Neutronics and thermal-hydraulics coupling analysis using the FLUENT code and the RELAP5-3D code for a molten salt fast reactor

Abstract

The present study proposes a method for analyzing the thermal hydraulics coupling with the neutron point kinetics of a molten salt fast reactor. First, it has been analyzed by the FLUENT code that the flow behaviors in the reactor under the steady state conditions are greatly influenced by the installation condition of the inlet piping. Among them, the case where the inlet pipes are not offset is selected as a representative case. Next, the method of coupling the neutronics and thermal-hydraulics using UDF of the FLUENT code is explained when a transient analysis is conducted. In the case of using the system code RELAP5-3D, it is explained that thermophysical property binary files of the molten salts are generated, and the kinetic parameters for the implemented point kinetics model are modified in consideration of the inflow and outflow of delayed neutron precursors. A ULOF event has been simulated by the FLUENT code and the RELAP5-3D code, and it has been confirmed that the short-term evolutions in the comparison of reactor power and outlet temperature by the two codes show similar trends. It has been confirmed that when all the pumps are tripped, the core average temperature increases and the negative temperature reactivity coefficient feedback reduces the reactor power to the decay heat level. An event of single pump trip is also calculated with two codes, and it has been confirmed that the molten salt fast reactor has self-control characteristics on the basis of the negative temperature reactivity coefficient.