Accident Analysis And Sensitivity Study On AP1000 Losing AC Power

In this paper, the transient simulation and the sensitivity at the loss of AC power of the AP1000 nuclear power plant were investigated.A relatively complete model of AP1000 nuclear power plant was established by using RELAP5 program to analysis the reliability and safety of the Passive Safety System when losing. The Ac power The transient simulation showed that the nuclear reactor can shutdown safely when the signal of the main pump low speed appears; the Passive Safety System starts properly while the residual thermal and power of nuclear core can be removed safely; the main parameters of the primary coolant loop are within the safety margin. The accident analysis demonstrated that it is adequate to remove the residual thermal through the safety valve, atmospheric relief valve of the stream generator, the natural circulation cooling system, and the Passive Residual Heat Removal System (PRHR). The damage of nuclear fuel and cladding is evitable and the pressure of RCS would not exceed the regulation. The PRHR can ensure the safety of AP1000 nuclear power plant under the loss of AC power accident condition.The correctness of the model established in the paper was validated by comparing the transient simulation results of RELAP5 with the calculations of Westinghouse LOFTRAN program. The sensitivity of model and the sensitivity of primary parameters were analyzed. The reactor core nodal and the steam generator heat-conduction nodal were altered to investigate the influence that different models have upon the nuclear power plant's safety. And in order to clarify the influence that operating parameters have upon nuclear power plant, the primary parameters of heat transfer area of the PRHR C-type heat exchange tube and the average operating temperature of the primary loop were altered.Model Sensitivity studies showed that the number of core nodal has relatively significant influence on the calculation of steady-state temperature. With the reduction of the number of core nodal, steady-state temperature of the primary loop would increase while the stead-state operating condition would deviate; the hysteresis of the fuel assembly thermal release would happen after the loss of AC power, and the primary loop system pressure drop speed would be reduced, CMT startup would delay while the maximum flow rate would increase.; the speed of the fall of the steam generator water level would accelerate, PRHR would start ahead of time while the maximum flow rate increase too. The section of the nodal of steam generator tube has small influence on steady-state conditions. The startup of CMT and PRHR is not affected by the section of the steam generator tube, the flow is almost unchanged, no effect happens on passive safety systems. The comparison between the heat exchange capabilities of the steam generator showed that the section almost has no influence on the heat exchange capability.Changing the passive residual heat removal heat exchanger area was found having no effect on the steady-state of AP1000 nuclear power plant. Under accident condition,CMT systems would stay unchanged, PRHR would operate ahead of time while its maximum flow rate would increase, and RCS system's maximum temperature would not change. The sensitivity studies showed that the startup of CMT would delay,PRHR would operate ahead of time, and the maximum flows of CMT and PRHR would have significantly increase; RCS would have corresponding increase in system peak temperature while the speed of the fall of the temperature would reduce;compared with the data in the accident analysis, the thermal taken away by steam generator would increase, and finally result in the increase of the time for the thermal equilibrium.