Prediction of Critical Heat Flux during Downflow in Fully Heated Vertical Channels

Author:

Shah Mirza M.1

Affiliation:

1. Engineering Research Associates, 10 Dahlia Lane, Redding, CT 06896, USA

Abstract

Boiling with downflow in vertical channels is involved in many applications such as boilers, nuclear reactors, chemical processing, etc. Accurate prediction of CHF (Critical Heat Flux) is important to ensure their safe design. While numerous experimental studies have been done on CHF during upflow and reliable methods for predicting it have been developed, there have been only a few experimental studies on CHF during downflow. Some researchers have reported no difference in CHF between up- and downflow, while some have reported that CHF in downflow is lower or higher than that in upflow. Only a few correlations have been published that are stated to be applicable to CHF during downflow. No comprehensive comparison of correlations with test data has been published. In the present research, literature on CHF during downflow in fully heated channels was reviewed. A database for CHF in downflow was compiled. The data included round tubes and rectangular channels, hydraulic diameters 2.4 mm to 15.9 mm, reduced pressure 0.0045 to 0.6251, flow rates from 15 to 21,761 kg/m2s, and several fluids with diverse properties (water, nitrogen, refrigerants). This database was compared to a number of correlations for upflow and downflow CHF. The results of this comparison are presented and discussed. Design recommendations are provided.

Publisher

MDPI AG

Reference28 articles.

1. Barnett, P.G. (1963). An Investigation into the Validity of Certain Hypotheses Implied by Various Burnout Correlations, United Kingdom Atomic Energy Authority. Rep. AEEW-R-214.

2. The upflow and downflow critical heat flux of water and freon in a vertical tube and its flow direction factor;Chen;At. Energy Sci. Technol.,1993

3. Gambill, W.R., and Bundy, R.D. (1961). HFIR Heat-Transfer Studies of Turbulent Water Flow in Thin Rectangular Channels, Oak Ridge National Laboratory for the US Atomic Energy Commission. ORNL-3079, TID-4500.

4. Papell, S.S., Simoneau, R.J., and Brown, D.D. (1966). Buoyancy Effects on Critical Heat Flux of Forced Convective Boiling in Vertical Flow, NASA. NASA Technical Note D-3672.

5. Papell, S.S. (1977). Combined Buoyancy and Flow Direction Effects on Saturated Boiling Critical-Heat Flux in Liquid Nitrogen, NASA. NASA TM X-68086.

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