Neutronic analysis of european sodium-cooled fast reactor using Monte Carlo

Abstract

The design and safety analyses of current and future nuclear reactors require continuous improvement of computational accuracies. To this end, multi-physics approaches including detailed, coupled neutronic, thermal hydraulic, and fuel pin mechanic assessments are being increasingly used. In this work, we have studied the neutronic performance of a 3600 MWth oxide-fuelled sodium-cooled fast reactor core developed in the framework of the Euratom 7th Framework Programme Collaborative Project on European Sodium Fast Reactor. A new version of Monte Carlo neutron transport code (MCNP6) was used to evaluate a number of safety-relevant characteristics. These included the effective neutron multiplication factor and pin-by-pin power distribution in the ESFR core. As such, the peak power pin was identified and its axial power profile determined. These characteristics were evaluated both for the beginning of life as well as the end of cycle situations, explicitly taking into account changes in core material composition during burn-up. Obtained results will serve as an input to detailed safety assessments of fuel pin behavior in nominal and off-nominal conditions.