A new fission matrix correction method to estimate the source distribution in nuclear reactor core

The Monte Carlo method has been widely used in nuclear reactor physics. However, the traditional Monte Carlo method suffers from slow convergence and statistical uncertainty, usually leading to a long computation time. The fission matrix acceleration method has been proposed to accelerate the traditional Monte Carlo criticality calculations, however, it still requires significant computational resources to build fission matrices. A new approach, used in the RAPID code, utilizes a fission matrix combination method to estimate the true fission matrix by combining a set of pre-calculated fission matrix coefficients. Except for the Monte Carlo calculation of the pre-calculated fission matrices, the combination does not rely on the Monte Carlo simulation and saves a large amount of computational costs for subsequent calculations. However, the fission matrix combination method can have significant errors in estimating the source distribution especially with a large enrichment gap between fuels. In this paper, we will explain the issues of the fission matrix combination method in source distribution calculations and present a new method to correct the combined fission matrices. The method is tested on a model with two different enrichment assemblies to calculate both the multiplication factor and the source distribution. The source distribution derived from the corrected fission matrix shows a good agreement compared to the reference, which is the Serpent calculation.