Convective heat transfer to CO2 at a supercritical pressure flowing vertically upward in tubes and an annular channel

The Super-Critical Water-Cooled Reactor (SCWR) has been chosen by the Generation IV International Forum as one of the candidates for the next generation nuclear reactors. Heat transfer to water from a fuel assembly may deteriorate at certain supercritical pressure flow conditions and its estimation at degraded conditions as well as in normal conditions is very important to the design of a safe and reliable reactor core. Extensive experiments on a heat transfer to a vertically upward flowing CO2 at a supercritical pressure in tubes and an annular channel have been performed. The geometries of the test sections include tubes of an internal diameter (ID) of 4.4 and 9.0 mm and an annular channel (8 × 10 mm). The heat transfer coefficient (HTC) and Nusselt numbers were derived from the inner wall temperature converted by using the outer wall temperature measured by adhesive K-type thermocouples and a direct (tube) or indirect (annular channel) electric heating power. From the test results, a correlation, which covers both a deteriorated and a normal heat transfer regime, was developed. The developed correlation takes different forms in each interval divided by the value of parameter Bu. The parameter Bu (referred to as Bu hereafter), a function of the Grashof number, the Reynolds number and the Prandtl number, was introduced since it is known to be a controlling factor for the occurrence of a heat transfer deterioration due to a buoyancy effect. The developed correlation predicted the HTCs for water and HCFC-22 fairly well.