This paper is dealing with the corrosion and growth behavior of M5® and recrystallized low tin Zircaloy-4 irradiated as stress-free tubes in conditions representative of grids and guide-tubes of PWR fuel assemblies. The low-tin Zircaloy-4 tubes have reached equivalent burn-ups up to 93 GWd/tU (corresponding to a fluence of 21.0×1025 n·m−2, E>1MeV), and the M5® tubes have reached equivalent burn-ups up to 79 GWd/tU (corresponding to a fluence of 17.1×1025 n·m−2). Postirradiation growth was measured by comparing the distances between holes regularly distributed along the rod to the initial measurements. Postirradiation outer diameter oxide thicknesses were measured by eddy currents. Detailed characterizations of oxide layers have been realized through optical microscopy and scanning electron microscopy. Hydrogen uptake has been obtained through global hydrogen content measurements and metallographic examination of hydrides. Finally, the irradiation-induced microstructure of the metallic matrix has been observed by transmission electron microscopy. On these tubes, the free growth reached 1.9 % at 93 GWd/tU on Zircaloy-4 and 0.3 % at 79 GWd/tU on M5®. The external oxide thicknesses are far greater on Zircaloy-4 (∼60 μm at 93 GWd/tU) than on M5® (∼7.5 μm at 79 GWd/tU), and the oxidation rate is eight times higher on Zircaloy-4 than on M5®. Very high hydrogen content is achieved on Zircaloy-4, up to ∼1600 ppm (due to the low wall thickness and two-sided corrosion), whereas the maximum value on M5® is ∼100 ppm (despite the same geometry and corrosion conditions). On both alloys, the hydrogen pickup fraction is of the same order for these experimental empty rods as for previously analyzed fuel rod claddings, which may indicate the absence of a heat flux effect on the hydrogen uptake. Finally, potential correlation between growth, corrosion and hydrogen uptake will be discussed. Taking account of the results obtained on both experimental tubes and fuel rods, the effects of the presence or absence of heat flux through the wall thickness and of hydride rim at the metal/oxide interface are discussed, especially concerning the high burn-up corrosion acceleration on Zircaloy-4.

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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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