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International Heat Transfer Conference 7

ISSN: 2377-424X (online)
ISSN: 2377-4371 (flashdrive)

PREDICTION OF CRITICAL HEAT FLUX IN PWR FUEL ASSEMBLIES WITH NON-UNIFORM AXIAL HEAT FLUX DISTRIBUTION

D. G. Reddy
Columbia University, New York, NY 10027

C. Fighetti
Columbia University, New York, NY

M. Merilo
Nuclear Safety and Analysis Department, Electric Power Research Institute, 3412 Hillview Avenue, Palo Alto, California 94303, USA

Abstract

The performance of the Columbia subchannel CHF correlation [1, 2] in predicting CHF limits in fuel assemblies with non-uniform axial heat flux distribution was evaluated. A total of 933 CHF data points from 23 test sections simulating PWR fuel assemblies with eight axial heat flux distributions were utilized in this study. The experimental verification was performed by comparing the measured local heat flux at the location of CHF occurrence with the predicted CHF at that location, calculated using the local conditions obtained with the COBRA 111C subchannel code for the test inlet conditions and the measured bundle power. The correlation predicted the non-uniform axial heat flux distribution CHF data with an average error of 0.02% and standard deviation of 8.55%. The performance of the correlation improved significantly when a correction factor to account for the upstream heat flux effects was introduced. The modified correlation predicted the data with an average error of 0.13% and standard deviation of 6.92%.

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