Critical Heat Flux Measurements in a 16-Rod Simulation of a BWR Fuel Assembly

[+] Author and Article Information
S. Israel

United Nuclear Corporation, Elmsford, N. Y.

J. Casterline, B. Matzner

Department of Chemical Engineering, Columbia University, New York, N. Y.

J. Heat Transfer 91(3), 355-361 (Aug 01, 1969) (7 pages) doi:10.1115/1.3580174 History: Received July 31, 1968; Online August 25, 2011


Critical heat flux data were obtained for forced flow boiling in a 16-rod test section arranged in a square array. The tests were performed at 1000 psia and used a radial power distribution which represented the region about the hot corner in a BWR fuel assembly. The results are lower than data obtained in a 9-rod square array, having a uniform power distribution, based on the average bundle exit quality. These two sets of data are in fair agreement when compared on the basis of the highest subchannel exit quality. Comparisons of different sets of data show the effects of different rod spacers and bundle misalignment on the critical heat flux.

Copyright © 1969 by ASME
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