Microscopic examinations of advanced zirconium alloy cladding irradiated to burnups over 70 GWd/MTU have revealed de-lamination of surface layers of the thin oxide. The new observation is termed “oxide surface peeling” or OSP. Examinations have revealed the layered structure of the oxide. Metallographic examination revealed that the waterside oxide on different irradiated Zr alloy cladding had a layered structure similar to the autoclaved corrosion specimens examined earlier. However, the OSP observations discussed here apply only to irradiated cladding. OSP is not observed in autoclave corrosion. A featureless intact oxide sub-layer was present in the interior oxide at the metal/oxide interface for all alloys. On top of this featureless layer, there were additional sub-layers with fine circumferential fissures believed to be associated with the cyclic corrosion rate transitions. The number of sub-layers depended on the corrosion resistance of the alloy. Intermediate corrosion resistance alloy had many sub-layers forming an overall intermediate thickness oxide. Higher corrosion resistance alloys had fewer sub-layers on top of the barrier oxide layer. In some cases, small patches of the top surface layers of the thin oxide were peeled off because of radial cracks generated in the oxide layer caused by tensile stresses created by a hard pellet contact. Metallography of the underlying cladding showed that hydride localization was not associated with the oxide surface peeling; in contrast to previous experience on the low corrosion resistance older claddings, such as Zircaloy-4, where a major thickness fraction of the thick oxide extending to the underlying metal was removed (“spalled”). The oxide surface layer peeling does not lead to irradiated alloy ductility reduction or enhanced embrittlement. The impact of oxide surface peeling observations on fuel performance is discussed in the paper. For fuel designs with low margin against grid-to-rod fretting wear, OSP may reduce this margin further.