Transient pool boiling heat transfer due to increasing heat inputs in subcooled water at high pressures
- Kobe Univ. of Mercantile Marine (Japan)
- Kyoto Univ. (Japan)
Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q{sub max}, on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q{sub o}e{sup t/T}, with periods, {tau}, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q{sub max}. Two main mechanisms of q{sub max} exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q{sub max} for long period range belonging to the former mechanism becomes longer and the q{sub max}mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q{sub max} for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling.
- 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); Japan Society of Multiphase Flow, Kyoto (Japan)
- OSTI ID:
- 107012
- Report Number(s):
- NUREG/CP-0142-Vol.1; CONF-950904-Vol.1; ON: TI95017077; TRN: 95:020901
- 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 1, Sessions 1-5; Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)]; PB: 862 p.
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING NOT INCLUDED IN OTHER CATEGORIES
WATER COOLED REACTORS
SAFETY ANALYSIS
TRANSIENT OVERPOWER ACCIDENTS
POOL BOILING
HEAT TRANSFER
HYDRAULICS
NUCLEATE BOILING
DIAGRAMS
EXPERIMENTAL DATA
SUBCOOLING
REACTOR SAFETY