A Fuel Burnup Molten Salt Reactor Multiphysics Benchmark

Abstract

Molten Salt Reactors (MSRs) are being researched as possible reactors to be built in the next generation of nuclear reactors. The interest in building these reactors has required improvement in current numerical methods for modeling the physics of these reactors. Current codes will need to update their features to include the physics of molten salt reactors or new codes will need to be developed in order to account for these physics. A numerical benchmark for modeling MSRs has been created for the modeling of these physics, however this benchmark could be updated to include the physics of isotope depletion (burnup). The work presented in this thesis entails using the finite volume method in an open-source code, FiPy, to solve the partial differential equations required for the modeling of the generic MSR design in the original benchmark. Python scripts were then utilized to visualize the results produced by FiPy and compare them to the values obtained in the original MSR benchmark. Once the model in FiPy is verified against the original MSR data, the benchmark could then be extended to include the additional physics modeling of isotope depletion. The Monte Carlo code, Serpent, was utilized to generate burnup cross sections based on the original MSR benchmark parameters and these cross sections were inserted into the python code to extend the original benchmark to include burnup. The results from FiPy matched closely with the original benchmark results for the steady-state problems with little to no discrepancy between them. The burnup cross sections were generated to simulate a six month operating period and were input into the FiPy simulations to generate the change in eigenvalue for FiPy over the original fully coupled problem for an MSR. Results were obtained for these burunp calculations and suggest FiPy was capable of handling newly-generated cross sections to model the physics of isotope depletion in an MSR.