For the high level waste accumulating in the Federal Republic of Germany, according to the current prognosis spent fuel elements with a heavy metal mass of about 10,500 t, the current concept provides for an interim storage period of up to 40 years. According to current knowledge, however, the permissible storage periods will be exceeded, which requires an extension of the safety demonstration. Within the framework of a research project funded by the BMUV, the present doctoral thesis aims to develop and apply a multiscale method that provides information on the sequence of the multi stage radiation induced damage of the cladding tubes of the spent fuel elements and effects on the structural mechanical properties of the cladding tube. After the fuel matrix, the cladding tubes of the fuel rods of the fuel assemblies (FAs) are the last barrier against an escape of the radionuclides contained in the fuel. Accordingly, the investigation of the structural integrity of the spent fuel elements is of crucial importance for further safety statements regarding long term interim storage. Among various damage processes, radiation induced damage to the cladding tube material and the resulting structural mechanical effects are of particular importance for the integrity of the FAs. This is a multi scale process sequence that starts at the atomic level of damage and extends to the macroscopic level with the structural mechanical change of the cladding tube material. Using the Monte Carlo method, the first step in the hierarchical improved modelling sequence is to determine the effects of the radiation field on the cladding material (Zircalloy 4) of the fuel element. The energy distribution of primary defects for a typical LWR operating phase was obtained with the neutronic damage code SPECTER. Next, the primary defects cascades in the lattice of Zry 4 material were simulated using the molecular dynamics program LAMMPS to determine the spatial distribution, energy spectrum, and density of interstitial atoms and vacancies at the end of the cascade. Following the end of the cascade, defects interact with each other and diffuse through the multi crystalline metal lattice, which becomes increasingly damaged. This is done based on the KMC method (Object Kinetic Monte Carlo method), whereby the size and density distribution of defect clusters or microstructures and their degree of agglomeration are simulated or determined as a function of the irradiation dose. Furthermore, the temporal change of these radiation induced microstructures is investigated and determined for dry storage temperatures using a cluster dynamic routine and rate theoretical algorithms. With a view to testing and validating the overall multiscale model, the results of the model calculations and simulations are compared with those obtained from TEM (transmission electron microscopy) measurements on the distribution and structure of radiation induced microstructures. This investigation confirms the accuracy of the multiscale method within the range of variation of the experimental measurement data and the results of the individual simulation models. Due to the effects of the microstructural material damage (which also results from the multi scale simulations) on the material properties of the cladding tube material, the structural mechanical material parameters are investigated and a significant change is found using the characteristic parameters yield strength and radiation growth ( elongation) as a function of dose and operating temperature during wet and dry storage (intermediate storage) as an example .Another aspect related to material damage is the impact of physico chemical processes such as hydride formation and precipitation, which, in combination with radiation damage, cause the material properties to change. For an assessment of the component behaviour or the integrity of the cladding tubes, the further development of the microstructural damage in the direction of crack initiation and propagation as a result of thermomechanical stresses and external mechanical loads is of particular importance. These effects are the subject of other fracture mechanics investigations based on the method of continuum mechanics and finite elements. In summary, it can be stated that with the development and application of a multiscale computational model, the simulation of the entire process chain of radiation induced material damage, starting from the atomic level, has been realised and the procedure enables precise knowledge about the temporal change of the safety relevant structural mechanical property variables of the irradiated cladding tube material. The evaluation results obtained by this work can fill a critical data gap in studies related to the extension of interim storage periods.