Computational fluid dynamics analysis of core bypass flow and crossflow in a prismatic very high temperature gas-cooled nuclear reactor based on a two-layer block model

•A CFD model was built based on a two-layer block experimental facility at Texas A&M University.•The coolant characterizations within the uniform and wedge-shaped crossflow gap regions were investigated.•The influence on the coolant distribution from the bypass flow gap width was studied.•Discretization and iterative errors involved in the simulations were quantified.

The very high temperature gas-cooled nuclear reactor (VHTR) has been designated as one of the promising reactors that will serve for the Next Generation (Generation IV) Nuclear Plant. For a prismatic VHTR core, the bypass flow and crossflow phenomena are important design considerations. To investigate the coolant distribution in the reactor core based on the two-layer block facility built at Texas A&M University, a three-dimensional steady-state CFD analysis was performed using the commercial code STAR-CCM+ v6.04. Results from this work serve as a guideline and validating source for the related experiments. A grid independence study was conducted to quantify related errors in the simulations. The simulation results show that the bypass flow fraction was not a strong function of the Reynolds number. The presence of the crossflow gap had a significant effect on the distribution of the coolant in the core. Uniform and wedge-shape crossflow gaps were studied. It was found that a significant secondary flow in the crossflow gap region moved from the bypass flow gap toward coolant holes, which resulted in up to a 28% reduction of the coolant mass flow rate in the bypass flow gap.