CFD study of two-phase cross flow and heat transfer in sub-channels of pressurized water reactor fuel assemblies

Abstract

Abstract The safety of nuclear power systems is extremely important. The development of nuclear power is predicated on the safety of the system. Under certain operating conditions, the coolant flowing through the fuel assembly will undergo phase change, resulting in gas-liquid two-phase flow. The existence of the two-phase flow will change the stability of the coolant flow and thus affect the reliability of the core instrumentation under accident conditions. The presence of the two-phase flow will also change the heat transfer coefficient of the coolant, thus affecting the temperature distribution on the surface of the instrument guide tube. This paper adopts the computational fluid dynamics method. On the one hand, this paper calculates the two-phase cross flow of the two-channel. By changing the inlet boundary conditions to achieve the elimination of diversion cross flow, the two-phase cross flow phenomenon was observed, which laid the foundation for future experiments. On the other hand, calculations of the surface temperature of the instrument guide tube. The surrounding flow field show that the temperature of the surface of the instrument guide tube is higher in two phases than in one phase. The gas phase moves to the surface of the instrument guide tube.