Extended development of a Monte Carlo code OpenMC for fuel cycle simulation of molten salt reactor

Abstract Adopting liquid fuel and on-line reprocessing scenario are the unique features of molten salt reactor (MSR). Considering lattice code for thermal reactor is not suitable for MSR calculation, batch-wise fuel salt reprocessing scheme and “two-step” deterministic calculations cause some approximations in simulation of MSR on-line continuous fuel reprocessing, this study focuses on the development of a direct calculation method for MSR fuel cycle simulation based on an open source Monte Carlo code OpenMC. A fictitious decay constant and an external source term were introduced into traditional burnup equation for accurate simulation of on-line fuel reprocessing. Finally, a MSR fuel cycle simulation code OpenMCB-MSR was developed by coupling OpenMC and burnup code with Python script. Two pressurized water reactor (PWR) pin cell benchmarks without considering fuel on-line reprocessing, and one actual MSFR fuel cycle benchmark considering fuel on-line reprocessing were used for verification of OpenMCB-MSR. By comparison with reference results, the maximum differences of kinf calculated by OpenMCB-MSR are no more than 400 pcm both for two PWR pin cell benchmarks. Moreover, apart from isotopes with small absolute nuclide number density, all the other isotopes' nuclide number densities show a difference of less than 10%. For MSFR benchmark, the reactivity coefficients, isotope mass evolutions during fuel cycle process, mass evolution of fission products in core and breeding ratio calculated by OpenMCB-MSR show a good agreement with reference results. At equilibrium state, apart from isotopes with very small absolute mass, the differences of other isotopes' mass are no more than 5.0% in comparison with reference results. Benchmark verification results indicate that the method of MSR fuel cycle simulation based on OpenMC code is correct, and OpenMCB-MSR has capability of MSR fuel cycle analysis.