Abstract
Over the last decade, several international thermalhydraulics benchmarking efforts have been carried out to support the development of the Generation IV supercritical water-cooled reactor (SCWR) concept. These benchmarking efforts aimed to assess the readiness of computer codes to predict the thermalhydraulics behavior of supercritical fluids for nuclear fuel assembly applications. The results from the benchmarking also shed light on knowledge gaps. Throughout the years, several advancements in this area have been achieved, resulting in relevant conclusions and observations. Furthermore, experimental campaigns have been carried out worldwide to further our knowledge on the thermalhydraulics of supercritical fluids. The nuclear industry uses the subchannel approach to study the thermalhydraulics behavior of nuclear fuel assemblies in detail. In Canada, the subchannel code advanced solution of subchannel equations in reactor thermalhydraulics—pressure velocity (ASSERT-PV) is the qualified code for subchannel applications. ASSERT-PV was modified to handle supercritical conditions, resulting in an interim code version. This publication presents relevant subchannel analyses using the interim supercritical version of ASSERT-PV for fuel assemblies cooled with supercritical fluids.
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