CFD Analysis and Assessment for Cross-Flow Phenomena in VHTR Prismatic Core

The very high temperature gas-cooled reactor (VHTR) is a uranium-fueled, graphite-moderated, and helium-cooled reactor envisioned as one of the promising future nuclear reactor concepts due to its high efficiency, safety, and variety of applications. In this reactor concept, core bypass flow and cross flow are currently considered as key issues because of their inherent uncertainties and complication of prediction. Recently, the computational fluid dynamics (CFD) method has received a great deal of attention as a method for understanding the flow behavior in the VHTR core, including the bypass and the cross flow. However, validation of this method has not been sufficiently conducted yet based on real experimental data. For this reason, prediction capability of the CFD method was validated in this study by comparing the predictions with the existing multi-hole experimental data obtained by Groehn as a part of the core thermal-hydraulics design study for the NHDD PMR-200. A two-stacked fuel block with wedge-shaped cross gap was simulated for computational domain as the experimental setup. The flow loss coefficients and the velocity distributions of the cross flow from the experiment were compared to the CFD predictions extensively. As a result, good agreements between the CFD predictions and the experimental results were observed, confirming prediction capability of the CFD method for the complicated cross flow in the VHTR core. Furthermore, the velocity distributions and pressure distribution in the cross gap were investigated to identify the characteristics of the cross flow.