Validation of Nuclide Depletion Capabilities in Monte Carlo Code MCS using Spent Fuel Isotopic Measurements

Abstract

As part of the validation of the three-dimensional (3D) continuous-energy neutron-photon transport Monte Carlo code MCS, isotopic predictions from MCS depletion calculations are compared to post irradiation examination (PIE) data from uranium dioxide (UO2) fuel samples obtained from Takahama-3 pressurized water reactor (PWR) spent nuclear fuel (SNF). The depletion simulations are performed with the ENDF/B-VII.1 library and an equivalent pin cell model. The isotopic inventory predicted are compared to measured actinides/fission products concentrations and analyzed. For most of the nuclides, MCS predictions agree within ±8% of the experiment. Preliminary results obtained with the ENDF/B-VIII.0 library show improved agreement for 241Am, 245Cm and 154Eu, due to the one-group cross-sections. The performance and accuracy of the MCS results are further demonstrated through comparison to the deterministic code STREAM.