[en] In several countries utilising nuclear power plants as energy source, e.g. Finland, Germany and Sweden, the concept of direct geological disposal of spent nuclear fuel (SNF) is considered. This suggests to store SNF after discharge from the nuclear reactor for few years under water in reactor-site storing pools until transfer out of the reactor building to an interim storage facility. In case of dry interim storage, the irradiated fuel assemblies are emplaced in multi-purpose transport and storage casks, such as the CASTOR® of Gesellschaft für Nuklear-Service. Eventually, the SNF is transferred to a deep geological repository. For instance in Germany, Spain and the United States of America, multi-purpose casks with SNF are stored in dry interim storage facilities, mostly located in close vicinity to the nuclear power plants, where the SNF was discharged from. It is foreseen to keep the SNF in transport and storage casks until a final repository for high-level waste (HLW) will be in operation. Begin of operation of final repositories for HLW are expected in the next years (Finland and Sweden), respectively decades (e.g. Switzerland and other European countries). Since transport and storage casks are not licensed for final disposal, the SNF needs to be transferred into a suitable final disposal container. The integrity of the fuel claddings during the transfer from transport and storage casks to final disposal containers is of utter importance, to ensure that radionuclides (RN) will not be released by failure of the cladding tube. Already during reactor operation, the Zircaloy cladding tubes undergo different processes, such as transient elongation of the fuel rod, oxidation of the fuel- and the water-faced sides, formation of zirconium hydrides, as well as mechanical stress, which correlates with the fuel pellet swelling at higher burn-ups. Furthermore, in case of long-term safety analyses for final repositories, the contact of groundwater with the stored nuclear waste is considered. Depending on the classification of the national waste management organisations of the individual countries, at this, the cladding tube is considered as a technical barrier whose failure leads to a release of radionuclides, in particular activation products from the cladding as well as RNs from the fuel following the water penetration through the breached cladding to the SNF. In this context, various activation and fission products, such as, for example C-14, Cl-36, Cs 135 and I-129 are of significant interest due to their long half-lives, expected rapid release from the waste and high mobility in a repository system. This Ph.D. thesis addresses the occurrence of the beta and gamma radiation-emitting radionuclides Cl-36, Cs-137 and I-129, which are segregated to some extent from the fuel to the pellet-cladding interface during reactor operation. As part of these investigations, the inventory of radionuclides at the interface, respectively the inventory of the irradiated fuel and the Zircaloy cladding tube will be determined experimentally and compared to activation calculations obtained by Monte Carlo N-particle code and CINDER as well as webKORIGEN calculations. For the precise identification of the Cl-36 and I-129 inventory, it is necessary to develop a radiochemical separation process, which allows a quantitative analysis of both of the nuclides. Since Cl-36 has no and I-129 very low energy gamma-lines, an experimental inventory determination is only possible by a complex separation of these two isotopes from other radionuclides. The method developed in this work allows the profound separation and quantification of Cl-36 and I 129 from other interfering radionuclides, present in excess by many orders of magnitude in highly radioactive specimens, by using various extraction and precipitation steps. The results presented in this work provide information on the actual proportion of Cl-36 generated by neutron activation in highly radioactive waste, as well as on the amount of the initial Cl-35 impurity. Furthermore, it is shown, that I-129 is transported by a temperature driven process from the hot pellet centre to the peripheral area, adjacent to the colder cladding, resulting in the enrichment of the volatile radionuclide. In addition to the determination of Cl-36 and I-129 activities, other radionuclides (i.e. transuranium isotopes, fission and activation products) in the highly radioactive samples were determined and quantified with respect to enrichment factors in e.g. the fuel-cladding interaction layer in the spent nuclear fuel. The experimental radionuclide measurements performed within this thesis are in good agreement with values derived from calculations and are providing further knowledge on the distribution of radionuclides in spent nuclear fuel and thus also to the more accurate determination of the source term for irradiated fuel. In the second part of the thesis, different spectroscopic methods like scanning electron microscopy and energy / wavelength dispersive X-ray spectroscopy (SEM-EDS / -WDS) as well as X-ray photoelectron spectroscopy (XPS) and X-ray absorption spectroscopy (XAS) are used for the analysis of the fuel-cladding interface. Here, examinations of the occurrence of caesium / chlorine / iodine containing phases within the fuel-cladding interaction layer of different irradiated fuel-types (UO${}_X$ and mixed oxide fuel (MOX)) as well as different burn-ups (50.4 GWd/t${}_{HM}$ and 38.0 GWd/t${}_{HM}$) are performed. It is evident, that the chemistry at the interface is far more complex than often reported in the past literature, such as solely the occurrence of CsI, UO${}_2$ or a mixed compound of zirconium and uranium ((Zr, U)O${}_{2-<!--\; ???\; -->x}$). To the knowledge of the author, it is the first time that synchrotron radiation based Cl and I K edge measurements are performed on actual spent nuclear fuel fragments and Zircaloy cladding segments, resulting in an iodine-bearing compound with structural similarity to CsI and a yet to be identified chlorine-bearing compound. These measurements aim to provide a better understanding on the occurrence of iodine and chlorine-bearing agglomerates in spent nuclear fuel. The last part of the thesis focusses on possible corrosive effects of fission and activation product agglomerates on the interface layer, impacting the cladding integrity of spent nuclear fuel rods with regard to a prolonged dry interim storage. Hereto, U-O-Zr-Cs-Cl-I-containing phases are deposited on unirradiated Zircaloy and stored under inert gas and elevated temperature for an extended period of time. Afterwards the resulting samples are analysed by SEM-EDS and XPS to draw conclusions on their corrosion and embrittlement behaviour. Results obtained from this experiment indicate pitting corrosion processes on the Zircaloy cladding under the above mentioned conditions, enabled by the halogen-bearing species. Especially the role of the impurity chlorine in nuclear fuel and cladding material is highlighted by this experiment as its effects on possible cladding degradation under interim storage conditions is till now afflicted with uncertainties. However, it is evident, that more experiments are needed to elucidate this effect. The applied experimental conditions are not fully representative for the interim storage environment. The effect of radiation is not considered and a significant oxidation of the introduced UO${}_2$ was found unexpectedly, despite working under Ar atmosphere. The work of this thesis is performed in part within the EURAD work package “Spent fuel characterization and evolution until disposal”.