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International audience; 238Pu-doped UO2 was characterised periodically to investigate the long-term effects of α-self-irradiation on spent nuclear fuel. Samples of two compositions (1.25 and 5% wt 238Pu), were studied by means of X-ray diffraction (XRD), Raman spectroscopy, and transmission electron microscopy (TEM). The cumulated dose received by the samples was measured in displacements per atom (dpa) to allow comparison between the two samples and to the literature values. XRD characterization showed that the lattice swelling approached saturation at 0.3% of the lattice parameter of the cubic cell at 0.4 dpa damage accumulation. Microstrain decreased for the whole duration of the study, as a result of point defect recombination into dislocation loops that were observed by TEM for dpa as low as 0.072. Raman highlighted the broadening of the T2g band and the increase and broadening of the defect triplet band. Both batches of samples followed the same trend if plotted against dpa, proving that for these dopant concentrations the self-irradiation effects do not depend on the rate of radiation damage creation in the matrix.

International audience; • SIMS and TEM techniques were combined to compare the thermal behaviour of Cs and Xe in UO 2 . • Both elements form bubbles with different growth kinetics. • At 1600 °C, caesium is found to be highly mobile in the UO 2 matrix while Xe distribution does not evolve.

International audience; This study aims to evaluate the influence of a metallic fission product, molybdenum, on the behaviour of caesium in UO2 at high temperature (1600°C) under reducing atmosphere. Both elements were successively introduced by ion implantation (firstly Mo, then Cs) at room temperature at a fluence of 1016 ions/cm². We show that in these co-implanted samples, the defects created by the Mo implantation render possible the nucleation of nanometric size bubbles during Cs implantation at room temperature. SIMS and TEM techniques were coupled in order to characterize the migration of both elements, as well as the UO2 microstructure evolution after annealing at 1600°C. We found that Mo is more mobile in presence of Cs, which may be related to an increase of available vacancies in the material produced by Cs implantation. On the contrary, the concentration profiles of Cs remain quite similar than when it is solely implanted, and its release percentage remain the same both in absence and presence of Mo. After annealing, both elements are found to be associated as Mo metallic precipitates and Cs bubbles distributed over the same depth (~150 nm). Comparing with the Cs bubbles distribution in absence of Mo, it is clear that the formation of Mo metallic precipitates hinders the Cs bubble migration. Our study highlights the role of dislocations on the nucleation of bubbles and metallic precipitates.

In the context of a comprehensive campaign for the characterisation of transmutation fuels for next generation nuclear reactors, the melting behaviour of mixed uranium-americium dioxides has been experimentally studied for the first time by laser heating, for Am concentrations up to 70 mol. % under different types of atmospheres. Extensive post-melting material characterisations were then performed by X-ray absorption spectroscopy and electron microscopy. The melting temperatures observed for the various compositions follow a markedly different trend depending on the experimental atmosphere. Uranium-rich samples melt at temperatures significantly lower (around 2700 K) when they are laser-heated in a strongly oxidizing atmosphere compressed air at (0.300 ± 0.005) MPa, compared to the melting points (beyond 3000 K) registered for the same compositions in an inert environment (pressurised Ar). This behaviour has been interpreted on the basis of the strong oxidation of such samples in air, leading to lower-melting temperatures. Thus, the melting temperature trend observed in air is characterized, in the purely pseudo-binary dioxide plane, by an apparent maximum melting temperature around 2850 K for 0.3 < x(AmO2) < 0.5. The melting points measured under inert atmosphere uniformly decrease with increasing americium content, displaying an approximately ideal solution behaviour if a melting point around 2386 K is assumed for pure AmO2. In reality, it will be shown that the (U, Am)-oxide system can only be rigorously described in the ternary U-Am-O phase diagram, rather than the UO2-AmO2 pseudo-binary, due to the aforementioned over-oxidation effect in air. Indeed, general departures from the oxygen stoichiometry (Oxygen/Metal ratios ≠ 2.0) have been highlighted by the X-ray Absorption Spectroscopy (XAS). Finally, to help interpret the experimental results, thermodynamic computations based on the CALPHAD method will be presented.

The development of tailored targets for the production of radioactive isotopes represents an active field in nuclear research. Radioactive beams find applications in nuclear medicine, in astrophysics, matter physics and materials science. In this work, we study the use of graphene both as carbon source for UO2 carbothermal reduction to produce UCx targets, and also as functional properties booster. At fixed composition, the UCx target grain size, porosity and thermal conductivity represent the three main points that affect the target production efficiency. UCx was synthesized using both graphite and graphene as the source of carbon and the target properties in terms of composition, grain size, porosity, thermal diffusivity and thermal conductivity were studied. The main output of this work is related to the remarkable enhancement achieved in thermal conductivity, which can profitably improve thermal dissipation during operational stages of UCx targets.

Uranium-molybdenum (UMo) alloy powder embedded in an Al matrix is considered as a promising candidate for fuel conversion of research reactors. A modified system with a diffusion barrier X as coating, UMo/X/Al trilayer (X = Ti, Zr, Nb, and Mo), has been investigated to suppress interdiffusion between UMo and the Al matrix. The trilayer systems were tested by swift heavy ion irradiation, the thereby created interaction zone has been analyzed by scanning transmission electron microscopy (STEM) and energy-dispersive X-ray spectroscopy (EDX). Detailed structural characterization are presented and compared to earlier μ-XRD analysis.

Within Europe 241 Am is a feasible alternative to 238 Pu that can provide a heat source for radioisotope thermoelectric generators (RTGs) and radioisotope heating units (RHUs) to be used in Electrical power sources used in outer planet missions. In the EPSO-SPACE* project developed as an exploratory research at JRC, a new type of nuclear power source, to provide heat and electricity for deep space missions is considered. It will consist of a 241 Am based compound with aerogel thermal insulation and thermocouples in close contact with the heat source. The assembly is inside an iridium encapsulation to withstand all relevant accident scenarios. The design will be modular and combinable for required electrical output.Several candidate Americium compounds will be investigated for chemical stability at high temperature and for self-irradiation damage. New thermo-electric converter materials containing actinides will be assessed, and a robust encapsulation designed. Safety analyses will be performed including launch explosion and re-entry accidents. The research will conclude in a conceptual design of a prototype power source. In the first part of this study, americium dioxide will be considered from the point of view of its chemical durability and of its behavior against radiation damage and helium formation, two aspects to be carefully investigated due to the high alpha-activity of the americium. Transmission electron Microscopy (TEM) and helium thermal desorption spectrometry (TDS) experiments will be described and results on aged (more than 30 years) AmO2 reported. Some comparison with 238 PuO2 based RTG’s will be discussed.

UO2 samples doped with the strong α-emitter 238Pu were periodically characterised to study the potential effects of α-self-irradiation on spent nuclear fuel during storage. The degradation of the thermal properties and microstructure of the samples were monitored using the Laser Flash technique (LAF), X-Ray Diffraction (XRD), and Scanning and Transmission Electron Microscopy (SEM, TEM). The thermal conductivity decreased with a non-linear behaviour as a function of the α-dose. Three clearly different stages in the thermal conductivity degradation could be detected, and saturation was reached at relatively low doses (0.035 dpa). In combination with analyses of the microstrain measured by XRD, and TEM images, these three phases could be attributed to three stages of radiation-induced defect generation and clustering. SEM inspection showed the thermal conductivity degradation cannot be due to loss of integrity or grain boundary opening. The radiation damage produced resulted in the formation of dislocation loops as observed by TEM.