As the science and knowledge service of the European Commission, the Joint Research Centre's mission is to support EU policies with independent evidence throughout the whole policy cycle.
A JRC co-authored study  has examined the evolution of components of spent nuclear fuel by comparing actual spent fuel with lab results obtained on fuel analogues in simulated, accelerated timescale. Most of the trends observed were found to be comparable with characteristics of actual spent fuel. Ongoing programmes are also addressing the retrievability of spent fuel after extended storage and its behaviour under accident conditions.
Such comparisons help make safety assessments of the alterations that occur in conditions of prolonged storage of spent fuel while waiting for its final disposal in a geological repository. The two most common types of spent nuclear fuel were examined: irradiated uranium dioxide (UO2) and uranium (U) – plutonium (Pu) mixed oxide fuel (MOX).
JRC scientists gathered a large set of experimental data to study the effects of accumulated alpha-decay damage and helium in UO2 at different scales, from a microstructural level (accumulation of crystal defects, helium bubbles formation and lattice parameter swelling) up to macroscopic property changes such as hardening and thermal conductivity degradation. In order to reproduce cumulative effects expected during relatively long storage times within acceptable laboratory timescales, accelerated damage build-up conditions were applied by using unirradiated mixed U-Pu oxides with high specific alpha-activity.
The outcome of the experiments showed a non-negligible extent of lattice swelling, which refers to the physical dimensions of the crystal structure in the UO2 fuel matrix. If the microstructural lattice swelling results in macroscopic swelling of the fuel pellets in spent fuel rods, additional stresses could be applied on the metallic cladding which contains the fuel, which may pose a safety concern. For a full comparison, further studies are being carried out.
The study also pointed to saturation of most macroscopic property changes – meaning a stabilised condition with no further alterations – for a simulated timescale corresponding to spent fuel after decades or centuries of storage (the exact time scale depends on parameters like fuel composition and burnup).
The retention of the mechanical integrity of spent nuclear fuel rods during and after storage is affected by the radioactive decay process. Other mechanisms that may cause property degradation are related to the temperature history experienced by the spent fuel rods, and are determined by the behaviour of hydrogen present in the cladding.
Results obtained under accelerated conditions need to be compared to the properties of actual spent nuclear fuel, in order to define the applicability of short term lab scale experimental data to the prediction of potential spent fuel rod alterations over many decades or centuries as a function of storage conditions.
After discharge from the core of a nuclear reactor, spent nuclear fuel rods are cooled for several years in water pools at the reactor site. Once the heat generated by radioactive decay processes has become sufficiently low, the spent fuel is moved to dry or wet interim storage, where it is stored before being moved to a final disposal in a geological repository or to a reprocessing plant. Typically, interim storage facilities are licensed to store spent fuel for a few decades (e.g. 40 years in Germany).
Due to the expanding timeline for the definition and implementation of the disposal in geological repository, in many countries the duration of the interim storage will have to be extended to longer periods, possibly encompassing a time scale of the order of a century. A sound scientific basis of data and knowledge is therefore required to assess the safety of spent fuel during extended storage, and in all retrieval, transportation and repackaging operations thereafter. In particular, it is important to determine and characterise all mechanisms that may affect the mechanical integrity of the fuel and its first containment barrier, the cladding.