Plutonium and minor actinides are generated in nuclear fuel in non-negligible amounts during the normal operation of nuclear reactors. Proper management is mandatory due to the associated long-term radiation hazards. Moreover, reuse and recycling would be beneficial in terms of the circular economy. Different durable plans for safe storage in geological repositories as well as recycling strategies in advanced nuclear reactors are currently under consideration, but also alternative uses are investigated. Research at Joint Research Centre (JRC) in Karlsruhe is performed to support the safety and safety assessment of these applications.

One of the most promising solutions to reduce the amount of plutonium and minor actinides is their use and/or transmutation in dedicated reactors. In particular, the development of safe production processes for minor actinides-bearing fuels is one of the critical tasks for transmutation technology. In this context, a novel synthesis route, the hot compressed water decomposition of oxalate, was used to prepare homogeneous (U,Pu)O₂, (U,Pu,Am)O₂ and (U,Am)O₂ fuel samples. The relative advantages and the drawbacks of the method are discussed in comparison to established methods, and the main scientific findings are summarised.

In search of chemically stable americium compounds with high power densities for space applications, a number of ceramic materials was prepared and characterized in our Minor Actinide (MA) laboratory. Such ceramics are foreseen as power sources for space applications like Radioisotope Thermoelectric Generators (RTGs), and they have to endure extreme conditions including high vacuum, temperatures, and radiation fields. We summarize and compare the results of different americium ceramics synthesised at JRC Karlsruhe, with respect to their americium content, crystallographic stability in terms of swelling and amorphization resulting from self-irradiation due to alpha decay, stability under vacuum and different atmospheres, the behaviour of the ²³⁷Np decay product, presence of a natural analogue, etc.

Part of the Am-ceramic compounds synthesized for space applications are also suitable as waste forms in the management of plutonium and minor actinides. Thus, our systematic studies on compounds of Pu and Am (such as monazite, pyrochlore, or zircon-like) provide useful information on the long-term stability of various candidate ceramic waste forms under internal irradiation.

Nearly all advanced nuclear fission concepts are based on closed fuel cycles, and today the reprocessing of spent nuclear fuel has reached maturity in terms of uranium and plutonium recycling. Research remains necessary, however, to address the problem of the minor actinides (Am, Np, Cm). For Am, partitioning followed by transmutation is since long proposed. At SCK CEN, heterogeneous transmutation of Am in a dedicated Accelerator Driven System is being investigated since several decades. Specific to this concept is that the transmutation uses (U, Am)O₂ targets with elevated Am concentrations. Although the concept is straightforward, the practical problems related to the fabrication of such (U, Am)O₂ targets remain challenging. Recently, advances have been made in the so-called infiltration route in which porous uranium oxide microspheres are loaded with a nitric acid solution of Am followed by calcination and sintering. The present contribution reports recent advances in tailoring the porosity of uranium oxide spheres, their loading with inactive surrogate infiltrant (Nd³⁺), and first results with active infiltrant (Am³⁺).

The study of uranium oxides at different conditions is of paramount importance in the nuclear field, especially regarding characterization of the spent nuclear fuel behavior in dry storage scenarios. This work presents an assessment of the structural evolution occurring during the oxidation of the UO₂ spent fuel matrix into U₃O₇ in air. In particular, we report the results of X-ray diffraction and Raman spectroscopy analyses obtained on a variety of powdered samples prepared in order to cover a specific stoichiometry range in UO_2+x, with x varying from 0 to 0.30. The oxidation degree of each sample is confirmed by thermogravimetric analysis. Over the hyperstoichiometry range UO_2.00-UO_2.20 three structure transitions are detected, giving rise to three distinct regions associated with consecutive structural rearrangements. As for the UO_2.24-UO_2.30 range, the appearance of a tetragonal distortion and its increasing presence with the increase in oxidation degree is observed. These outcomes improve the understanding of non-stoichiometric uranium oxides, what can be used as a basis for further research on the stability of doped UO₂ matrices, such as ATF (Accident Tolerant Fuel) matrices, under in situ conditions, simulating both interim and final disposal.

_{Oxidation of uranium dioxide (UO2) nuclear fuel occurs during accident scenarios and storage conditions. The excess oxygen is incorporated into the fluorite structure and the resulting atomic-scale defect configuration significantly influences important bulk properties such as thermal conductivity and fission gas release. Previous experimental and modelling efforts have proposed distinct oxygen defect cluster configurations; however, most characterization techniques lack sensitivity to the local atomic structure or the oxygen sublattice and the resulting data cannot be used to validate predicted defect clusters. Here, we present results on single-phase UO2+x systems (x = 0.07 and 0.15) combining advanced experimental and modelling techniques to create high fidelity atomistic models of the oxygen defect clusters. In situ high-temperature neutron total scattering measurements with high sensitivity to the oxygen sublattice were performed at the Nanoscale-Ordered Materials Diffractometer (NOMAD) instrument at the Spallation Neutron Source (Oak Ridge National Laboratory). The data acquired at 600 °C and 1000°C were analyzed via Reverse Monte Carlo modelling techniques which consider both the long- and short-range structures. The analysis reveals evolving behavior as a function of oxygen content with simple clusters in the low O:M regime (UO2.07) and more complex, extended defects for higher oxygen concentrations (UO2.15). Our findings have implications in improving and validating potentials for Molecular Dynamics simulations to advance larger fuel performance codes.}