A formal approach for the prediction of the critical heat flux in subcooled water
Description
The critical heat flux (CHF) in subcooled water at high mass fluxes are not yet satisfactory correlated. For this scope a formal approach is here followed, which is based on an extension of the parameters and the correlation used for the dryout prediction for medium high quality mixtures. The obtained correlation, in spite of its simplicity and its explicit form, yields satisfactory predictions, also when applied to more conventional CHF data at low-medium mass fluxes and high pressures. Further improvements are possible, if a more complete data bank will be available. The main and general open item is the definition of a criterion, depending only on independent parameters, such as mass flux, pressure, inlet subcooling and geometry, to predict whether the heat transfer crisis will result as a DNB or a dryout phenomenon
Additional details
Publishing Information
- Imprint Title
- Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24
- Imprint Pagination
- 825 p.
- Journal Page Range
- p. 2506-2518.
- Report number
- NUREG/CP--0142-Vol.4
Conference
- Title
- 7. international meeting on nuclear reactor thermal hydraulics.
- Dates
- 10-15 Sep 1995.
- Place
- Saratoga, NY (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 27021133
- Subject category
- S70: PLASMA PHYSICS AND FUSION TECHNOLOGY; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- CRITICAL HEAT FLUX; DRYOUT; HEAT TRANSFER; HYDRAULICS; MASS TRANSFER; MATHEMATICAL MODELS; PWR TYPE REACTORS; REACTOR COOLING SYSTEMS; THERMONUCLEAR REACTOR COOLING SYSTEMS; THERMONUCLEAR REACTORS; VERIFICATION
- Descriptors DEC
- COOLING SYSTEMS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; HEAT FLUX; POWER REACTORS; REACTOR COMPONENTS; REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-950904--Vol.4.