Development and assessment of a new rod-bundle CHF correlation for China fuel assemblies

Abstract One of the main objectives of subchannel analysis for a pressurized water reactor (PWR) core or fuel assembly is to predict the critical heat flux (CHF) which usually employs the correlation based on the local parameters. The effects of spacer grid, cold wall, and non-uniform heat flux are essential for the sake of accurate calculation of local parameters. The subchannel code CORTH is applied to determine the local thermal-hydraulic parameters at the occurrence of departure from nucleate boiling (DNB) for each set of bundle CHF data of China fuel (CF) assemblies. In order to develop a new CHF correlation for CF assembly, the minimum DNB ratio (MDNBR) point and the burn-out (BO) point methods are both employed. The three-step form was adopted for the correlation which considers the effects of spacer grids, cold walls, and non-uniform heat flux. The coefficients of the correlation were determined based on the nonlinear regression method. The analysis and assessment results against the experimental data indicate that the correlation can predict the CHF with high prediction accuracy and correct parametric trends. In comparison with MDNBR point method, the BO point method is recommended to predict the risk of DNB because it is more reasonable and conservative and has a better predication rate. When implemented with the subchannel code CORTH, the developed correlation can predict the thermal-hydraulic performances of CF assemblies.