Abstract
Correlation for rod-bundle critical heat flux (CHF) is of vital importance for design and assessment of a Pressurized Water Reactor (PWR). In our previous study, the basic form of a dimensionless rod-bundle CHF correlation has been developed based on stepwise regression with CHF data obtained in test bundles of uniform axial heat flux. It proves that stepwise regression is an effective method to develop simple, but reliable CHF correlation consisting of only dimensionless parameters with rigorous physical meanings. As a continuation of the previous investigation, revision of the basic form was conducted. Dimensionless correction factors accounting for the cold-wall effect due to unheated guide tube and the effect of non-uniform axial heat flux were developed. For the data set of 486 rod-bundle CHF points obtained in 7 test bundles, the average ratio of the measured to predicted CHF is 1.002 with a standard deviation from the mean of 5.51%.
Sign-in/Register to access full text options 
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a callDisclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.
