Sensitivity analysis of feedback factors by neutronics and thermal-hydraulics coupling code based on FLUENT software

With the great improvement of computer performance, analyzing the complex flow and heat transfer phenomena under transient condition by coupling CFD and neutronics has attracted lots of attention nowadays. We developed a neutronics-thermal hydraulics coupled code for transient analysis of pool type lead cooled fast reactors based on FLUENT UDF. The coupled code FLUENT/PK was validated by critical and sub-critical reactor under unprotected transient condition. The validation results show the correctness and feasibility of CFD method in safety analysis of reactors. The coupled code was used to analyze the lead cooled fast reactor SNCLFR-100 (which had been proposed by USTC) under different transient conditions. The calculated results show that there exists great difference if the introducing time of reactivity is different. The sensitivities of the feedback factors were analysed for the reason that these factors will be quite different at different operation time of the reactor core, and the result shows that Doppler constant has the most important effect on reactor safety compared with coolant temperature coefficient, fuel rod axial expansion coefficient and cladding expansion coefficient.