Analysis of Heat Transfer Characteristics of Canadian SCWR Fuel Assembly Concept

Abstract

Abstract Canada is participating in the Generation IV (Gen IV) International Forum with a main focus on the pressure-tube-type supercritical water-cooled reactor (SCWR) concept. The subchannel code ASSERT-PV modified for SCWR applications was used to design the SCWR fuel assembly, specifically the fuel bundle. Several assumptions were required to model the fuel assembly, including the perfect insulation of (i) the central flow tube (i.e., no heat transfer through the central tube) and (ii) the pressure tube (i.e., no heat loss to the moderator). These two assumptions were considered as conservative, but they were not analyzed or assessed for their validity or accuracy. ASSERT-PV was upgraded to model the heat loss to the moderator, and an external CATHENA system code model was coupled to ASSERT-PV to model the heat transfer to the central flow tube. This paper describes these additional heat transfer components, and presents an assessment of these two assumptions for their impact on the prediction of maximum fuel cladding temperature.