EXPERIMENTAL STUDIES ON CRITICAL HEAT FLUX IN TIGHT LATTICE ROD BUNDLES

The tight fuel lattice of pressurized water reactors(PWR) is helped to reduce the volume ratio of water-uranium, to increase the conversion ratio, to decrease the volume of core. It is especially useful for very high burnup and high volume power flux. The design of tight-lattice pressurized water reactors requires the knowledge of critical heat flux (CHF) in tight rod bundles. The tight hexagonal 19-rod bundles is used in this test. There are 4 wires wrapped in outside wall of each rod to support and locate. Experimental investigations on CHF behavior in the two kind bundles of helix angle 3 degrees and 5 degrees were performed. The CHF data points have been obtained in a range of parameters: pressure 8.0-16.6 MPa, mass flux 164.6-3283.0 kg/m(2)s and bundle exit steam quality -0.315 to 0.747. It is found that the CHF value of helix angle 5 degrees bundle was more higher than that of helix angle 3 degrees bundle in the same T/H condition. The effect of different parameters on CHF in the tight rod bundle is similar to that in the open literature. The CHF correlations of helix angle 50 bundle was obtained based on the test data.