Home / Store / Journals / Electronic Articles / Nuclear Technology / Volume 75 / Number 2
B. S. Pei, Y. B. Chen, Chunkuan Shih, W. S. Lin
Volume 75 / Number 2 / November 1986 / Pages 134-147
Format:electronic copy (download)
The performance of five critical heat flux (CHF) correlations with the COBRA IIIC/MIT-1 code was evaluated. These correlations were evaluated against a data group comprised of 2943 axial nonuniform, rod bundle, first-order, and higher rank CHF data points of pressurized water reactor (PWR) core geometries. Among these five CHF correlations, EPRI-1 is the most accurate and has the widest operating ranges. Two kinds of correction factors—cold-wall correction factors and the CHF local quality correction factor—were developed and introduced to EPRI-1 to improve its accuracy in PWR CHF predictions. An in-depth evaluation of the EPRI-1 correlation in the prediction of CHFs of three fuel element abnormalities was also performed. Heat flux spikes and blocked channel conditions have negligible effects on CHFs. For the adverse effects of rod bowing on CHFs, the severity of rod bowing effects depends on the percentage of gap closure between rods, and also on the presence of any thimble tube (cold wall) adjacent to the distorted subchannel. Rod bowing effect parameter correlations under cold-wall conditions were developed. These rod bowing effect parameter correlations were tested; it was proved that they could closely describe the rod bowing effects with no apparent remaining residual trends.
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