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Title: Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

Technical Report ·
DOI:https://doi.org/10.2172/115071· OSTI ID:115071
; ;  [1]
  1. Rensselaer Polytechnic Institute, Troy, NY (United States)

This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

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:
115071
Report Number(s):
NUREG/CP-0142-Vol.3; CONF-950904-Vol.3; ON: TI95017079; TRN: 95:022971
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, Volume 3, Sessions 12-16; Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)]; PB: 1001 p.
Country of Publication:
United States
Language:
English