[en] In nuclear plants, some accidental situations can result in air exposure of Pressurized Water Reactor (PWR) fuel assemblies: air ingress following a breach in the reactor vessel, de-flooding during handling, spent fuel storage pool de-flooding. Deprived of cooling source, the assemblies temperature raises and the fuel cladding, made out of zirconium based alloys, oxidize. Compared to a steam oxidation, the degradation kinetic of the cladding is higher, on the one hand because of the high enthalpy of the zirconium-oxygen reaction (compared to zirconium-steam reaction), on the other hand because of the nitrogen contribution to the degradation. Temperature escalation and reaction runaway are expected and can rapidly lead to the loss of integrity of the cladding tubes. The objective of this PhD thesis was to affine the understanding of the high temperature air oxidation mechanisms of the two mostly used zirconium alloys in French PWR, Zircaloy-4 and M5. Special attention has been paid to clarify the role of nitrogen. As-received Zircaloy-4 and M5 claddings segments have been oxidized in a thermo-balance in air in isothermal conditions at temperatures between 800 C and 1000 C. Several characterization techniques (micro-Raman spectroscopy, EPMA, XRD, optical and scanning electron microscopies...) have been used to analyze the oxide layers. Identification and evolution of the different phases (monoclinic, tetragonal and cubic zirconia, zirconium oxynitride and ZrN) has been evidenced and analyzed at several step of the oxidation process. Oxidation mechanisms have been proposed and the better oxidation resistance of the M5 alloy, compared to Zircaloy-4 alloy, has been explained. The collected information will allow improvement of modeling aiming to predict the behavior of the claddings in various accidental situations with air ingress (temperature transients, evolution of the gas phase composition...). (author)