Many critical heat flux (CHF) correlations have been developed for water-cooled rod clusters representing typical pressurized water reactor (PWR) or boiling water reactor fuel element geometries with relatively wide rod lattices. The fuel elements of an advanced pressurized water reactor (APWR), however, have a tight fuel rod lattice in order to increase fuel utilization. It was therefore decided to produce a new CHF correlation valid for rod bundles with tight lattices. The already available WSC-2 correlation was chosen as a basis. The geometry-dependent parameters of this correlation were again determined by root-mean-square fitting from the experimental data of CHF test performed within the framework of the light water breeder reactor program at Bettis Atomic Power Laboratory. These tests include triangular arrays of rod bundles with very tight lattices. The effects of spiral spacer ribs were based on experimental data from Columbia University. The present correlation was compared with various tests performed with rod bundles with wire wrapped spacers. Application of the new CHF correlation to conditions typical for an APWR shows that the predicted CHFs are smaller than those calculated with the usual PWR CHF correlations.