Coupled neutronics and thermal-hydraulics simulation of molten salt reactors based on OpenMC/TANSY

摘要

An OpenMC/TANSY code system has been developed in this paper for the coupled neutronics and thermal-hydraulics simulations of MSRs. The homogenized cross section data library is generated using the continuous-energy Monte-Carlo code OpenMC which provides significant modeling flexibility compared against the traditional deterministic lattice transport codes. The few-group cross sections generated by OpenMC are provided to TANSY which is based on OpenFOAM to perform the full-core coupled neutronics and thermal-hydraulics simulations. In order to verify the OpenMC/TANSY code system, the Molten Salt Fast Reactor benchmark problem was calculated and both the neutronics results and thermal-hydraulics results were compared with those obtained by other researchers. For application of the OpenMC/TANSY codes sequence, the simulation of a representative molten salt reactor core MOSART has been performed. First, to verify the generation of the few-group cross sections, the neutronics results obtained by the "two-step" scheme were compared with those obtained by full-core Monte-Carlo solution. Good agreement can be observed for the multiplication factor as well as the power distributions. Then the full-core coupled neutronics and thermal-hydraulics simulation was performed. The distribution of the important neutronics and thermal-hydraulics parameters are presented and analyzed in detailed in this paper. For the further study of the characteristics of MSRs, several effects like the external-loop transit time, inlet velocity and inlet temperature on the effective delayed neutron fraction and critical fuel concentration have been analyzed. The numerical results indicated that the TANSY code with the cross section library generated by OpenMC has the capability for the steady-state analysis and reactor core design of MSRs. (C) 2017 Elsevier Ltd. All rights reserved.