基于SCALE／TRITON的单流双区熔盐堆燃耗计算方法
美国橡树岭国家实验室开发的SCALE／TRITON程序广泛用于反应堆临界安全、中子物理、辐射屏蔽和灵敏度与不确定度等方面的计算分析。基于SCALE／TRITON程序，采用等效体积、均匀混合和平均截面等三种外部耦合方法，处理单流双区熔盐堆的燃耗计算，解决了SCALE／TRITON程序在计算中不能精确反映流动燃料周期性均匀混合的问题。研究表明平均截面法与均匀混合法的计算结果几乎完全一致，与橡树岭文献结果也能很好符合，等效体积法因为没有考虑堆芯分区结构的差异而导致计算结果与其他两种方法偏离较大。基于SCALE／TRITON发展的平均截面法，放宽了对步长的要求，具有准确性好、计算效率高的优点，适用于熔...
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| Published in: | 核技术 Vol. 40; no. 8; pp. 72 - 78 |
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| Format: | Journal Article |
| Language: | Chinese |
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中国科学院上海应用物理研究所 嘉定园区 上海 201800
2017
中国科学院先进核能创新研究院 上海 201800 中国科学院大学 北京 100049%中国科学院上海应用物理研究所 嘉定园区 上海 201800 |
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| Abstract | 美国橡树岭国家实验室开发的SCALE／TRITON程序广泛用于反应堆临界安全、中子物理、辐射屏蔽和灵敏度与不确定度等方面的计算分析。基于SCALE／TRITON程序，采用等效体积、均匀混合和平均截面等三种外部耦合方法，处理单流双区熔盐堆的燃耗计算，解决了SCALE／TRITON程序在计算中不能精确反映流动燃料周期性均匀混合的问题。研究表明平均截面法与均匀混合法的计算结果几乎完全一致，与橡树岭文献结果也能很好符合，等效体积法因为没有考虑堆芯分区结构的差异而导致计算结果与其他两种方法偏离较大。基于SCALE／TRITON发展的平均截面法，放宽了对步长的要求，具有准确性好、计算效率高的优点，适用于熔盐堆两区（或多区）的堆芯设计与燃耗性能分析，具有重要的应用意义。 |
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| AbstractList | 美国橡树岭国家实验室开发的SCALE／TRITON程序广泛用于反应堆临界安全、中子物理、辐射屏蔽和灵敏度与不确定度等方面的计算分析。基于SCALE／TRITON程序，采用等效体积、均匀混合和平均截面等三种外部耦合方法，处理单流双区熔盐堆的燃耗计算，解决了SCALE／TRITON程序在计算中不能精确反映流动燃料周期性均匀混合的问题。研究表明平均截面法与均匀混合法的计算结果几乎完全一致，与橡树岭文献结果也能很好符合，等效体积法因为没有考虑堆芯分区结构的差异而导致计算结果与其他两种方法偏离较大。基于SCALE／TRITON发展的平均截面法，放宽了对步长的要求，具有准确性好、计算效率高的优点，适用于熔盐堆两区（或多区）的堆芯设计与燃耗性能分析，具有重要的应用意义。 TL99; 美国橡树岭国家实验室开发的SCALE/TRITON程序广泛用于反应堆临界安全、中子物理、辐射屏蔽和灵敏度与不确定度等方面的计算分析.基于SCALE/TRITON程序,采用等效体积、均匀混合和平均截面等三种外部耦合方法,处理单流双区熔盐堆的燃耗计算,解决了SCALE/TRITON程序在计算中不能精确反映流动燃料周期性均匀混合的问题.研究表明平均截面法与均匀混合法的计算结果几乎完全一致,与橡树岭文献结果也能很好符合,等效体积法因为没有考虑堆芯分区结构的差异而导致计算结果与其他两种方法偏离较大.基于SCALE/TRITON发展的平均截面法,放宽了对步长的要求,具有准确性好、计算效率高的优点,适用于熔盐堆两区(或多区)的堆芯设计与燃耗性能分析,具有重要的应用意义. |
| Abstract_FL | Background: The standardized computer analysis for licensing evaluation (SCALE) developed in the Oak Ridge National Laboratory (ORNL) of USA has been widely used in criticality safety, neutron physics, radiation shielding, and sensitivity and uncertainty analysis. However, the burnup calculation for single-fluid, two-zone molten salt reactor (MSR) has not been well dealt with in SCALE/TRITON due to the cell information card (Celldata) which is used in unit cell calculations to generate problem-dependent multigroup cross sections. Purpose: This study aims to develop and evaluate possible solutions to the problem above. Methods: Based on external program, three methods (i.e., homogeneous mixing method, equivalent volume method and average cross section method), are developed without any modification of the existing codes in SCALE6 and they are tested in a MSR with two-zone core. Test results are compared and analyzed. Results: Comparison of the three methods indicates that the results obtained by average cross section method are almost equal to those obtained by homogeneous mixing method and moreover they accord well with the results given in ORNL's work, whilst the equivalent volume method is not sufficient to describe the difference of unit cells in the core. Conclusion: The average cross section method with a relatively high computational efficiency and accuracy is recommended for burnup calculation in the MSR with two or more zones when using SCALE/TRITON. |
| Author | 崔德阳 夏少鹏 余呈刚 蔡翔舟 陈金根 |
| AuthorAffiliation | 中国科学院上海应用物理研究所嘉定园区,上海201800 中国科学院先进核能创新研究院,上海201800 中国科学院大学,北京100049 |
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| Author_FL | YU Chenggang XIA Shaopeng CAI Xiangzhou CHEN Jingen CUI Deyang |
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| Keywords | 熔盐堆 Average cross section Enriched uranium 燃耗 Burnup 富集铀 MSR 平均截面 |
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| Notes | Background： The standardized computer analysis for licensing evaluation （SCALE） developed in the Oak Ridge National Laboratory （ORNL） of USA has been widely used in criticality safety, neutron physics, radiation shielding, and sensitivity and uncertainty analysis. However, the burnup calculation for single-fluid, two-zone molten salt reactor （MSR） has not been well dealt with in SCALE/TRITON due to the cell information card （Celldata） which is used in unit cell calculations to generate problem-dependent multigroup cross sections. Purpose： This study aims to develop and evaluate possible solutions to the problem above. Methods： Based on external program, three methods （i.e., homogeneous mixing method, equivalent volume method and average cross section method）, are developed without any modification of the existing codes in SCALE6 and they are tested in a MSR with two-zone core. Test results are compared and analyzed. Results： Comparison of the three methods indicates that the results obtained by average cross |
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| PublicationYear | 2017 |
| Publisher | 中国科学院上海应用物理研究所 嘉定园区 上海 201800 中国科学院先进核能创新研究院 上海 201800 中国科学院大学 北京 100049%中国科学院上海应用物理研究所 嘉定园区 上海 201800 |
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| SubjectTerms | 富集铀 平均截面 熔盐堆 燃耗 |
| Title | 基于SCALE／TRITON的单流双区熔盐堆燃耗计算方法 |
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