Two experiments were performed in the TREAT reactor using seven-rod bundles of 27-in.-long pressurized Zircaloy-clad UO2 fuel rods to determine fuel rod failure characteristics under water reactor loss-of-coolant accident (LOCA) conditions. Fissioning in the UO2 pellets provided the most realistic duplication available of heat transfer from stored energy and decay heat expected in a reactor LOCA. The center rod of each experiment was previously irradiated in the ETR and cladding temperatures of 1800 and 2400°F were reached in a flowing steam atmosphere in the two TREAT experiments. Maximum cladding expansion averaged 36 and 60% in the two experiments with ruptures occurring over a 2¼-in. axial length. The rate of volume expansion from clad swelling was calculated and the onset of rapid expansion correlated well with the ultimate stress. Average coolant channel blockage at the worst axial location was 48% in the first experiment and 91% in the second experiment. Fission product release was <0.5%, and the release of some fission products was inhibited by the smaller rupture opening in the second experiment.