Development of a sub-channel thermal hydraulic analysis code and its application to lead cooled fast reactor

The most important candidate for 4th generation nuclear system is lead or lead alloy cooled fast reactor. In the design phase of a lead cooled reactor, it is a top priority to finish the analysis work for the core. The detailed sub-channel analysis code KMC-Sub (Keda multi-physics and multi-scale coupling platform) has been developed to analyze steady state thermal hydraulic issues of SNCLFR-100 (Small Modular Lead-cooled Natural Circulation Fast Reactor, which was designed by University of Science and Technology of China). The model used in this code had taken the influence of cross flow into account, both forced and natural circulation can be simulated, to assess the development status of KMC-sub, experimental data from ORNL 19 pin tests (sodium cooled) and CAS 61 rods test (lead bismuth eutectic cooled) are compared to results from the code. The author found that in most flow rate and power density regime, the results coincides well with the tests data. After the V&V work, the code was used to analyze the flow and temperature distribution in important assemblies of SNCLFR-100 core, which showed that the design is reasonable and feasible.