The Thermalhydraulic Characteristics of the CANDU-6 Reactor Channel with a CANFLEX-RU Fuel Bundle

Abstract

This paper describes the thermalhydraulic characteristics of a CANDU-6 reactor's fuel channel loaded with CANFLEX-RU (CANDU Flexible Fuelling - Recycled Uranium) bundles. The distributions of the channel flow rate, channel exit quality, critical channel power(CCP) and critical power ratio(CPR) for the CANFLEX-RU fuel channels are evaluated by using the NUCIRC code. This code is incorporated with the recent models for pressure drop and critical heat flux(CHF) of the CANFLEX fuel bundle as well as a 37-element bundle. Especially, the effects of pressure tube creep and bearing pad height on the thermalhydraulic characteristics are also examined by comparing the various results on the uncrept, 3.3% and 5.1 % crept channels loaded with CANFLEX bundles with the 1.4 mm height or 1.7 mm height bearing pads with those of the 37-element bundle. The distributions of channel flow rate and CCP for the CANFLEX-RU bundle show a typical trend for the CANDU-6 reactor channel, and the CPR is maintained above 1 .455 at least in the uncrept channel. The CANFLEX-RU fuel bundle is considerably less sensitive to the CCP reduction due to pressure tube creep than the 37-element bundle. The CCP enhancement of the CANFLEX-RU fuel bundle due to the raised bearing pads is estimated to be about 2 % - 6%.