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International audience; GenIV fast reactors (FR) development would increase the sustainable use of the uranium resources and mitigate long term impact of nuclear waste generated by nuclear power production.These reactors, however, require the design and development of nuclear fuel (NF) withstanding the most extreme irradiation and temperature conditions. High plutonium content MOX fuels are envisaged (for example 30% Pu-enriched in MYRRHA). A deep and complete understanding of the behaviour of such MOX fuels for FR is hence one of the key aspects towards the licensing process of the reactors. The high alpha-decay activity from the plutonium content in MOX fuel leads to the formation of large amount of defects which will induce measurable effects even before irradiation in the reactor.The fluorite structure common to all actinide dioxides is known as being a crystal configuration resistant to radiation damage. Nevertheless, high alpha-dose can affect the stability of materials envisaged as fuels. Furthermore, when the alpha-particle comes to rest it becomes a helium atom that can alter the microstructure of the material, e.g. by forming microscopic bubbles.At the downstream of the fuel cycle, the radiotoxicity of the irradiated fuel will be mainly due to the minor actinides (MA) accumulation, including plutonium. Over the medium-term period alpha-decay will also significantly contribute to the heat load, e.g. during extended storage in the case of LWR.In the present work we report on new experimental observations obtained by TEM, LAF, XRD, RAMAN and DSC of the self-irradiation alpha-damage effect on Uranium-Plutonium MOX fuel samples with a variety of Plutonium content and of different isotopic composition, different storage time and hence different damage levels.The results for samples with higher alpha-dose can be used to forecast the ageing consequences affecting the behaviour of spent fuel over several centuries or millennia of storage/disposal.

International audience; Spent nuclear fuel (SNF) of light water reactors (LWR) is constituted of a matrix of unfissioned UO2 containing a small fraction of fission products (FP) and minor actinides (MA). These MA are mainly α-emitters with very long half-lives: therefore, SNF will keep α-self-irradiating for millennia after discharge from the reactor, resulting in widespread changes of the thermophysical properties of the material. In order to properly coordinate and license any disposal strategy for SNF, the long-term effect of α-self-irradiation has to be known and anticipated. Due to the lack of real old SNF, a proxy system to study the effect of α-irradiation on the SNF matrix is provided by the synthesis of UO2 doped with short-lived α-emitters. In this way, a significant amount of radiation damage can be stockpiled over a laboratory timescale in a simpler surrogate system that does not combine radiation damage with chemical or density gradients (built-in in real SNF). In the present work, UO2 doped with 238Pu was produced to study the effect of α-self-irradiation on the crystalline disorder and lattice swelling. The composition of the samples was carefully chosen, based on the dopant specific activity, in order to reach saturation of the lattice parameter swelling within the 3 years timespan of the project. Samples were periodically characterized by means of XRD and Raman spectroscopy up to 0.4 dpa, equivalent to a spent nuclear fuel with a 40 GWd/t UO2 burnup stored during 300 years, or representative of a 65 GWd/t spent MOX fuel (45 % Pu) after 25 years of storage. Lattice swelling as a function of dpa was assessed with very good accuracy and benchmarked against literature data: saturation was reached at a value of 0.3 % around 0.4 dpa. For the first time, microstrain was also monitored in (U,Pu)O2 as a function of self-irradiation. Periodic Raman spectroscopy acquisitions on (U,Pu)O2 as a function of the dose represent an innovative probe, and they showed fast and progressive degradation of the structural order up to 0.1 dpa, and a slow but persistent increase up to 0.4. SEM characterizations were additionally performed and highlighted that no loss of structural integrity is associated with this microstructural evolution, at least within the measured dpa range.

International audience

Best poster presentation award of symposium EN17 at MRS Spring Meeting in Phoenix; International audience; Spent nuclear fuels include a variable amount of minor actinides incorporated within the UO2 matrix as a result of the ongoing nuclear reactions during operation time. These actinides contribute to the radioactivity of the spent fuel, whose most long-living component is constituted by α-emission, lasting up to millennia ad even farther. The accumulation of α-damage and radiogenic He affects the microstructure of the material and might result detrimental for the spent fuel integrity. For this reason, it is fundamental to know exactly how the spent fuel evolves in order to accurately predict its behaviour in the long term.To investigate such a slow evolution in a laboratory timeframe, a suitable way is the preparation of UO2 surrogates doped with highly-emitting actinides that will produce high levels of damage and radiogenic He in a shorter time. In this work, UO2 disks containing tailored amounts of highly α-emitting 238Pu were produced by powder coprecipitation and sintering and successively characterized. The coprecipitation route has been chosen in order to obtain a solid solution of the two oxides from the very beginning of the process, while the successive sintering ensured a homogeneous distribution of the α-dopant as well as a nearly complete densification (95%TD, compliant with the real nuclear fuel). The two chosen Pu percentages are 2.5 and 10% and were selected so that all the monitored evolving material properties should come to saturation within 2 years.The samples were then divided into 3 batches and stored at different temperatures: liquid nitrogen (approximately -195°C), room temperature (about 25°C) and wet storage conditions (around 200°C). Particular precaution was taken in the transportation between the storage and the actual characterisation. The choice of these three temperatures should allow differentiating the nature of the defects formed and their temperature-dependant evolution, in particular during spent fuel storage.The samples evolution was monitored by mean of several characterization techniques: XRD (structure), TEM (microstructure), DSC (defect energy), LAF (thermal properties), Helium thermal desorption spectrometry, mechanical testing, and Raman spectroscopy. The outcome of this work should help predicting the long term behaviour of spent fuel.

International audience