Abstract

The use of porous media to simplify the thermohydraulic of a nuclear reactor is the topic of recent research. As a case study, the rector of 200kW installed at Missouri University of Science and Technology is modeled in this paper. To help this objective, a fundamental CFD examination was completed to supplement the neutronics investigation on the present reactor. Characteristics of thermal energy removal from a typical research reactor are modeled by numerical thermal transport in porous media. The neutron flux is modeled by the nodal expansion method. For the thermo-hydraulic part, a three-dimensional governing equation is solved by an iterative method to find the steady-state solution of fluid flow and temperature in loss of coolant condition where the heat produced in the reactor core is removed by free convection. The profiles of heat flux for various power levels are benchmarked. Pressure, temperature and velocity contours in the power passage were assessed at 300kW and 500kW power levels. To reduce the computational cost, a porous media approach for the whole geometry was utilized. The numerical results agree with the experimental results. The developed model can be used for safety and reliability analysis for various loss of coolant accidents.

Keywords: Natural convection; Porous media; Two-phase flow; Loss of coolant; Nodal expansion method

Nomenclature: Cp: Specific Heat; k: Thermal Conductivity; K: Permeability of the Porous Medium; q: Heat Flux; Da: Darcy Number; Nu: Nusselt Number; Pr: Prandtl Number u; v: Velocity Vector; P: Pressure; T: Temperature; x, y: Cartesian Coordinates; X, Y: Dimensionless Cartesian Coordinates