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.