Neutronic Analysis of MOX-1000MWth Fast Reactor by two-step system code MCS/RAST-F

Abstract

RAST-F is a novel full-core analysis code that adopts a two-step approach, combining a multi-group cross-section generation Monte Carlo code (MCS) with a multi-group nodal diffusion solver. In this paper, we aim to illustrate the viability of utilizing the MCS/RAST-F combination for the analysis of fast reactors. We present verification results for the coupled nodal code applied to the MOX-1000MW thermal SFR benchmark cores. We employ two different multi-group cross-section calculation methods to enhance the agreement between the nodal and reference solutions. The reference solution is derived from the MCS code, utilizing continuous-energy nuclear data. our analysis reveals a high level of agreement between MCS/RAST-F and the reference solution. Notably, We observe discrepancies of less than 100 pcm in keff. This study affirms the reliability and effectiveness of the two-step approach, MCS/RAST-F, as a robust tool for the three-dimensional simulation of reactor cores with fast spectra.