as computations of reactivity changes caused by the minor changes of the reactor state and benchmark computations of the neutron flux of the reactor pressure vessel. To serve these needs, analyses of the influencing factors in the computed results using MCNP are present in the thesis.
After comprehending the physics of neutron interaction being the essence of MCNP code and the method about how to use MCNP, this paper sets up computation model basing on two typical benchmark problems: the criticality computation of the MOX fuel research reactor and the flux computation of the reactor pressure vessel. During the process, there are the operative skills advanced in the paper. Finally, the most important work that analyzes the influencing factors in the computed results has been finished.
The criticality computation of the MOX fuel research reactors firstly calculates the effective multiplication factors and the relative power distributions. And secondly it proves that the computed results rerunning the problems with a different random number sequence always have certain statistical fluctuation that reduces gradually along with the number of history increasing by the way of changing the initial fission source (SRCTP) from a previous MCNP criticality calculation. As following, under the condition of the statistical fluctuation weakly enough, the difference between the computed results using different data libraries is discussed, it pointed out that using S(α,β) thermal scatting treatment is the essential condition to obtain the appropriate effective multiplication factors.
In the neutron flux computation of the reactor pressure vessel, the statistical fluctuation existing in computed results and the influence of the data libraries being on the computed result are further confirmed. There is the discussion about the reasons of difference between the computed results using different variance reduction techniques and the influence of procedure edition on the computed results. Finally, it indicates that the computation process is correctly according to the fast flux relative distribution of the pressure vessel internal surface presenting in the benchmark problem.
引文
[1] 谢仲生，邓力著．中子输运理论数值计算方法．西安：西北工业大学出版社，2005：5页 401-418页
[2] 谢仲生，张育曼，张建民等．核反应堆物理数值计算．北京：原子能出版社，1996：167-172页
[3] 裴鹿成，张孝泽．蒙特卡洛方法及其在粒子输运理论中等应用．北京：科学出版社，1980
[4] Christopher J. Werner. Simulation of delayed neutrons using MCNP Progress in Nuclear Energy, 2002,41(4):385-389P
[5] R. Arthur Forster, Lawrence J.Cox, Richard F. Barrett, ed. MCNP Version 5. Nuclear Instruments and Methods in Physics Research B, 2004,213:82 -86P
[6] Michael R. James, G.W. McKinney, John S. Hendricks, ed. Recent enhancements in MCNPX: Heavy-ion transport and the LAQGSM physics model. Nuclear Instruments and Methods in Physics Research A, 2006,562:819- 822P
[7] Dennis M. Duggan. Improved radial dose function estimation using current version MCNP Monte-Carlo simulation: Model 6711 and ISC3500 125I brachytherapy sources. Applied Radiation and Isotopes, 2004, 61:1443-1450P
[8] 钟兆鹏，施工，胡永明．MCNP程序在反应堆临界计算中的应用．核动力工程，2003，24(1)：8-11页
[9] 经荥清，杨永伟，许云林．蒙特卡罗方法用于HTR-10首次临界燃料装料预估的校算．核动力工程，2005，26(1)：28-34页
[10] 陈伟，张颖，陈立新等．铀氢锆研究堆堆芯脉冲参数计算．第九届反应堆数值计算和粒子输运暨2002年反应堆物理会议．北京，2002：268-272页
[11] 景春元，朱继洲．MCNP／4B在脉冲堆物理计算中的应用．第八届全国核反应堆数值计算和粒子输运学术会议．深圳，2000
[12] 邱小平，黎学川，王建华．次临界堆芯参数变化对K_(eff)值的影响．南华大学学报(自然科学版)，2005，19(1)：42-46页
[13] 王晓宇，冯开明，黄锦华．启明星实验装置设计方案中K_(eff)与最优化参数的计算．核聚变与等离子体物理，2004，24(2)：123-128页
[14] M.Y. Khalila, M.K. Shaatb, A.Y. Abdelfattahb. Neutronic conceptual design of the ETRR-2 cold-neutron source using the MCNP code. Physica B, 2005,358:36-41P
[15] I. Khamis, I. Sulieman. Monte Carlo simulation of a conceptual thermal column in the Syrian miniature neutron source reactor using MCNP-4C. Annals of Nuclear Energy, 2006,33:622-626P
[16] 邱立青，贾斗南，邓才玉等．蒙特卡罗方法用于HFETR堆芯γ释热的可行性研究．核动力工程，2006，27(4)：13-15页
[17] Wang Kan, T P Lou, E Greenspan, ed. Benchmarking and Validation of MOCUP. Proceedings of The 2000 ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the next Millennium, Pittsburgh, PA, USA, 7-11 May, 2000:15202P
[18] 余纲林，王侃，王煜宏．MCBurn-MCNP和ORIGEN耦合程序系统．原子能科学技术，2003，37(3)：250-254页
[19] 薛小刚，周培德，柯国土等．基于蒙特卡罗方法的三维燃耗计算研究．核科学与工程，2005，25(1)：24-29页
[20] 蒋校丰，谢仲生．蒙卡-燃耗程序系统及ADS基准题的计算．核科学与工程，2003，23(4)：325-331页
[21] D. Ridikas, R. Plukiene, A. Plukis, ed. Fusion-fission hybrid system for nuclear waste transmutation (1): Characterization of the system and burn-up calculations. Progress in Nuclear Energy, 2006,48:235-246P
[22] R. Plukiene, A. Plukis, D. Ridikas, ed. Fusion-fission hybrid system for nuclear waste transmutation (Ⅰ): From the burn-up optimization to the testsof different data libraries. Progress in Nuclear Energy, 2006,48:247-258P
[23] 杨永伟，李浩泉，经荥清．用LAHET和MCNP程序研究散裂中子靶的中子学性能．原子能科学技术，2005，39(3)：205-208页
[24] Y. Chen, U. Fischer. Rigorous MCNP based shutdown dose rate calculations: computational scheme, verification calculations and application to ITER. Fusion Engineering and Design, 2002, (63-64):107-114P
[25] M. Sohrabpoura, M. Hassanzadehb, M. Shahriarib etc. Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry, Applied Radiation and Isotopes, 2002,57:537-542P
[26] Regulatory Guide 1.190. Calculational and dosimetry methods for determining pressure vessel neutron fluence. U.S:Nuclear Regulatory Commission, 2001,3
[27] J. F. Carew, K. Hu, i. Aronson, ed. PWR and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions. Washington. U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research, 2001
[28] Yoshihiro Ogawaa, Itsumasa Urabeb, Hiroyuki Sagawab, ed. Measurement and analysis of leakage neutron energy spectra around the Kinki University Reactor UTR-KINKI. Nuclear Instruments and Methods in Physics Research A, 2002,476:435-439P
[29] 孔巢城，李泉凤，朱立等．电子加速器屏蔽室空间散射的蒙特卡罗计算．高能物理与核物理，2004，28(4)：431-436页
[30] E. Mainardi, F. Premuda, E. Lee. Comparison and physical interpretation of MCNP and TART neutron and γMonte Carlo shielding calculations for a heavy-ion ICF system. Nuclear Instruments and Methods in Physics Research B, 2004,213:182-185P
[31] 杨玉中．基于蒙特卡罗输运技术的压力容器及辐照监督管中子注量计算方法研究及应用．第十届全国核反应堆数值计算和粒子输运学术会议暨 2004年度反应堆物理会议．秦山，2004：535-543页
[32] 邱忠明，何忠良，孔德萍．MCNP程序在秦山核电厂辐照监督管和RPV中子注量计算中的应用．第十一届全国核反应堆数值计算和粒子输运学术会议暨2006年度反应堆物理会议．哈尔滨，2006：655-661页
[33] 王义，程建平，范佳锦等．MRPC γ灵敏度计算．核电子学与探测技术，2003，23(1)：33-35页
[34] 张国光，黄龙生，张前美等．SI—PIN探测器中子直照灵敏度研究．核电子学与探测技术，2004，24(4)：372-375页
[35] 谈春明．利用MCNP程序对充气电离室探测性能的研究．核电子学与探测技术，2003，23(3)：236-239页
[36] 郭春营，罗永锋，林源根．NaI(T1)闪烁谱仪峰总比的蒙特卡罗计算．核电子学与探测技术，2002，22(5)：453-455页
[37] 李亢，钟友琴，黄锦延．MC方法在反应堆初级中子源计算中的应用尝试．第十一届反应堆数值计算和粒子输运暨2006年反应堆物理会议．哈尔滨，2006：69-71页
[38] F.H.H. Al-Ghorabie, S.S.A. Natto, S.H.A. Al-Lyhiani. A comparison between EGS4 and MCNP computer modeling of an in vivo X-ray fluorescence system. Computers in Biology and Medicine, 2001, 31:73-83P
[39] 王进亮，白净，罗建文等．蒙特卡罗方法和三维数字人体模型在放疗计划质量保证中的应用．生物物理学报，2005，21(3)：221-226页
[40] 邓力，谢仲生，李树．反应堆临界-燃耗耦合蒙特卡罗计算．原子能科学技术．2002，36(2)：168-174页
[41] M. Marseguerra, E. Padovani, S. A. Pozzi. Use of the MCNP-POLIMI code for time correlation safeguards measurement. Progress in Nuclear Energy, 2003,43(1-4):305-311P
[42] Enrico Padovani, Paolo Peerani, Mirko Da Ros, ed. Simulation of correlated counts from an Am-Li source. Nuclear Instruments and Methods in Physics Research A, 2006,557:599-606P
[43] W. K. Matthes. Monte Carlo simulation of gas-flow using MCNP. Annals of Nuclear Energy, 2005, 32:1495-1508P
[44] Judith F Briesmesister. MCNP-A general monte carlo n-particle transport code. Los-Alamos National Laboratory Report LA-12625-M, Version4B. 1997
[45] Mafizur Rahan, Takenori Suzaki, Takamasa Mori. Analytical study of two-region TCA critical experiment with PWR-type MOX fuel by using Monte Carlo code MVP. Japan: Japan Atomic Energy Research Institute, 2003, 007:1-68P
[46] John G.Williams, David W. Vehar, Frank H. Ruddy, ed. Impact of the ENDF/B-VI Cross Sections on the RPV Fluence Determination. American Society for Testing and Materials, West Conshohoken, PA 2000
[2] 谢仲生，张育曼，张建民等．核反应堆物理数值计算．北京：原子能出版社，1996：167-172页
[3] 裴鹿成，张孝泽．蒙特卡洛方法及其在粒子输运理论中等应用．北京：科学出版社，1980
[4] Christopher J. Werner. Simulation of delayed neutrons using MCNP Progress in Nuclear Energy, 2002,41(4):385-389P
[5] R. Arthur Forster, Lawrence J.Cox, Richard F. Barrett, ed. MCNP Version 5. Nuclear Instruments and Methods in Physics Research B, 2004,213:82 -86P
[6] Michael R. James, G.W. McKinney, John S. Hendricks, ed. Recent enhancements in MCNPX: Heavy-ion transport and the LAQGSM physics model. Nuclear Instruments and Methods in Physics Research A, 2006,562:819- 822P
[7] Dennis M. Duggan. Improved radial dose function estimation using current version MCNP Monte-Carlo simulation: Model 6711 and ISC3500 125I brachytherapy sources. Applied Radiation and Isotopes, 2004, 61:1443-1450P
[8] 钟兆鹏，施工，胡永明．MCNP程序在反应堆临界计算中的应用．核动力工程，2003，24(1)：8-11页
[9] 经荥清，杨永伟，许云林．蒙特卡罗方法用于HTR-10首次临界燃料装料预估的校算．核动力工程，2005，26(1)：28-34页
[10] 陈伟，张颖，陈立新等．铀氢锆研究堆堆芯脉冲参数计算．第九届反应堆数值计算和粒子输运暨2002年反应堆物理会议．北京，2002：268-272页
[11] 景春元，朱继洲．MCNP／4B在脉冲堆物理计算中的应用．第八届全国核反应堆数值计算和粒子输运学术会议．深圳，2000
[12] 邱小平，黎学川，王建华．次临界堆芯参数变化对K_(eff)值的影响．南华大学学报(自然科学版)，2005，19(1)：42-46页
[13] 王晓宇，冯开明，黄锦华．启明星实验装置设计方案中K_(eff)与最优化参数的计算．核聚变与等离子体物理，2004，24(2)：123-128页
[14] M.Y. Khalila, M.K. Shaatb, A.Y. Abdelfattahb. Neutronic conceptual design of the ETRR-2 cold-neutron source using the MCNP code. Physica B, 2005,358:36-41P
[15] I. Khamis, I. Sulieman. Monte Carlo simulation of a conceptual thermal column in the Syrian miniature neutron source reactor using MCNP-4C. Annals of Nuclear Energy, 2006,33:622-626P
[16] 邱立青，贾斗南，邓才玉等．蒙特卡罗方法用于HFETR堆芯γ释热的可行性研究．核动力工程，2006，27(4)：13-15页
[17] Wang Kan, T P Lou, E Greenspan, ed. Benchmarking and Validation of MOCUP. Proceedings of The 2000 ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the next Millennium, Pittsburgh, PA, USA, 7-11 May, 2000:15202P
[18] 余纲林，王侃，王煜宏．MCBurn-MCNP和ORIGEN耦合程序系统．原子能科学技术，2003，37(3)：250-254页
[19] 薛小刚，周培德，柯国土等．基于蒙特卡罗方法的三维燃耗计算研究．核科学与工程，2005，25(1)：24-29页
[20] 蒋校丰，谢仲生．蒙卡-燃耗程序系统及ADS基准题的计算．核科学与工程，2003，23(4)：325-331页
[21] D. Ridikas, R. Plukiene, A. Plukis, ed. Fusion-fission hybrid system for nuclear waste transmutation (1): Characterization of the system and burn-up calculations. Progress in Nuclear Energy, 2006,48:235-246P
[22] R. Plukiene, A. Plukis, D. Ridikas, ed. Fusion-fission hybrid system for nuclear waste transmutation (Ⅰ): From the burn-up optimization to the testsof different data libraries. Progress in Nuclear Energy, 2006,48:247-258P
[23] 杨永伟，李浩泉，经荥清．用LAHET和MCNP程序研究散裂中子靶的中子学性能．原子能科学技术，2005，39(3)：205-208页
[24] Y. Chen, U. Fischer. Rigorous MCNP based shutdown dose rate calculations: computational scheme, verification calculations and application to ITER. Fusion Engineering and Design, 2002, (63-64):107-114P
[25] M. Sohrabpoura, M. Hassanzadehb, M. Shahriarib etc. Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry, Applied Radiation and Isotopes, 2002,57:537-542P
[26] Regulatory Guide 1.190. Calculational and dosimetry methods for determining pressure vessel neutron fluence. U.S:Nuclear Regulatory Commission, 2001,3
[27] J. F. Carew, K. Hu, i. Aronson, ed. PWR and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions. Washington. U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research, 2001
[28] Yoshihiro Ogawaa, Itsumasa Urabeb, Hiroyuki Sagawab, ed. Measurement and analysis of leakage neutron energy spectra around the Kinki University Reactor UTR-KINKI. Nuclear Instruments and Methods in Physics Research A, 2002,476:435-439P
[29] 孔巢城，李泉凤，朱立等．电子加速器屏蔽室空间散射的蒙特卡罗计算．高能物理与核物理，2004，28(4)：431-436页
[30] E. Mainardi, F. Premuda, E. Lee. Comparison and physical interpretation of MCNP and TART neutron and γMonte Carlo shielding calculations for a heavy-ion ICF system. Nuclear Instruments and Methods in Physics Research B, 2004,213:182-185P
[31] 杨玉中．基于蒙特卡罗输运技术的压力容器及辐照监督管中子注量计算方法研究及应用．第十届全国核反应堆数值计算和粒子输运学术会议暨 2004年度反应堆物理会议．秦山，2004：535-543页
[32] 邱忠明，何忠良，孔德萍．MCNP程序在秦山核电厂辐照监督管和RPV中子注量计算中的应用．第十一届全国核反应堆数值计算和粒子输运学术会议暨2006年度反应堆物理会议．哈尔滨，2006：655-661页
[33] 王义，程建平，范佳锦等．MRPC γ灵敏度计算．核电子学与探测技术，2003，23(1)：33-35页
[34] 张国光，黄龙生，张前美等．SI—PIN探测器中子直照灵敏度研究．核电子学与探测技术，2004，24(4)：372-375页
[35] 谈春明．利用MCNP程序对充气电离室探测性能的研究．核电子学与探测技术，2003，23(3)：236-239页
[36] 郭春营，罗永锋，林源根．NaI(T1)闪烁谱仪峰总比的蒙特卡罗计算．核电子学与探测技术，2002，22(5)：453-455页
[37] 李亢，钟友琴，黄锦延．MC方法在反应堆初级中子源计算中的应用尝试．第十一届反应堆数值计算和粒子输运暨2006年反应堆物理会议．哈尔滨，2006：69-71页
[38] F.H.H. Al-Ghorabie, S.S.A. Natto, S.H.A. Al-Lyhiani. A comparison between EGS4 and MCNP computer modeling of an in vivo X-ray fluorescence system. Computers in Biology and Medicine, 2001, 31:73-83P
[39] 王进亮，白净，罗建文等．蒙特卡罗方法和三维数字人体模型在放疗计划质量保证中的应用．生物物理学报，2005，21(3)：221-226页
[40] 邓力，谢仲生，李树．反应堆临界-燃耗耦合蒙特卡罗计算．原子能科学技术．2002，36(2)：168-174页
[41] M. Marseguerra, E. Padovani, S. A. Pozzi. Use of the MCNP-POLIMI code for time correlation safeguards measurement. Progress in Nuclear Energy, 2003,43(1-4):305-311P
[42] Enrico Padovani, Paolo Peerani, Mirko Da Ros, ed. Simulation of correlated counts from an Am-Li source. Nuclear Instruments and Methods in Physics Research A, 2006,557:599-606P
[43] W. K. Matthes. Monte Carlo simulation of gas-flow using MCNP. Annals of Nuclear Energy, 2005, 32:1495-1508P
[44] Judith F Briesmesister. MCNP-A general monte carlo n-particle transport code. Los-Alamos National Laboratory Report LA-12625-M, Version4B. 1997
[45] Mafizur Rahan, Takenori Suzaki, Takamasa Mori. Analytical study of two-region TCA critical experiment with PWR-type MOX fuel by using Monte Carlo code MVP. Japan: Japan Atomic Energy Research Institute, 2003, 007:1-68P
[46] John G.Williams, David W. Vehar, Frank H. Ruddy, ed. Impact of the ENDF/B-VI Cross Sections on the RPV Fluence Determination. American Society for Testing and Materials, West Conshohoken, PA 2000