基于蒙特卡罗方法制作用于中子输运SN程序的多群截面库

Fabrication of Multi-group Neutron Transport Cross Section Library for SN Program Based on OpenMC

在反应堆计算中,确定论计算软件如ANISN计算速度快,适合复杂的物理-热工耦合计算任务。但对于确定论计算程序,其计算精度主要受制于多群截面数据库的制作。本工作基于蒙特卡罗计算软件OpenMC制作用于确定论程序的多群截面库。首先利用OpenMC进行建模计算,然后分区分能群统计总反应率、裂变中子产生率、吸收反应率、中子通量以及高阶勒让德散射率,最后通过自主编写的Fortran截面转换程序得到BUGLE-96格式的多群截面数据。为了验证所制作截面库的可靠性,将新制作的截面库提供给ANISN程序进行基准题计算,计算结果和蒙特卡罗程序及BUGLE-96库进行对比。结果表明,基于OpenMC和自主编写的截面转换程序制作的截面库用于ANISN计算时,K^(eff)和通量与蒙特卡洛程序计算结果相吻合,并且比使用BUGLE-96库计算结果偏差更小,验证了本方法制作中子输运SN程序的多群截面库的有效性。

In reactor calculation,deterministic calculation software such as ANISN is fast and suitable for complex physical thermal coupling calculation tasks.But for the deterministic calculation program,the calculation accuracy is mainly restricted by the making of multi-group cross section library.In this paper,based on the Monte Carlo software OpenMC,a multi-group cross section library for deterministic programs is developed.Firstly,modeling and calculation are carried out by using Open-MC.Then,the total reaction rate,neutron fission rate,absorption reaction rate,neutron flux and high-order Legendre scattering rate are calculated by different regions and energy groups.Finally,multi-group cross section data in BUGLE-96 format is obtained by Fortran cross section conversion program.In order to verify the reliability of the cross-section library,the new cross-section library is provided to ANISN program for benchmark calculation,and the calculation results are compared with Monte Carlo program and BUGLE-96 library.The results show that the K^(eff)and flux calculated by the cross section library based on OpenMC and the cross section conversion program are consistent with those calculated by Monte Carlo program,and the error is smaller than that calculated by BUGLE-96 library,which verifies the effectiveness of the multi-group cross section library of neutron transport SN program.