摘要
During the simulation of AP1000 nuclear power plant,the values of input parameters, core nodalization methods and calculation models, may have important influence on the code outputs. Therefore, it is necessary to identify and evaluate the influence of these parameters and modeling approaches quantitatively. Based on the best estimate thermal-hydraulic system code RELAP5,sensitivity analyses have been performed on core partition methods,parameters and models in AP1000 nuclear power plant,such as the core channel number,pressurizer node number,and feedwater temperature. The results show that code channel number,code channel node number, and the pressurizer node number have apparent influences on the coolant temperature variation and pressure drop in the reactor. The feedwater temperature is a sensitive factor to the steam generator( SG) outlet temperature and the SG outlet pressure. In addition,the influence of the cross-flow model on coolant temperature variation and pressure drop through the reactor is insignificant,both in steady state and loss of power transient. Furthermore, some suitable parameters and modes also have been put forward for the nuclear system simulation.
During the simulation of AP1000 nuclear power plant,the values of input parameters, core nodalization methods and calculation models, may have important influence on the code outputs. Therefore, it is necessary to identify and evaluate the influence of these parameters and modeling approaches quantitatively. Based on the best estimate thermal-hydraulic system code RELAP5,sensitivity analyses have been performed on core partition methods,parameters and models in AP1000 nuclear power plant,such as the core channel number,pressurizer node number,and feedwater temperature. The results show that code channel number,code channel node number, and the pressurizer node number have apparent influences on the coolant temperature variation and pressure drop in the reactor. The feedwater temperature is a sensitive factor to the steam generator（ SG） outlet temperature and the SG outlet pressure. In addition,the influence of the cross-flow model on coolant temperature variation and pressure drop through the reactor is insignificant,both in steady state and loss of power transient. Furthermore, some suitable parameters and modes also have been put forward for the nuclear system simulation.
