[en] Full text: The long-term storage of spent nuclear fuel in geologic repositories envisages the isolation of the material from environmental factors of influence by a multi-barrier system. The cladding of the material is considered to act as one of the barriers in the near field close to the surface of the material, and therefore plays an important role in the isolation of radionuclides from the geosphere during final storage. The interaction of cladding and the surface of spent fuel samples (UOX, BU 30 GWd/t and MOX, BU 12 GWd/t) was investigated with static long-term leaching tests. The cladding tubes of spent fuel rodlets were perforated in order to simulate defects. Subsequently, the samples were immersed in water for a period of about 4 years. Analysis of the bulk solution composition was performed with ICP-MS after an immersion period of 3 weeks and 4 years. After 4 years of treatment the fuel elements were separated from the solution and cut into specimen for investigation with optical microscopy and SEM-EDX, as well as measurement of physical parameters (gap width, micro hardness). A macroscopical optical investigation of the pre-set defects showed the formation of a layer of yellow and brown products on the outer surface of the cladding close to the openings, indicating that the spent fuel matrix was partially oxidized to U(VI) which was leached into the solution. An investigation of the cut specimens with optical microscopy showed that the formation of alteration products in the fuel-cladding gap proceeded in a different way for the MOX and UOX material sample. In the case of the UOX fuel, a porous layer of deposits is formed on the cladding which results in irregular gap width data. Almost no alteration products were found on the inner cladding surface of the MOX material. The Zircaloy alteration layer was investigated with SEM-EDX. A qualitative and quantitative assessment of the alteration product formation was performed by thermodynamic equilibrium calculations using the overall solution properties of the system as input. Since the immersion tests were conducted under an N₂ atmosphere with an O₂ content of 4 %, oxidation of the UO₂ matrix to U(VI) is expected. The plutonium present in the MOX sample was initially present in the +IV oxidation state (PuO₂). The calculations predict that UO₂ is converted to U(VI) (with UO_2.333 as the solubility-controlling phase) and PuO₂ is present as Pu(IV), both in solution and in the solid, under the conditions given. The Zircaloy material is thermodynamically predicted to form the corrosion product ZrO₂ under the experimental conditions. Since ZrO₂ forms a monoclinic phase at ambient temperatures, it is expected that the material present in the layer will not form binary mixtures of the solid-solution type with the oxides PuO₂ and UO₂, but possibly with the oxides of U(VI). This suggests that ZrO₂, once formed on the cladding surface, might show a possible retention capacity for hexavalent U, but not so for Pu and several of the fission product elements in the spent fuel. This hypothesis is coherent with the observed high release rate of fission products in the case of the MOX sample. (author)