乏燃料废包壳残留物质核素分析计算
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摘要
为分析计算乏燃料废包壳残留物质的核素含量,以M310型核电机组及燃料组件为分析对象,建立了乏燃料废包壳残留物质核素含量分层计算模型,用SCALE程序计算分析了~(244)Cm含量、总Pu含量及~(244)Cm/Pu比等主要参数随燃耗及冷却时间的变化。计算结果表明,~(244)Cm含量、总Pu含量及~(244)Cm/Pu比随燃耗及冷却时间的变化均可用三阶多项式拟合。本文工作为废包壳残留物质非破坏性测量方法研究提供了数据支持。
In order to analyze the nuclide content of spent fuel waste cladding residue, a multilayer calculation model was established for M310 nuclear power plant and fuel assembly. The variation of ~(244)Cm content, total Pu content and ~(244)Cm/Pu ratio with burnup and cooling time was calculated by SCALE code. The results show that the variation of ~(244)Cm content, total Pu content and ~(244)Cm/Pu ratio with burnup and cooling time can be fitted by third-order polynomial. This study provides data support for non-destructive analysis method of waste cladding residue.
In order to analyze the nuclide content of spent fuel waste cladding residue, a multilayer calculation model was established for M310 nuclear power plant and fuel assembly. The variation of ~(244)Cm content, total Pu content and ~(244)Cm/Pu ratio with burnup and cooling time was calculated by SCALE code. The results show that the variation of ~(244)Cm content, total Pu content and ~(244)Cm/Pu ratio with burnup and cooling time can be fitted by third-order polynomial. This study provides data support for non-destructive analysis method of waste cladding residue.
引文
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[2] 乔盛忠,朱荣保,郏惠忠,等. 模拟废包壳中残留铀含量的有源中子测定[J]. 原子能科学技术,1994,28(1):8-15. QIAO Shengzhong, ZHU Rongbao, JIA Huizhong, et al. Determination of the residual uranium content in simulant leached hulls[J]. Atomic Energy Science and Technology, 1994, 28(1): 8-15(in Chinese).
[3] 高彦锋,张孝泽,王瑞宏. 用MCNP程序计算核燃料废包壳缓发裂变中子形成的热中子通量密度[J]. 原子能科学技术,1993,27(3):198-205. GAO Yanfeng, ZHANG Xiaoze, WANG Rui-hong. Computing with MCNP the thermal flux density due to the delayed fission neutron generated in processing the clads of burned-up fuel roads[J]. Atomic Energy Science and Technology, 1993, 27(3): 198-205(in Chinese).
[4] 任凤仪. 国外核燃料后处理[M]. 北京:原子能出版社,2006.
[5] JOEL R, DEOKJUNG L, KORD S. CASMO-5/CASMO-5M: A fuel assembly burnup program[R]. US: Studsvik Scandpower, 2011.
[6] 于超,朱庆福. 基于图论算法的多群点燃耗程序研制[R]. 北京:中国原子能科学研究院,2012.
[7] 于超,朱庆福. 蒙特卡罗燃耗计算程序 MCNTRANS的开发与验证[J]. 原子能科学技术,2013,47(10):1 824-1 828. YU Chao, ZHU Qingfu. Development and validation of Monte-Carlo burnup calculation code MCNTRANS[J]. Atomic Energy Science and Technology, 2013, 47(10): 1 824-1 828(in Chinese).
[8] SANDERS C E, GAULD I C. Isotopic analysis of high-burnup PWR spent fuel samples from the Takahama-3 reactor, NUREG/CR-ORNL/TM-2001/259[R]. US: ORNL, 2002.
[9] YOSHINORI N, KENYA S, TAKENORI S. Technical development on burn-up credit for spent LWR fuels[R]. US: ORNL, 2002.
[2] 乔盛忠,朱荣保,郏惠忠,等. 模拟废包壳中残留铀含量的有源中子测定[J]. 原子能科学技术,1994,28(1):8-15. QIAO Shengzhong, ZHU Rongbao, JIA Huizhong, et al. Determination of the residual uranium content in simulant leached hulls[J]. Atomic Energy Science and Technology, 1994, 28(1): 8-15(in Chinese).
[3] 高彦锋,张孝泽,王瑞宏. 用MCNP程序计算核燃料废包壳缓发裂变中子形成的热中子通量密度[J]. 原子能科学技术,1993,27(3):198-205. GAO Yanfeng, ZHANG Xiaoze, WANG Rui-hong. Computing with MCNP the thermal flux density due to the delayed fission neutron generated in processing the clads of burned-up fuel roads[J]. Atomic Energy Science and Technology, 1993, 27(3): 198-205(in Chinese).
[4] 任凤仪. 国外核燃料后处理[M]. 北京:原子能出版社,2006.
[5] JOEL R, DEOKJUNG L, KORD S. CASMO-5/CASMO-5M: A fuel assembly burnup program[R]. US: Studsvik Scandpower, 2011.
[6] 于超,朱庆福. 基于图论算法的多群点燃耗程序研制[R]. 北京:中国原子能科学研究院,2012.
[7] 于超,朱庆福. 蒙特卡罗燃耗计算程序 MCNTRANS的开发与验证[J]. 原子能科学技术,2013,47(10):1 824-1 828. YU Chao, ZHU Qingfu. Development and validation of Monte-Carlo burnup calculation code MCNTRANS[J]. Atomic Energy Science and Technology, 2013, 47(10): 1 824-1 828(in Chinese).
[8] SANDERS C E, GAULD I C. Isotopic analysis of high-burnup PWR spent fuel samples from the Takahama-3 reactor, NUREG/CR-ORNL/TM-2001/259[R]. US: ORNL, 2002.
[9] YOSHINORI N, KENYA S, TAKENORI S. Technical development on burn-up credit for spent LWR fuels[R]. US: ORNL, 2002.