快堆用蒙特卡罗多群数据库制作方法研究

基于离散角方法，开发了蒙特卡罗多群数据库生成程序MGXSMC，该程序可以实现从输入文件读取截面数据或者从指定格式的截面库中读取截面，产生可供蒙特卡罗程序MCNP或RMC计算的数据库，并且可自动生成相应的索引文件列表。采用二维两群不带反射层的国际原子能机构（IAEA）压水堆（PWR）基准题和铅基快堆（RBEC-M）基准题对MGXSMC程序加工产生的核数据进行验证，计算结果表明，采用P5阶近似多群截面与连续点截面计算的有效增殖系数（keff）结果相差24 pcm（1pcm=10-5），而采用P0阶近似多群截面与连续点截面计算的keff结果相差较大。由此说明蒙特卡罗多群数据库的制作方法和所开发的程序是正确的，同时，中子各向异性散射对铅基快堆计算结果影响较大，故制作蒙特卡罗多群数据库时应加入中子散射角数据。 

Research on Method of Generating Monte Carlo Multigroup Library for Fast Reactor

Based on the discrete angle method, a Monte Carlo multi-group cross section generation program MGXSMC was developed. This program can read the cross section data from an input file or read the cross section from a library in a specified format to generate the multi-group cross section for MCNP or RMC. The corresponding index file list can be automatically generated. The two-dimensional two-group IAEA pressurized water reactor (PWR) benchmark and lead-based fast reactor (RBEC-M) benchmark were used to verify the cross section library generated by the MGXSMC program. The calculation results show that the difference between the calculated result of the P5 order approximate multigroup section and the continuous point cross section is 24 pcm (1pcm = 10-5), and the difference of the keff result calculated by the P0 order approximate multigroup section and the continuous point section is large. This shows that the method and the program developed for the Monte Carlo Group Section Library are correct. At the same time, the neutron anisotropic scattering has a large impact on the calculation results of the lead-based fast reactor. Therefore, when the Monte Carlo Group Section library is produced, the neutron scattering angle data should be added.