Neutronic calculations and thermalhydraulic application using CFD for the nuclear research reactor NUR at steady state mode

Abstract

Neutronic calculations are performed for the X-1 configuration of NUR nuclear research reactor core, followed by a CFD thermalhydraulic modeling of the reactor core hot channel. The neutronic study is established by considering the nuclear reactor core as being composed of homogenized zones. Calculated power densities in the fuel elements are used to identify the hottest channel in the nuclear reactor core and to determine the power peaking factors. To do this, the WIMSD-CITVAP neutron calculation line was used. This was followed by a 3D thermalhydraulic modeling. The mathematical model is based on equations of conservation of mass, momentum and energy. These equations are coupled to a k−ω SST turbulence model. The calculated power peaking factors are introduced into the power distribution equation which is applied to the outer area of the fuel meat. This equation set-up is solved, on a numerical domain mesh, using the FLUENT CFD code. In order to validate the method developed in this work, an application was made and the temperature distribution profile of the cladding was plotted. A comparison with a published work has shown that the calculated relative differences do not exceed 2.4 % and that the temperature profile obtained in this work is the most conservative. A parametric study was conducted to confirm the physically valid behavior of the model developed in this work and the obtained results also confirm that Onset of Nucleate Boiling (ONB) and cladding melting cannot occur under the conditions of the present study.