Numerical Simulation of the Effect of Molten Salt Channel Number on Temperature Distribution in the Molten Salt Core Reactor

Abstract

Abstract A critical aspect in the design of a nuclear reactor that needs to be considered is thermal analysis. This parameter relates to molten salt circulation, which serves as a fuel, and coolant in Molten Salt Reactor (MSR) core reactors. This study simulates the temperature distribution performance of the MSR core by considering the effect of the number of molten salt channels with inlet temperature variations. Knowing the temperature distribution will predict the characteristics of heat transfer in the form of natural convection for the survival of the reactor. The reactor used a Molten Salt Breeder Reactor (MSBR) design developed by Oak Ridge National Laboratory (ORNL). The Ansys software was used to simulate a 3D model under steady-state conditions. The geometry was divided into two regions: solid (graphite material) and fluid (molten salt). The thickness of the moderator and radius of the molten salt channel were 0.0508 and 0.0208 m, respectively. The moderator height was 1.98 m. The number of molten salt channels varied up to four. The results showed a temperature difference of 9 to 15 K when the channel increased from one to four at 839 K inlet temperature.