Phenomenal Investigations of the Thermal-hydraulic Responses of Multi-Dimensional Simulation and Modeling in Core and Downcomer of L2-5 Test of LOFT Using RELAP5-3DK
In this study, RELAP5-3DK code is used to simulate the L2-5 test of Loss of Fluid Test (LOFT) facility. RELAP5-3D is a multi-dimensional reactor system thermal-hydraulic analysis code. The conservatism evaluation models of LOCA analysis specified in 10CFR50.46 Appendix K are incorporated into RELAP5-3D code and the new code is named RELAP5-3DK. In the present simulation, the core and downcomer of the test facility were modeled as interconnected three dimensional components. The results of the simulation were compared with the results of the RELAP5-3DK one-dimensional analysis. Four analyses were performed to study the impact of multi-dimensional modeling of core and downcomer on the results of LOCA analysis. The results show that when both core and downcomer are modeled three-dimensionally, the predicted PCT was lowered by 60℃. The cladding surface temperature drops significantly earlier in the 3D calculation than that in the 1D calculation. Neither 1D nor 3D calculation reproduces the experiment results. It seems that the predictions of the 1D calculation are closer to the experimental results. Modeling core and downcomer three-dimensionally has significant impact on the flow behaviors of coolant in the reactor vessel. There is a tendency that coolant flows into the hot channel from lower plenum. The tendency can be attributed to a phenomenon called chimney effect. This effect is even stronger when downcomer is also modeled three-dimensionally. Larger amount of coolant flows into the channel reduces the PCT. Modeling downcomer multi-dimensionally changes the behaviors of circumferential flow in the annular region between vessel wall and core baffle substantially. The bypass flow from the intact side to the broken side of downcomer is significantly reduced. This phenomenon increases the amount of downward flow of coolant into the lower plenum and reduces the accumulative break flow. The impact of 3D core modeling on the bypass flow and accumulative break flow is relatively minor.
Nuclear Safety Transient Analysis LOCA
M. Lee K. Y. Chiang
Department of Engineering and System Science National Tsing Hua University No. 101, Section 2, Kuang Department of Engineering and System Science National Tsing Hua University No. 101, Section 2, Kuang
国际会议
上海
英文
741-753
2010-10-10（万方平台首次上网日期，不代表论文的发表时间）