Zirconium alloys are commonly used in Pressurized Water Reactor as fuel rod cladding tubes. After irradiation and cooling in pool, the Spent Nuclear Fuel Assemblies are either transported for wet storage to a devoted site or loaded in suitable casks for dry storage on Nuclear Power Plant site. During dry transportation or at the beginning of dry storage, at temperatures around 400°C, the cladding experiences a creep deformation under the hoop stress induced by the internal pressure of the fuel rod. During creep a recovery of the radiation damage can occur that can affect the subsequent mechanical properties. The mechanical behavior of the cladding has been investigated in laboratory on two neutron irradiated cladding materials: fully recrystallized Zr-1%Nb and stress relieved Zircaloy-4. Creep tests under internal pressure were conducted at 400°C and 420°C. After depressurization and cooling, ring tensile tests were carried out at room temperature. In addition, Transmission Electron Microscopy observations have been performed after testing. The post-creep mechanical response exhibited a decrease of the strength compared to the asirradiated material. This decrease is associated with a significant recovery of the ductility, which becomes close to the ductility of the unirradiated material. The Transmission Electron Microscopy examinations, especially conducted on recrystallized Zr-1%Nb ring samples, revealed that the radiation defects have been annealed. It was also observed that, as for the unirradiated material, the deformation occurred homogeneously throughout the grains. No dislocation channel was indeed observed contrary to the as-irradiated material. These observations explain the recovery of the strength and of the ductility after post-irradiation creep that may occur during dry transportation or at the beginning of dry storage.

Zirconium alloys are commonly used in Pressurized Water Reactor as fuel rod cladding tubes. After irradiation and cooling in pool, the Spent Nuclear Fuel Assemblies are either transported for wet storage to a devoted site or loaded in suitable casks for dry storage on Nuclear Power Plant site. During dry transportation or at the beginning of dry storage, at temperatures around 400°C, the cladding experiences a creep deformation under the hoop stress induced by the internal pressure of the fuel rod. During creep a recovery of the radiation damage can occur that can affect the subsequent mechanical properties. The mechanical behavior of the cladding has been investigated in laboratory on two neutron irradiated cladding materials: fully recrystallized Zr-1%Nb and stress relieved Zircaloy-4. Creep tests under internal pressure were conducted at 400°C and 420°C. After depressurization and cooling, ring tensile tests were carried out at room temperature. In addition, Transmission Electron Microscopy observations have been performed after testing. The post-creep mechanical response exhibited a decrease of the strength compared to the asirradiated material. This decrease is associated with a significant recovery of the ductility, which becomes close to the ductility of the unirradiated material. The Transmission Electron Microscopy examinations, especially conducted on recrystallized Zr-1%Nb ring samples, revealed that the radiation defects have been annealed. It was also observed that, as for the unirradiated material, the deformation occurred homogeneously throughout the grains. No dislocation channel was indeed observed contrary to the as-irradiated material. These observations explain the recovery of the strength and of the ductility after post-irradiation creep that may occur during dry transportation or at the beginning of dry storage.

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