Experimental investigation of flow transient critical heat flux at light water reactor conditions

Journal Article · · International Communications in Heat and Mass Transfer; (United States)

OSTI ID:6408640

Lee, Chienhsiung ^[1]; Lin, Kuowei ^[2]

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1. Inst. of Nuclear Energy Research, Lung-Tan (Taiwan, Province of China). Thermohydraulic Lab.
2. National Central Univ., Chungli (Taiwan, Province of China). Dept. of Mechanical Engineering

An experimental investigation of flow transient critical heat flux (CHF) for water at pressures 6.9 and 15.5 MPa is conducted under the simulation of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) conditions. The influence of flow decay rates on local mass velocity and quality at CHF is explored to ensure the adequacy of local-condition approach in the CHF predictions. The assessment of transient upstream correction factor to modify the predictions of steady-state CHF correlations during the BWR flow transients are presented. A mechanistic CHF model based on the dryout of a sublayer and CHF correlations are evaluated in the predictions of PWR flow transient CHF.

• OSTI ID: 6408640
• Journal Information: International Communications in Heat and Mass Transfer; (United States), Vol. 20:4; ISSN 0735-1933
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
CRITICAL HEAT FLUX
TRANSIENTS
PWR TYPE REACTORS
DRYOUT
EXPERIMENTAL DATA
FUNCTIONAL MODELS
HEAT TRANSFER
DATA
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
HEAT FLUX
INFORMATION
NUMERICAL DATA
POWER REACTORS
REACTORS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100* - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled