Abstract
A prismatic gas-cooled very high temperature reactor (VHTR) is being developed under the next generation nuclear plant program (NGNP) for the United States Department of Energy. It is of interest to know the flow and temperature distributions in the core during normal operation and, especially for the thermal hydraulic phenomena in the reactor core during particular accident scenarios. The accident analysis is initiated from normal operating conditions. The accuracy of the simulation for the normal operating condition depends on many design considerations for the reactor core. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow, which occurs in the interstitial regions between fuel blocks. Present study begins with the investigation of core bypass flow phenomena for normal operating conditions, and then investigates a loss of flow accident (LOFA). The LOFA occurs when the coolant circulators are lost for some reason, causing a loss of forced convection through the core. One of the mechanisms that may occur after a LOFA for the transport of heat out of the core is by the natural convection of the coolant. It is also the objective of the present work to characterize the phenomenon of natural convection after a LOFA.Computational fluid dynamics (CFD) codes, which have simulation capabilities based on the physics of fluid flow and heat transfer, are widely used in various industrial fields. This study investigates core thermal hydraulic phenomena with the assistance of commercial CFD codes. However, the accuracy of CFD computations is affected by several parameters as well as the calculation domain, which is dependent on the capacity of computer. Present study begins with the smallest model, a 2D single channel, and then expands the CFD model gradually to partially validate and investigate the thermal hydraulic phenomena for the reactor core. Therefore, confidence in the accuracy of the biggest model involving the 1/12 core section is increased by having performed several pilot investigations.Turbulence models that perform well are then used to make steady bypass flow calculations in a symmetric one-twelfth sector of a prismatic block that includes bypass flow. Increasing surface roughness increases the maximum fuel and helium temperatures as do increases in gap width. However, maximum coolant temperature variation due to increased gap width is not changed by surface roughness. Partial validations and recommendations on turbulence model selection are also conducted in LOFA transient computations. Moreover, the present study has found that it is necessary to employ representative geometries of the core to estimate the heat transfer. By taking advantage of global and local symmetries, a detailed estimate of the strength of the resulting natural circulation and the level of heat transfer is obtained in the two sub-region model for limited capacity of computer. The effects of bypass flow after the LOFA are no longer as significant as for normal operation. Many detailed and accurate results are obtained from the 1/12 core model to characterize the phenomena occurring during the natural convection after a LOFA. These situations suggest that a larger computational domain might need to be employed for a LOFA transient condition.