摘要
用蒙特卡罗方法及可适应多种截面的三维多群P3 中子输运蒙特卡罗程序MCMG ,计算比较了不同裂变材料使用向量裂变谱与矩阵裂变谱的计算结果 ,通过与MCNP程序结果的比较 ,确信两种裂变谱对计算结果的影响是存在的 ,但对计算结果不会带来本质影响 .
When the fission reaction happens in neutron transport calculation,the energy of the fission neutron can be usually determined by the fission spectrum.Since the coefficient of the spectrum type depends on the incident neutron energy,the multigroup neutron fission spectrum should strictly be a matrix form.In general transport calculation,the 235 U vector fission spectrum is usually chosen as the standard fission spectrum.In order to make clear what kind of effect will be produced after this treatment,the Monte Carlo method and the multigroup P 3 Monte Carlo neutron transport code MCMG are used to analyze the difference between two types of fission spectrums.By comparison the results with MCNP code,it is certain that the difference exists,but it does not have any effect on the correction of the calculated results.At the same time,the calculation results of the different neutron cross section libraries have been compared. [
出处
《计算物理》
CSCD
北大核心
2002年第1期67-72,共6页
Chinese Journal of Computational Physics
基金
中国工程物理研究院 ( 2 0 0 10 6 6 0 )
关键词
矩阵裂变谱
向量裂变谱
临界计算
蒙特卡罗法
中子
matrix fission spectrum
vector fission spectrum
critical calculation
Monte Carlo
