Heat transfer to a supercritical pressure fluid flowing in sub-bundle channels

Abstract

Supercritical pressure water cooled reactor (SCWR) has been regarded as an innovative nuclear reactor. For the design and development of the SCWR, heat transfer performance under supercritical pressure is one of the most important indicators. In this paper, experimental data are presented on the heat transfer to a supercritical pressure fluid flowing vertically upward and downward in a small diameter heated tube and two sub-bundle channels with three heater rods and seven heater rods, using HCFC22 as the test fluid. Downstream of grid spacer for the sub-bundles, heat transfer enhancement was observed in the upward flow, but not in the downward flow. The enhancement was remarkable especially when the heat transfer deterioration occurs in the fully developed region unaffected by the spacer. The heat transfer correlation for the downstream region of the spacer was developed in the normal heat transfer of subbundles. In the fully developed region for the sub-bundle, occurrence of the heat transfer deterioration was suppressed or degree of the deterioration was moderated. At high mass velocity for downward flow in the seven rod sub-bundle, oscillation of heat transfer was observed in the region of the enthalpy over the pseudocritical point.

Experimental study on heat transfer to supercritical water in 2×2 rod bundle with wire wraps

“…They assessed eight selected correlations against the experimental data, and found that the correlations of Jackson [16,17] and Ornatsky et al [2] provided the best prediction accuracies. Mori et al [12] performed experiments on heat transfer of supercritical Freon R22 flowing vertically upward and downward in 3-rod and 7-rod bundles. They found that heat transfer was enhanced downstream of the grid spacer in the upward flow, but not in the downward flow.…”

Supercritical-Pressure Light Water Cooled Reactors

“…In association with development of the Super FR and Super LWR, many experiments have been performed using hydrochlorofluorocarbons as the surrogate fluid for water (Yamashita et al 2003;Mori et al 2005Mori et al , 2008Mori et al , 2009aMori et al , b, 2012Kaida et al 2011;Kariya et al 2013;Oka et al 2011Oka et al , 2013. Results of three types of unit assembly cell calculations performed independently of the specific core layout were incorporated with results of core analysis, and then the effects of neutronic heterogeneity of surrounding FAs were taken into account in the cell homogenized macroscopic cross sections.…”

“…In the experiments on heat transfer and pressure drop, various types of test sections with different channel geometries were used: a single tube, a single rod, a three-rod bundle, and a seven-rod bundle (Yamashita et al 2003;Mori et al 2005Mori et al , 2008Mori et al , 2012Kaida et al 2011). Mainly, the heat transfer coefficient and pressure drop were measured under steady-state and transient conditions.…”

“…To clarify basic thermal hydraulic characteristics in small hydraulic diameter tubes, heat transfer and pressure drop data have been obtained for supercritical pressure fluid flowing vertically upward and downward in a circular tube test section of small inside diameter (Mori et al , 2008(Mori et al , 2012Kaida et al 2011). Chlorodifluoromethane, HCFC22, was used as the surrogate fluid for water.…”

General Assessment of Convection Heat Transfer Correlations for Multiple Geometries and Fluids at Supercritical Pressure

“…Unfortunately, plain rod bundle data were unavailable for comparison. Finally, Figures 8 and 9 show the heat transfer profiles of the leading correlations against the Freon data by Richards [15] for R-12 in a 7-rod bundle and Mori et al [47] for R-22 in a 3-rod bundle, respectively. All selected leading correlations showed reasonable representation of the Freon data for both rod bundle geometries.…”