Effects of self-shielding and scattering matrix treatments in detailed burnup calculation

山本 章夫*; 遠藤 知弘*; 渡邉 友章  

Yamamoto, Akio*; Endo, Tomohiro*; Watanabe, Tomoaki

A sensitivity analysis was conducted to evaluate the impact of nuclide-wise self-shielding and scattering matrix treatments in a detailed burnup calculation involving more than 4,000 nuclides. The analysis was performed using a newly developed neutronics analysis code, GenesisRPC, based on the neutron transport code GENESIS for two- and three-dimensional geometries. The results, obtained for UO and MOX fuel systems, indicate that for nuclides with number densities below 10 [1/barn/cm], the effects of self-shielding factors and the P0 (isotropic) scattering matrix are negligible. Additionally, the impact of the Pn (anisotropic) scattering matrix was found to be insignificant for nuclides with number densities below 10 [1/barn/cm] in the present calculation conditions, suggesting that only nuclides with high initial concentrations need to be considered.