摘要
堆芯材料中核素的平均宏观截面及反应率是影响堆芯倍增因子的一个重要因素,关于平均宏观截面的计算MCNP中没有介绍,但是MCNP能输出核素的中子截面图、中子能谱图上的点的x-y坐标及包含这些点的坐标文本文档。结合对一个基准体系中的球形均匀裸堆(Bigten)的Keff的蒙特卡罗模拟计算,研究了一种计算核素的平均宏观截面和反应率的方法——用C++编写的MFC程序实现对MCNP输出的中子截面图和中子能谱图上的点的坐标所对应的文本文档的读取、处理,根据多群理论,计算了核素的平均宏观截面,同时能得出核素的反应率。计算结果的相对误差都在10%以内,探讨了在不同数据库下堆芯倍增因子Keff变化的原因。
The average macroscopic cross sections of nuclides of the materials used in reactor is an important factor that affect the multiplication factor of a core. However, the calculation of macroscopic cross sections is not referred in MCNP. MCNP can display the x-y coordinates of the points and the documents about x-y coordinates of the points on the neutron nuclides cross sections and neutron energy spectrum. In this article, combining the simulation calculation of K off of a spherical homogeneous bare reactor （Bigten） of a benchmark assemblies with MCNP, we have studied an approach of calculating the average macroscopic cross sections of the nuclides -- using MFC to deal with the output files about the x-y coordinates of the points on cross sections and neutron ener- gy spectrum. According to the muhi-group theory, we calculate the average macroscopic cross sections of the nu- clides and the reaction rate at the same time. The relative errors of the results are all within 10%. The reason for the change of K eff under different data libraries are then investigated.
