Analysis Of Pwr Nuclear Reactor Core Using Subchannel Method

Pressurized water reactors are mostly chosen for nuclear power plant throughout the world. They take up 60% of the reactors being used in the world. In China, up to 87% of the nuclear reactors are pressurized water reactors. Detailed thermal hydraulic analysis is not only important to nuclear reactor design purpose, but also important to the safety of the reactor's operation. Recently, subchannel method is used for detailed analysis of the reactors, one of which is the subchannel computer program series COBRA developed by Pacific Northwest Laboratories in USA. In China, these programs are used for thermal hydraulic analysis process in many operation and research companies in nuclear industry. However, the self-design capability of the nuclear island, which is one of the key PWR technologies for the long term development, should not only depend on commercial software developed in foreign country, but also in need of program or software suitable for PWR core's thermal hydraulic analysis developed in our country.A computer program based on subchannel method was developed for thermal hydraulic research for PWR core. The computer simulation based on 900MW PWR nuclear power plant was established in this article. Certain concentric subchannels were divided from center to the boundary of the core, along the cross-section of core's fuel assembles and coolant tunnels. The subchannel's continuity conservation, momentum conservation and energy conservation were established. The turbulent and diversion crossflow were considered between adjacent subchannels. All subchannels were correlated and computation was made from the core's inlet to core's outlet. The coolant's mass velocity, enthalpy was obtained in each subchannel along core's axial direction, therefore the surface temperature of the fuel rods were obtained along core's axial direction as well.By using the theoretical model and computer program being developed, the steady-state computation and transient-state computation with regard to rapid core's power surge concerning boron dilution accident, which based on 900MW PWR nuclear reactor, were established. 100% rating power were considered in the steady-state computation. The results of steady-state computation basically reconciled with the concerned published PWR operational data. Transient-steady were made based on the assumption of the core's power surge to 300% rating within 2 minutes. The core's parameters variation progress within the concerning time were acquired. Results showed that film boiling emerged when power reached about 271% rating power. As a result, the heat transfer deterioration made peak temperature on the fuel rods surface about 2 times higher than its value in normal operation condition.