Advanced Calculation System Using Monte Carlo Method for Analyses of Isotopic Composition of Spent Fuel and Radiation Flux Distribution in BWR RPVs

摘要

Toshiba is currently developing an advanced calculation system using the Monte Carlo transport method for analyses of the actinides and fission product nuclides of spent fuel and the radiation flux in the reactor pressure vessel (RPV) of a boiling water reactor (BWR). Detailed determination of actinides and fission product nuclides is important in criticality evaluation for the transportation and storage of spent fuels. We have therefore developed the MCNP-BURN2 parallelization Monte Carlo burnup calculation system for this purpose. Using the design code and post-irradiation examination (PIE) analysis, we confirmed the accuracy of MCNP-BURN2. The radiation flux in the BWR RPV is calculated by the TORT (S_N method)/MCNP (Monte Carlo method) coupling method. In this method, the radiation angular flux distribution on the core surface is obtained by TORT, so the calculation for the outside of the core using MCNP is carried out with sufficient sampling at the source and the exact model for the outside structure of the core.