The results of heat transfer to supercritical water flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of the tubes of 5.2-mm outside diameter and 485-mm heated length are presented. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within the range from 800 to 3000 kg/m2·s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2 (heat flux rate up to 2.5 kJ/kg). Temperature regimes of the annular channel and 3-rod bundle were stable and easily reproducible within the whole range of the mass and heat fluxes, even when a deteriorated heat transfer took place. The data resulted from the study could be applicable for a reference estimation of heat transfer in future designs of fuel bundles.
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17th International Conference on Nuclear Engineering
July 12–16, 2009
Brussels, Belgium
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4353-6
PROCEEDINGS PAPER
Heat Transfer to Supercritical Water in Vertical Annular Channel and 3-Rod Bundle
V. G. Razumovskiy,
V. G. Razumovskiy
National Technological University of Ukraine “KPI”, Kyiv, Ukraine
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Eu. N. Pis’mennyy,
Eu. N. Pis’mennyy
National Technological University of Ukraine “KPI”, Kyiv, Ukraine
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A. Eu. Koloskov,
A. Eu. Koloskov
National Technological University of Ukraine “KPI”, Kyiv, Ukraine
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I. L. Pioro
I. L. Pioro
University of Ontario Institute of Technology, Oshawa, ON, Canada
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V. G. Razumovskiy
National Technological University of Ukraine “KPI”, Kyiv, Ukraine
Eu. N. Pis’mennyy
National Technological University of Ukraine “KPI”, Kyiv, Ukraine
A. Eu. Koloskov
National Technological University of Ukraine “KPI”, Kyiv, Ukraine
I. L. Pioro
University of Ontario Institute of Technology, Oshawa, ON, Canada
Paper No:
ICONE17-75212, pp. 233-238; 6 pages
Published Online:
February 25, 2010
Citation
Razumovskiy, VG, Pis’mennyy, EN, Koloskov, AE, & Pioro, IL. "Heat Transfer to Supercritical Water in Vertical Annular Channel and 3-Rod Bundle." Proceedings of the 17th International Conference on Nuclear Engineering. Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance. Brussels, Belgium. July 12–16, 2009. pp. 233-238. ASME. https://doi.org/10.1115/ICONE17-75212
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