A prismatic gas-cooled very high temperature reactor (VHTR) is being developed under the next generation nuclear plant program (NGNP) of the U.S. Department of Energy, Office of Nuclear Energy. In the design of the prismatic VHTR, hexagonal shaped graphite blocks are drilled to allow insertion of fuel pins, made of compacted tristructural-isotropic (TRISO) fuel particles, and coolant channels for the helium coolant. One of the concerns for the reactor design is the effects of a loss of flow accident (LOFA) where the coolant circulators are lost for some reason, causing a loss of forced coolant flow through the core. In such an event, it is desired to know what happens to the (reduced) heat still being generated in the core and if it represents a problem for the fuel compacts, the graphite core or the reactor vessel (RV) walls. One of the mechanisms for the transport of heat out of the core is by the natural circulation of the coolant, which is still present. It is desired to know how much heat may be transported by natural circulation through the core and upwards to the top of the upper plenum.
It is beyond current capability for a computational fluid dynamics (CFD) analysis to perform a calculation on the whole RV with a sufficiently refined mesh to examine the full potential of natural circulation in the vessel. The present paper reports the investigation of several strategies to model the flow and heat transfer in the RV. It is found that it is necessary to employ representative geometries of the core to estimate the heat transfer. However, 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 to the top of the upper plenum is obtained.
Modeling Strategies to Compute Natural Circulation Using CFD in a VHTR After a LOFA