On Flow Distribution and Heat Transfer in a Graphite Moderated Molten Salt Reactor

Abstract

Graphite moderated nuclear reactor cores experience irradiation induced dimensional change as a function of graphite temperature and fast neutron fluence. In this paper, the impact of changes in channel geometry of a liquid fueled Molten Salt Reactor (MSR) is simulated with Monte-Carlo method based neutronics. Thermal-hydraulic phenomena are subsequently investigated with system code thermal hydraulics and Computational Fluid Dynamics (CFD). Specifically, gaps between graphite blocks introduced by irradiation-induced shrinkage are investigated, demonstrating the corresponding neutronic penalty on reactivity, and risk of semi-stagnant flow for a specific core lattice design. A comparison of CFD and 1D thermal hydraulics demonstrates the effect of flow entrainment between channels and gaps, and buoyancy assisted mixed convection in regions of laminar flow. Results indicate that the fuel temperature in gaps is approximately 25% higher than fuel in the adjacent flow channel for a representative 150 MWth design.