There are different parameters such as increase or decrease of heat removal from primary circuit which affect the coolant mixing phenomena in the vessels of pressurized water reactors (PWRs). Determination of mixing level is very important from reactor safety and control aspects. In this study, the thermal hydraulic test of coolant mixing within the reactor pressure vessel of Bushehr nuclear power plant (BNPP) has been simulated (3-D modeling) using ANSYS CFX 18.0. In this test, the fluid mixing due to primary circuit heat removal decrease has been investigated and the goal of this research has been defined as finding the coolant temperature distribution, computing the primary circuit loops mixing coefficients and other thermal hydraulic parameters of coolant in the whole reactor zones specially the reactor core which is the most important. To achieve this, the geometry of whole reactor considering all components have been modeled and the governing equations of reactor flow field (the Reynolds Averaged Navier-Stokes equations utilizing SST k-ω turbulence model) have been solved in CFX. Comparison of simulation results and experimental results of BNPP startup test shows the average error of 6.45 % and 10.92 % for mixing coefficient of loops and core inlet, respectively. According to the implemented simplifications, the results have good accuracy.
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