Transient thermal-hydraulic analysis for fuel assembly blockage accident by coupling CFD code with the neutron kinetics model

Abstract

Lead-based reactor coolants have many excellent characteristics but the coolants may seriously erode the fuel cladding and the structural materials and the flow area will be reduced by the corrosion products. This phenomenon is called blockage accident.In this paper, the blockage accident of single subassembly for a lead-based reactor has been analyzed by applying the commercial computational fluid dynamics (CFD) software coupled with the point reactor kinetics model. Firstly, the neutronics/thermal-hydraulics coupling model was established by user defined function (UDF) of FLUENT software based on the point reactor kinetics function with one group delayed neutron. Then the neutronics/thermal-hydraulics coupling model was validated with a single fuel pin model by comparing the fuel power, the cladding temperature and the coolant temperature with the calculation results of the RELAP5 code. Secondly, after considering mesh independence, turbulence models and other factors, the steady state of a single subassembly with 61 fuel pins at full power for a lead-based reactor designed by INEST/FDS Team was simulated in the normal operating conditions. Finally, a single subassembly for the center blockages of the flow channel was simulated by the neutronics/thermal-hydraulics coupling model. The temperature distributions and the maximum temperature were compared with that of the normal operating conditions and the effect of axial position for blockage was analyzed. It shows that the core outlet temperature and the highest temperature of cladding are increasing after blockage accident happening. And the highest temperature of cladding is lower calculated by neutronics/thermal-hydraulics coupling model than the condition without considering reactivity feedback. And the maximum cladding temperature for the 500 mm axially location blockage accident is higher.