The aim of the project is to investigate the mechanical properties of irradiated cladding tubes under conditions of long-term dry storage. In order to prevent the cladding tubes from failing during dry storage, statutory stress and strain criteria are prescribed. The circumferential stress in the cladding tube depends on the modulus of elasticity, which in turn is influenced by other phenomena such as oxidation, structure of the oxide layer and the metallic material, hydrogen content and hydrogen platelet precipitation and orientation, precipitation of secondary phases and irradiation damage during and after irradiation in the reactor. This can lead to a decrease in ductility and an increase in embrittlement of the cladding tube. By studying the variation of mechanical properties such as modulus of elasticity and hardness, on a nano to micrometric scale, conclusions will be drawn on the evolution of these properties during storage over a long period of time to find out how the microstructure of the Zircaloy cladding tube, hydride content and morphology as well as the interaction of pellet with the cladding tube affect the mechanical properties of the cladding tube. In addition to the modulus of elasticity and hardness, other discrete physical events, such as incipient deformability, phase transformations or twinning, which are induced by deformation, are also to be investigated. Zircaloy in contact with nuclear fuel and Zircaloy from the plenum without contact with a fuel pellet will be used. By comparing these samples, conclusions can then be drawn about the influence of irradiation damage, oxide layer thickness, burn-up and the absence of nuclear fuel cladding interaction on the mechanical properties.