During high-temperature transients on Zircaloy-clad nuclear fuel rods, both external and internal cladding oxidation have been observed. The result of such oxidation is to embrittle the cladding, making it susceptible to brittle fracture. External oxidation of Zircaloy in a steam environment is relatively well understood, while the internal processes have received little attention. This paper examines the internal oxidation process that occurs as a result of oxygen transport from the uranium oxide fuel pellets to the Zircaloy cladding. An oxygen diffusion model has been derived that solves a coupled two-medium problem with a moving boundary in Zircaloy due to the α-β phase change. Analytical results for the time and temperature dependence of such Zircaloy oxidation were found to compare favorably with both in- and out-of-pile experimental data, indicating that the underlying oxidation process is diffusion controlled.