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

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 OpenMC. 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.