Abstract
Accurate prediction of rod-bundle critical heat flux (CHF) is a main challenge in thermal-hydraulic design and safety assessment of Pressurized Water Reactor (PWR). The most widely applied approach is empirical correlation developed based on rod-bundle experiments carefully tested with geometry and power distribution representative to PWR working conditions. The standard procedure is to correlate the experimentally obtained CHF values as empirical functions of local thermal-hydraulic conditions obtained with subchannel analysis. However, up-to-date empirical correlations developed according to this manner are in general quite complicated with a number of dimensional coefficients and correction factors with less or non-physical meanings. In this study, stepwise regression method was used to develop a novel, dimensionless rod-bundle CHF correlation covering typical PWR working conditions. First, various dimensionless parameters affecting CHF were selected as candidate independent variables. With stepwise regression, the form and coefficients of the proposed CHF correlation were dynamically optimized and determined.
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