Thermal-hydraulic and Safety Analysis of Space Hundred-kilowatt Lithium-cooled Fast Reactor

As the increase of deep space exploration requirements, space highpower level nuclear reactor system is imperative. In this paper, for the space lithium-cooled fast reactor combined with the Stirling cycle, the models of core, Stirling generator, radiation radiator, pump and related pipeline were established. A system thermalhydraulic analysis code was developed based on the Fortran language. The fidelity of the Stirling model was verified by the experimental data with the maximum relative error of 17.3%. The space lithiumcooled power system model was established. The system code was verified by comparing the steadystate calculated value with the design value with the maximum relative error of 13.3%. The transient analysis of the typical reactor accident was analyzed. The results show that the system has good safety performance due to the negative reactivity feedback of the core and the peak fuel temperature is within the safety limit. This paper provides theoretical support for the thermalhydraulic and safety analysis of the hundredkilowatt space reactor.