Study On Thermal-Hydraulic Of Fuel Assembly Of AP1000

A nuclear power plant always shows very complex thermal-hydraulic behavior no matter under normal or accident conditions. It is compact in structure and actually a kind of heat source with a large core volume power density. Therefore, it is necessary to improve the reliability of a nuclear reactor to ensure that its core will not be broken under general accident conditions, and that no radioactive materials leak out of the core in severe accidents.A reliable analysis of thermal hydraulic of a nuclear reactor is required for obtaining more accurate core thermal parameters in stable operation and accident process. By using of the computational fluid dynamics program, Fluent, and one-dimensional sub-channel program COBRA-IV, the distributions of thermohydraulic parameters in fuel assemblies were obtained and studied in this article. The results given by Fluent and COBRA-IV were compared to verify the accuracy of Fluent in the three-dimensional calculation. Firstly, the variation of the outlet temperature of the 3×3 assembly was obtained by the two softwares when the loss of coolant accident occurs, and then a calculation was carried out to get the distribution of temperature and velocity fields under steady condition in the eighth fuel assembly of AP1000.The effect on flow resistance arising from the spacer grid was studied preliminarily and the drag coefficient of the spacer grid was calculated under different turbulence models.The calculated results by Fluent and COBRA-IV show that the trend of temperature distribution among the subchannels is consistent. And the deviation of outlet temperature is below 1%. By combing Fluent and COBRA-IV, it is easy to display the distribution of the axial flow velocity inside the subchannels and the impact of the spacer grid on lateral flow.And the lateral and axial velocity can be predicted more accurately. Due to the simplification of the model and the lack of related parameters, although the Drag coefficient of spacer grids calculated by Fluent has large deviation from the specified value in the COBRA-IV file, it still could be referenced to a certain degree.In general, the comprehensive result in three-dimensions as well as local characteristic of the thermal fluid can be obtained by using of Fluent and COBRA-IV. The distributions of temperature inside the fuel rods and the DNBR can be obtained quickly and efficiently too.Therefore, the efficiency can be improved with combining the two softwares, and the calculated results by this method can meet different requirements in the analysis of thermal hydraulic.