Critical Heat Flux in Vertical and Horizontal 37-Element Fuel Bundles

Abstract

Designers and analysts must be able to predict critical heat flux (CHF) in the 37-element fuel bundles used in CANDU reactors under various normal and abnormal operating conditions. Because of the difficulties of predicting flow and enthalpy distributions among the sub-cahnnels of fuel bundles and the complexity of the mechanisms governing CHF under these conditions, reliable analytical models do not exist for this purpose and designers and analysts must rely on correlations of relevant experimental data. At present, a conservative lower-bound envelope of early experimental data obtained in the vertical, in-pile U-1 loop at Chalk River Nuclear Laboratories of AECL is used as a design correlation (1).