Boiling Water Reactor Core Analysis by means of an Improved Porous Media Two-phase Flow Approach

TWOPORFLOW is a thermal-hydraulics simulation code currently under development at the Institute of Neutron Physics and Reactor Technology (INR) of the Karlsruhe Institute of Technology (KIT). It has the capability to simulate single- and two-phase flow in a structured or unstructured porous medium using a 3-D Cartesian geometry. TWOPORFLOW calculates the transient or steady state solution of the mass, momentum, and energy conservation equations for each fluid phase with a semi-implicit numerical procedure based on the implicit continuous Eulerian (ICE) method. TWOPORFLOW can simulate simple 1-D geometries like heated pipes, fuel assemblies resolving the sub-channel flow between rods or a whole nuclear core using a coarse mesh. Several closure correlations are implemented to model the heat transfer between solid and coolant, phase change, wall friction as well as liquid-vapor momentum coupling. These models have been not fully evaluated. Important models such as the turbulent mixing and void dispersion were missing. Consequently, the goal of this doctoral thesis is to extend and improve the simulating capability of TWOPORFLOW regarding the two-phase flow phenomena in Boiling Water Reactors (BWR) by implementing physical models e.g., turbulent mixing, void dispersion, and critical heat flux. In order to evaluate the quality of the implemented models and the overall prediction capability of TWOPORFLOW, experimental data from selected experiments are used to be compared with the simulation’s results. Finally, the enhanced prediction capability of TWOPORFLOW is applied to the simulation of two BWR reactor cores, namely the one of Oskarshamn-2 (Sweden) and Laguna Verde unit 1 (Mexico) for the first time showing promising results. Based on the performed investigations, possibilities for further development are identified in the areas of model development, numerical improvement, validation, code coupling, and parallelization