(untitled) (Brazilian Journal of Radiation Sciences (BJRS), 2023)
The suitable cooling of fuel elements in a nuclear reactor is an important requirement that must be met to avoid that the fuel temperature rises above the safety limits according to the reactor design. During the reactor operation, there are transients that could lead to a temperature that overcomes these limits, such as those related to the cooling system. The CFD codes are tools that could aid in the understand of the phenomenon during such transients, allowing to access details of the flow that are not possible, or are possible only with limitations, by using other kind of codes or experiments. In the present work, the results obtained using ANSYS-CFXÒ code for the IEA-R1 reactor during a steady state and transient of slow loss of cooling accident event are presented. The results obtained shown a good agreement with experimental data reported and works that used this reactor as case of study. These results are part of a research in which the main objective is to simulate the flow of the coolant in the fuel element channels during transients. These results would support an initial analysis of the flow during the transition from forced to natural convection that occurs when the coolant flow falls below a settled value and the valve on the bottom of the core opens by gravity, aiming to understand better phenomena involved and the limitations of the models available in the ANSYS-CFXÒ code.