MCS Multi-group Cross Sections Generation for Fast Reactor Analysis

Abstract

Current work introduces a brief methodology for multi-group (MG) cross sections (XSs) generation by Monte Carlo (MC) code MCS, which can be compatible with nodal diffusion code, PARCS. The applicability of the methodology is quantified on the sodium fast reactor ABR-1000 design with a metallic fuel loaded. The MG XSs generated by MCS with a 2D sub-assembly are well consistent with those of SERPENT 2. Furthermore, the solutions of beginning-of-cycle steady-state MG calculation of MCS/PARCS for a whole-core problem, including the core keff and power profiles, are compared to those of the MCS MC code. Overall, the code-tocode comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-meansquare error in assembly power less than 1.15%. Therefore, it is successfully demonstrated the employment of the MCS MG XSs generation for PARCS is a promising system to accurately perform neutronic analyses for fast reactors.