Depletion calculation of subcritical system with consideration of spontaneous fission reaction

Riyana, E. S.   ; Okumura, Keisuke ; Sakamoto, Masahiro ; Matsumura, Taichi ; Terashima, Kenichi  

Modification of the Monte Carlo depletion calculation code OpenMC was performed to enable the depletion calculation of the subcritical neutron multiplying system. With the modified code, it became possible to evaluate the quantity of short half-life fission products from spontaneous and induced fissions in the subcritical system. As a preliminary study, it was applied to the fuel debris storage canister filled with nuclear materials and spontaneous fission nuclides. It was confirmed that the code could successfully provide a quantity of short half-life FPs over time and provide the relationship between the activity ratio of Kr-88 to Xe-135 and effective neutron multiplication factor of the canister.