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.
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Disclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.