The fuel oxidation of UO 2 pellets in two types of defective Zircaloy-clad BWR fuel rods with small leaks has been examined along both pellet diametral and fuel rod axial directions. The post-defect irradiation time was a few months for the base-irradiated full length rod and several minutes for the power-ramped segment rod. No phase change to higher order oxides of U 40 9 or U 3O 8 was found, but hyperstoichiometric UO 2+ x with fluorite structure was still present for both fuels. The fuel oxidation significantly depended on the defect size and distance from the defect. The pellet volume-averaged O/M ratios at various axial locations were in the range of 2.02–2.06 for the base-irradiated fuel, and about 2.01 for the power-ramped fuel. The data revealed that pellet oxidation by steam proceeded notably even in a short period of several minutes and played a more important role for generating liberated hydrogen, which could cause secondary hydriding of Zircaloy cladding, in comparison with the inner wall oxidation of cladding.