Critical heat flux and heat transfer transition for subcooled flow boiling

Journal Article · Fri Feb 01 00:00:00 EST 1991 · Journal of Heat Transfer (Transcations of the ASME (American Society of Mechanical Engineers), Series C); (United States)

DOI:https://doi.org/10.1115/1.2910542· OSTI ID:5510683

Boyd, R D ^[1]

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1. Prairie View A and M Univ., TX (USA)

The emphasis in the engineering development of fusion reactor components has been on material development. If high heat fluxes are to be accommodated with the present emphasis, low-pressure thermal data will be needed. The objectives of this experiment were to (1) expand the critical heat flux, W/cm{sup 2} data base near 4.0 kilowatts cm{sup 2} and heated coolant channel length divided by coolant chamber diameter near 100.0 (near-term application), (2) add low-pressure quantitative data to our existing knowledge of the qualitative influence of coolant exit pressure on CHF, and (3) provide thermal data in a region applicable to high heat flux components for assessing existing and evolving CHF and local heat transfer coefficient correlations (long-term).

• OSTI ID: 5510683
• Journal Information: Journal of Heat Transfer (Transcations of the ASME (American Society of Mechanical Engineers), Series C); (United States), Vol. 113:1; ISSN 0022-1481
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
THERMONUCLEAR REACTORS
SUBCOOLED BOILING
CRITICAL HEAT FLUX
LOW PRESSURE
MULTIPHASE FLOW
TRANSITION HEAT
BOILING
ENTHALPY
FLUID FLOW
HEAT FLUX
PHASE TRANSFORMATIONS
PHYSICAL PROPERTIES
THERMODYNAMIC PROPERTIES
700204* - Fusion Power Plant Technology- Cooling Systems