Molten Salt Reactor Experiment Simulation Using ThorMOC

Abstract

Abstract High fidelity modeling and simulation of nuclear reactors with direct whole core transport calculation has become an important state of art for development of reactor computational tools with the rapidly improvement of computing power. In this work, a quasi-2D delayed neutron precursor (DNP) drift model has been implemented in a GPU-based whole core transport code ThorMOC to achieve high fidelity modeling and simulation of liquid-fueled molten salt reactors (MSRs). A rasterized coarse mesh finite difference (RCMFD) method is applied in ThorMOC to improve the convergence rate, and to handle complex MSR models with arbitrary boundaries by combining with long characteristics method. OpenMC is used to generate a multi-group cross section library and to provide reference results for stationary state calculations. The results including eigenvalue and power distributions from ThorMOC agree well with those from OpenMC for stationary state calculations of the Molten Salt Reactor Experiment (MSRE) model. The reactivity loss due to DNP drift is computed and is in a good agreement with the measured value in experiment.