In nuclear plants, some accidental situations can result in air exposure of Pressurized Water Reactor (PWR) fuelassemblies: air ingress following a breach in the reactor vessel, deflooding during handling, spent fuel storagepool deflooding. Deprived of cooling source, the assemblies temperature raises and the fuel cladding, made out ofzirconium 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-steamreaction), on the other hand because of the nitrogen contribution to the degradation. Temperature escalation andreaction 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 mechanismsof the two mostly used zirconium alloys in French PWR, Zircaloy-4 and M5®. Special attention has been paid toclarify the role of nitrogen.As-received Zircaloy-4 and M5® claddings segments have been oxidized in a thermobalance in air in isothermalconditions at temperatures between 800°C and 1000°C. Several characterization techniques (micro-Ramanspectroscopy, EPMA, XRD, optical and scanning electron microcopies...) have been used to analyze the oxidelayers. Identification and evolution of the different phases (monoclinic, tetragonal and cubic zirconia, zirconiumoxynitride and ZrN) has been evidenced and analyzed at several step of the oxidation process. Oxidationmechanisms have been proposed and the better oxidation resistance of the M5® alloy, compared to Zircaloy-4alloy, has been explained.The collected information will allow improvement of modeling aiming to predict the behavior of the claddings invarious accidental situations with air ingress (temperature transients, evolution of the gas phase composition…).
