MCNP温度相关中子截面库的研制及基准验证

Development and Benchmark Validation of TemperatureDependent Neutron CrossSection Library for MCNP

摘要: 本文在使用NJOY软件由ENDF格式的中子截面文件处理生成ACE (a compact ENDF) 格式的温度相关中子截面库的方法研究的基础上，开展温度相关中子截面库的研制及验证。研制过程中，选择了在反应堆设计和运行温度范围内的16个温度点。在温度相关中子截面库的验证过程中应用了4个基准题：带可燃毒物的轻水堆芯临界基准题、反应性多普勒系数基准题、标准CANDU组件燃料温度系数基准题和VHTRC温度系数基准题。验证计算结果表明，该温度相关中子截面库可运用于反应堆物理的计算分析中。

Abstract: A compact ENDF (ACE) data library of 321 nuclides at 16 temperature points was generated using NJOY software. A program used for generating a single nuclide cross section file was developed. The final validation of the resulted library and program was done using 4 temperature dependent benchmarks: a LWR core with burnable absorbers, fuel cell Doppler coefficient benchmark, VHTRC temperature coefficient benchmark and standard CANDU fuel bundle benchmark. In conclusion, the resulted ACE library and program in the work are available and correct.