Thermal-hydraulic Analysis Of A Nuclear Research Reactor Core Channel

The thermal-hydraulic nuclear reactor core channel analysis is done thanks to the conservation equations of mass, momentum and energy, for an incompressible fluid. The set of equation is solved numerically using the finite volume method. This approach is applied for a 02 MW and for a 10 MW nuclear research reactor. The temperature profiles of the coolant and the clad along the channel are plotted. For the case of 02 MW and for an upward flow, the obtained results were compared to those given by Boudali and Salhi and to those given by the code TERMIC. For the case of 10 MW and a downward flow, the obtained results were compared to the results given by Lu et al. and also to those given by the code TERMIC. The obtained results are very close to those obtained by the cited authors and the calculated relative differences are minor. The results obtained thanks to this method, are more conservative than the results given by the presented comparative studies.