Calculation of Neutronic Parameters using OpenMC for Potential Dispersed Fuels of MNSR

摘要

In view of Reduced Enrichment for Research and Test Reactor Program (RERTR) the study related to conversion of the Miniature Neutron Source Reactor (MNSR) core from High Enriched Uranium (HEU) fuel to Medium Enriched Uranium (MEU)/Low Enriched Uranium (LEU) fuel is carried out. The reactor was modeled and simulated using OpenMC code. The developed model was benchmarked using HEU core. The parameters compared were excess reactivity, control rod worth, and shutdown margin. Two potential fuels were used. MEU fuel used was 22.31% enriched U3Si2-Al and LEU fuel used was 19.81% enriched U3Si-Al. The resulted parameters were quite close to the original parameters values, means that these fuels can be candidate fuels for core conversion but before that kinetic parameters and reactivity temperature coefficients should also be calculated for stability check on these fuels.