Improvements and New Findings in Monte Carlo Method with Complex-valued Weights for Neutron Leakage-corrected Assembly Calculations

Kyoto University, Research Reactor Institute, 2-1010, Asashiro Nishi, Kumatori, Sennan-gun, Osaka, 590-0494, JAPAN

^* Corresponding Author, E-mail: toshihiro.yamamoto223@gmail.com

Abstract

The author of this paper recently proposed a Monte Carlo calculation algorithm to solve a complex transport equation with complex-valued weights. The algorithm enables one to generate neutron leakage-corrected group constants and anisotropic diffusion coefficients for a unit fuel pin cell or assembly. The group constants are subsequently used for multi-group deterministic core calculations. The technique, however, had some limitations in applying itself to general problems. Some improvements have been done in this paper. The reflective boundary condition has newly become available. It has been found that a cumbersome weight cancellation of fission sources with positive and negative weights can be omitted in general fuel assembly geometries. A homogenization method of diffusion coefficients for a fuel assembly has been proposed.