Void fraction distribution in a heated rod bundle under flow stagnation conditions

Conference · Fri Sep 01 00:00:00 EDT 1995

OSTI ID:111429

Herrero, V A; Guido-Lavalle, G; Clausse, A ^[1]

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1. Centro Atomico Bariloche and Instituto Balseiro, Bariloche (Argentina)

An experimental study was performed to determine the axial void fraction distribution along a heated rod bundle under flow stagnation conditions. The development of the flow pattern was investigated for different heat flow rates. It was found that in general the void fraction is overestimated by the Zuber & Findlay model while the Chexal-Lellouche correlation produces a better prediction.

• Research Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Systems Technology; American Nuclear Society (ANS), La Grange Park, IL (United States); American Institute of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers (ASME), New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); European Nuclear Society (ENS), Bern (Switzerland); Atomic Energy Society of Japan, Tokyo (Japan); Japan Society of Multiphase Flow, Kyoto (Japan)
• OSTI ID: 111429
• Report Number(s): NUREG/CP-0142-Vol.4; CONF-950904-Vol.4; ON: TI95017080; TRN: 95:021617
• Resource Relation: Conference: 7. international topical meeting on nuclear reactor thermal-hydraulics (Nureth-7), Saratoga Springs, NY (United States), 10-15 Sep 1995; Other Information: PBD: Sep 1995; Related Information: Is Part Of Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24; Block, R.C.; Feiner, F. [American Nuclear Society, La Grange Park, IL (United States)]; PB: 825 p.
• Country of Publication: United States
• Language: English

Related Subjects

22 NUCLEAR REACTOR TECHNOLOGY
FUEL CHANNELS
VOID FRACTION
WATER COOLED REACTORS
HEAT TRANSFER
HYDRAULICS
FUEL ELEMENT CLUSTERS
LOSS OF FLOW
TWO-PHASE FLOW
REACTOR SAFETY
TESTING