Supercritical water-cooled reactor (SCWR) is one of the Gen-IV reactors, which shows higher economy and safety. Reactor core of a SCWR employs a tight lattice in order to efficiently remove heat from nuclear fuels. In the narrow sub-channels of a tight lattice reactor core, a spacer has been used as a turbulence generator and a space-keeper between the fuel rods. Meanwhile, the spacer has significant influences on the heat transfer of the fuel rods. In order to investigate the effects of simple spacers on heat transfer to upwardly flowing supercritical fluid in a vertical annular channel, an experiment is underway at the supercritical model fluid thermal hydraulics test facility (SMOTH) with supercritical R134A. The equivalent diameter of the annular channel is 6 mm. The outer metal circular tube of the annular channel test section is electrically-heated. The blockage ratio of the simple spacers ranges from 0.2 to 0.4. Based on the geometry parameters of the test section, preliminary numerical investigations were carried out for the effects of simple spacer on the local heat transfer performance of supercritical R134A using commercial CFD code FLUENT. Heat transfer characteristics in the spacer downstream were analyzed with respect to the variations of heat flux, mass flux, pressure, blockage ratio and local enthalpy. And the reason for the different heat transfer enhancement under different conditions is given preliminarily. Finally, existing empirical correlations were selected to be compared with the results of CFD numerical simulation. The applicability of conventional subcritical heat transfer enhancement correlations for spacer grids to supercritical fluid was discussed.
- Nuclear Engineering Division
Numerical Investigation of Spacer Effects on Heat Transfer Enhancement in a Vertical Annular Channel at Supercritical Pressure
Pan, J, Xiao, Y, & Gu, H. "Numerical Investigation of Spacer Effects on Heat Transfer Enhancement in a Vertical Annular Channel at Supercritical Pressure." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 8: Computational Fluid Dynamics (CFD) and Coupled Codes; Nuclear Education, Public Acceptance and Related Issues. Shanghai, China. July 2–6, 2017. V008T09A030. ASME. https://doi.org/10.1115/ICONE25-66911
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