Development and Verification of a Few-Group Parameters Calculation Code TMSR-LINK for Molten Salt Reactor

Abstract Assembly few-group cross-sections are important parameters for nuclear design of reactor with the ‘two-step’ strategy based on deterministic theory. The unique physical mechanism and operation mode of liquid-fueled Molten Salt Reactor (MSR) make the calculation method of assembly few-group cross-sections different from that of the traditional solid-fueled reactor. In this work, a few-group parameters calculation code for MSR named TMSR-LINK is developed based on OpenMC, and two single assembly benchmarks of pressurized water reactor and MSR are first verified and analyzed; Simultaneously, macro cross-section libraries of the Molten Salt Reactor Experiment (MSRE) are prepared and provided to the MSR specific neutronics/thermal-hydraulics (TH) coupling code TMSR3D based on advanced nodal method, by which the reactivity loss caused by fuel flow and temperature distributions of graphite and fuel salt in the hottest channel are calculated and compared with the experimental data and the results from similar code; Finally, the steady-state characteristics of MSRE under different operation power are also probed. The results indicate that TMSR-LINK can provide a reasonable description for the benchmarks, which verified the accuracy of the developed code. The neutronics-TH coupling simulation results of the whole core for MSRE agree well with the experiments and exhibit a comparable precision against the similar code, this means that TMSR-LINK can provide accurate macro cross-section data for the MSR specific neutronics-TH coupling analysis code.