Neutronic Modeling of Molten Salt Reactors

Abstract

The main objective of this study is to obtain an analytical solution to multi-group diffusion equations of MSRs. As a starting point, one group criticality calculations including delayed neutrons are carried out both for no-recirculation and recirculation cases, respectively. The effect of fuel's velocity and recirculation loop length on the criticality of the system as well as the difference between recirculation and non-recirculation systems on the delayed neutron precursors is investigated. Then, neutronic modeling has been done for FUJI-233Um (450 MWth) MSR using MCNP-IV. The analytical solution employed in the one-group diffusion theory of MSR is extended to the multi-group case.