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A Theoretical Critical Heat Flux Model for Low-Pressure, Low-Mass-Flux, and Low-Steam-Quality Conditions

Wei-Hsiao Ho, Kuan-Chywan Tu, Bau-Shei Pei, Chin-Jang Chang

Nuclear Technology

Volume 103 / Number 3 / September 1993 / Pages 332-345


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The critical heat flux (CHF) is the maximum heat flux just before a boiling crisis; its importance as a measurement of nuclear reactor power capability design as well as in the safety analysis of reactors has been recognized. With emphasis on CHF behavior under subcooled and low-quality (i.e., <0.25) convective flow at low pressure (i.e., <4.9 MPa) and low mass flux (i.e., <1000 kg/m2·s), an improved model that uses the sublayer dryout theory has been developed. Based on experimental observations of CHF, the model assumes that CHF under such conditions is of the departure from nucleate boiling type. Based on the postulation that CHF is triggered by Helmholtz instability in the sublayer stem-liquid system, the model was developed by a simple energy balance of liquid sublayer evaporation as the vapor blanket tends to disturb the balance between the buoyance force and the drag force exerted upon it. The model is compared with the well-known Biasi et al. correlation as well as the Atomic Energy of Canada Limited lookup table against 102 uniformly heated round tube CHF data and 34 nonuniformly heated round tube CHF data. The comparison shows that the model provides better accuracy and a reasonable agreement between the predicted values and experimental CHF data.

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