Correlation of critical heat flux data for application to boiling water reactor conditions. Final report

A study was conducted of publicly available multirod critical heat flux (CHF) data to generate a correlation which may be used to predict the limits of nucleate boiling in a boiling water reactor. A critical quality-boiling length correlation was generated which includes the effects of rod-to-rod (local) peaking factor, different axial heat flux profiles, pressure, rod diameter and bundle geometry. The CHF data base for this correlation is primarily 16-rod, with heated lengths up to 12 feet, with axially uniform and non-uniform heat flux, in a pressure range from 600 to 1400 psia, and a mass velocity range from 0.25 to 1.5 Mlb/h ft/sup 2/. The correlation predicts the critical power of the test data used to create the correlation with a standard deviation of 5.5%, and a mean value of 0.9948. When all applicable BWR CHF data are compared to the correlation, it predicts the test bundle critical power with a standard deviation of 7.5%, with a mean value of 1.0295.