Thermo-hydraulic analysis of the supercritical water-cooled reactor core by porous media approach

Abstract

Super critical water reactors (SCWR) are high efficient version of conventional pressurized water reactor with the benefit of using super critical water as coolant and moderator. Researches and studies on nuclear reactor core are usually subdivided into two major fields, named: thermal–hydraulic and neutronic, in which, precise simulation of reactor behavior in both fields is highly required to ensure the designers that reactor will work in a safe margin. In this study, a thermal–hydraulic analysis of a typical SCWR with three pass core design is performed using a porous media approach. Based on this approach, each fuel assembly was modeled and was divided into a network of lumped regions; each of them was characterized by a volume average parameter. In such manner, while complex geometries are easily defined and dealt with, the thermal–hydraulic parameter and phenomena like friction, shear stress, cross-flows, convective heat transfer and etc. are strictly included in simulations. To validate the applied approach, the numerical analysis and available data in literature were compared and showed a good agreement with those reported by other authors.