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Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD) studies on the thermal hydraulic analysis of single-phase flow in the rod bundles of nuclear reactor core. On account of the tremendous amount of publications related to each aspect above, this brief review would focus on the most representative researches in the past decades. The aim of this paper is to provide an outline and present a comprehensive overview of the CFD research and guidelines on single-phase flow through bare rod bundle rod with mixing vane spacer and wrapped wire spacer in reactor core. In addition, the analysis and validation methods for various CFD calculating methods are concluded. The applications of CFD approach with different type of rod bundles are summarized. Finally, some remaining challenges for future work are proposed and discussed. [ABSTRACT FROM AUTHOR]

In order to explore the core corrosion characteristics of horizontal lead-bismuth reactor in oxygen-controlled environment, a liquid lead-bismuth eutectic (LBE) corrosion model including the growth and removal process of oxide layer was established in this paper. The coupling method of double-layered oxidation model and computational fluid dynamics (CFD) was used to simulate the long-term oxidation corrosion of horizontal LBE reactor core numerically. The results show that the average total oxide layer thickness of fuel rod surface is 1.59 μm, and the thickest oxide layer is 3.48 μm on the fuel rod surface at the core outlet after 20,000 h. The fuel rods maintain a double-layer oxide layer structure. The growth rate and removal rate of the oxide layer on fuel rod surface increase with the increase of inlet temperature and the decrease of inlet flow rate. The growth rate of the oxide layer on fuel rod surface increases with the increase of inlet oxygen concentration, while the removal rate decreases. At higher temperature above 420℃, lower flow velocity below 0.25 m/s and lower oxygen concentration below 2 × 10−8wt%, the magnetite layer will first be completely removed from the surface of the fuel rods at the core outlet. The liquid LBE oxidation corrosion model and coupled CFD method presented in this paper can be used for numerical simulation of oxidation corrosion in different units of LBE cooling system. [ABSTRACT FROM AUTHOR]

Thermal-hydraulic performance in the reactor core was an essential factor in the nuclear power plant design. In this study, we analyzed the three-dimensional (3-D) thermal-hydraulic characteristic of reactor core based on porous media method. Firstly, a 3-D rector pressure vessel (RPV) model was built, including the inlet leg nozzle, downcomer, lower plenum, reactor core, upper plenum and outlet leg nozzle. Porous media model was used to simplify the reactor core and upper plenum. The commercial CFD code ANSYS CFX was employed to solve the governing equations and provide the 3-D local velocity, temperature and pressure field. After appropriate parameters and turbulent model being carefully selected, the simulation was validated against the 1:5 scaled steady-state hydraulic test. The predicted hydraulic parameters (normalized flowrate distribution and pressure drop) were in good agreement with the test results. And the predicted thermal parameters agreed well with the designed values. The validation indicated that this method was practicable in analyzing the 3-D thermal-hydraulic phenomena in the RPV. Finally, the thermal-hydraulic features in reactor core were analyzed under the condition of the Steam Generator Tube Rupture (SGTR) accident. The simulation results showed that the coolant temperature increased gradually from the center to the periphery in the reactor core in the accident. But the temperature decreased to safety level rapidly after the reactor shutdown and safety injection operation. The reactor core could keep in a safe state if appropriate safety operations were performed after accidents. [ABSTRACT FROM AUTHOR]