The Transfer of Xenon-135 to Molten Salt Reactor Graphite

Abstract

The term “Molten salt reactors” refers to a broad class of nuclear reactors that use a molten alkali-halide salt as the primary coolant fluid. This paper pertains to thermal spectrum liquid fuel molten fluoride salt reactors with graphite moderator (MSRs), where the molten salt also dissolves the actinide fuel. Xenon isotope 135,135Xe, is a fission product that is produced during nuclear fission energy production, and it acts as a neutron poison. Due to the circulating nature of the fuel salt in MSRs, there is a qualitative difference in the behavior of 135Xe in an MSR compared to a solid fueled reactor. Some of the 135Xe produced in fission may end up in the pore space of the porous graphite moderator used in an MSR. This paper examines the transfer and storage of 135Xe in MSR graphite. Prior publications are reviewed, the porosity of the MSR graphite is examined, governing equations are detailed, film layer production and destruction are discussed, the graphite/salt interface is explored, transport pathways are considered, transfer processes are exposited, the effect of charged species is examined, the solubility of noble gases in molten fluoride salts is examined, the mass diffusion coefficient in molten salts is explored, and the calculation of mass transfer coefficients is described.