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Invited talk EN.17.04.01 given at MRS Fall Meeting 2019. Symposium EN17—Structure–Property Processing Performance Relationships in Materials for Nuclear Technologies.MRS Fall Meeting, Boston, MA, USA. December 1 - 6, 2019; International audience; The majority of periodic table elements are created within the conventional uranium dioxide fuel matrix during its time in pile. These elements are accommodated in quite different ways within the nuclear ceramic. Ba, Zr, Y and the lanthanide atoms are typically substituted within the uranium dioxide matrix. At high burnup, the number of phases that form and the differences between them can result in the development of highly complex structures and microstructures. Similar effects might be encountered in manufactured advanced actinide fuels for the transmutation of minor actinides or in fuels containing significant amounts of burnable poisons (typically Gd). Those systems constitute then a large class of compounds, generally with fluorite or fluorite-derived structures, exhibiting high radiation tolerance, possibly related to their peculiar ability to accommodate a variety of defects and to exist as nonstoichiometric compounds within a large homogeneity range. Nevertheless, a variety of behaviours and different phases is observed after quenching or annealing in several of the peudo-binary systems of type (Ln,U)O2-x, suggesting that the thermal history of those systems is of paramount importance. Some of these systems, like (U,Ce)O2-x are particularly complex, display an intriguing structural variety, and therefore present a fundamental interest. In this talk we would try to address some of those interesting structural features induced by aliovalent substitutions in UO2.

International audience; The development of realistic models for fission products behavior during a nuclear severe accident requires experimental data on fission products speciation into the fuel. In this context, several batches of dense UO$_2$ samples containing fission products surrogates under different chemical forms have been prepared and sintered, to be further submitted to thermal treatments in order to characterize fission products speciation under controlled temperature and oxygen potential conditions.To that end, a large number of as-fabricated samples from these experiments were analyzed by ICP-AES and ICP-MS after dissolution. A separation step by liquid chromatography on UTEVA resin was essential before the ICP-AES measurements to overcome the problems of spectral interferences and matrix effects caused by uranium.This study emphasizes the complementarity of these two techniques in nuclear fuel characterization. The advantage of the ICP-AES analysis on simultaneous device is explained in details. The importance of the integrated collision reaction cell in ICP-MS to avoid many problems of polyatomic interferences for the quantification of Cs and Ba is highlighted.

International audience; During operational conditions of pressurized water reactors (PWR) nuclear fuel is highly irradiated. This process induces many modifications of the initial properties, such as porosity, thermal conductivity or the apparition of a particular structure named High Burn-up Structure (HBS), characterized by sub micrometric grain size and high population of closed pores, among others (1). This work focuses on the evolution of the microstructure of simulant materials to model nuclear fuel under conditions similar to those found in the HBS. The chosen system to study was U-Nd-O, being Nd used as dopant since it is the most abundant lanthanide fission products. At low temperature and low concentrations of Nd, the system is partially miscible and presents a region where it is biphasic. This region is formally called as miscibility gap. Previous studies carried out on the U-Nd-O ternary system highlighted that the miscibility gap at room temperature exists when the concentration of Nd is higher than 9 at% (2,3). With rising temperature, the solubility increases and the biphasic system becomes monophasic. The objective of this paper is to analyze the system when it is submitted to a strain energy that appears when the initially homogeneous phase at elevated temperature crosses a biphasic domain at a lower temperature (enters the miscibility gap). Thus, the most salient feature of this study will be to observe if this internal strain would produce dislocations which would contribute to the formation of the HBS. In the current work, three different samples with 4, 17 and 25 at% of Nd were studied. The evolution of the crystalline structure in temperature and the presence of dislocations at room temperature as a consequence of phase separation are analyzed through means of High Temperature-X-Ray Diffraction (HT-XRD), High-Temperature Scanning Electron Microscopy (HT-SEM), and Transmission Electron Microscopy (TEM). Experimental results are confronted to thermodynamic modelling calculated through the CALPHAD method using the Thermo-Calc code. In this paper a first approach on the evolution of the microstructure as a function of the concentration of Nd is also proposed.

International audience; During the irradiation of nuclear fuel in a light water reactor (LWR), fission products (FP) are created and retained in the fuel, modifying its microstructure and properties (thermal conductivity, creep behavior, etc.). In the case of failure of all the three containment barriers, as it could the case during a severe accident, FPs could be released from the fuel to the environment. The chemical nature of FP released from the fuel strongly depends on the scenario and the accident conditions involved (temperature, type of fuel, burn-up, oxygen potential, etc.). Moreover, chemical nature greatly affects FP behavior during their transport in the reactor and hence the source term. FP speciation in the fuel was studied in order to provide experimental data needed to support the speciation mechanisms currently proposed in the literature. Within the framework of the International Source Term Program, the VERDON-3 and-4 tests were devoted to high burn-up MOX fuel behavior and FP releases respectively under oxidizing and reducing conditions at very high temperature (up to 2570 K and 2800 K respectively). Qualitative and quantitative analyses on these fuel samples made it possible to obtain valuable information on fission product behavior in the fuel during the test as a function of the atmosphere. This presentation will provide a series of detailed characterizations including scanning electron microscopy (SEM) and optical microscopic (OM) observations, as well as electron probe microanalyses (EPMA) and Secondary Ion Mass Spectrometry (SIMS) of the VERDON-3 and -4 fuel samples before and after the test.

International audience

International audience; Thanks to its precision, its reproducibility and its stability, Electron Microprobe is a well suited technique for accurately analyzing nearly all chemical elements at concentration levels down to few 10’s ppm with a spatial resolution of about 1 µm, which is relevant to microstructures in a wide variety of materials and mineral specimens. For irradiated samples, EPMA is also one of the technique of choices and can support nuclear fuel development, control, metallurgical or glass analysis. It reveals fine compositional details and the distribution of main and trace elements across the surface of the sample.CAMECA leader in scientific instruments has been manufacturing Electron Microprobe (EPMA) since 1958 and will present its latest CAMECA shielded EPMA model, SKAPHIA released in 2016 for hotlab facilities. One example of the LECA/STAR shielded EPMA in hotlab handling nuclear fuel will be highlighted. In this example, the microstructure and chemical evolution of fuel pellets submitted to two different nuclear severe accident scenarii are shown. EPMA helped to highlight the impact of the atmosphere (i.e the oxygen potential) on the behavior of the fuel and fission products during these tests. These studies are of particular importance to evaluate and better predict the consequences of such an accident in term of contamination.

International audience

International audience; Despite the very high level of control and safety in Nuclear Power Plant, they are not exempt from device malfunction, human errors or even natural disasters, which may lead to nuclear incidents or accidents. When consequences involve a high degradation of the reactor core (melting), associated with the release of Fission Products (FPs) and other radioactive materials from the reactor core, it is classified as a Severe Accident (SA, TMI-2, Chernobyl and Fukushima for instance). These released FPs may be transported by air masses or water and thus cover extensive areas and affect all living beings due to their radiological effect and may as well act as poisons. The amount and isotopic composition of the radioactive material released from the core is called Source Term, and its assessment has been the main objective of several international research programs for more than thirty years. These research programs are commonly classified into two main groups, according to their approach (1) Integral programs, such as PHEBUS FP, studying the response of a whole nuclear core during a severe accident, in a reduced scale; (2) On the other hand, analytical programs studying the fuel and FPs when submitted to accidental conditions, by means of Separate-Effect Tests (SET). Examples of the latter are the HI/VI, VEGA, VERCORS and VERDON programs. As result from all the research programs an extensive experimental database has been generated. However, up to now, predicting correctly the FPs release from UO2 and/or MOX fuels in SA conditions is still a significant and very important challenge since there are many remaining uncertainties. In order to improve these estimations, the global fuel and FPs behavior during the accidental sequence must be better understood, and specific emphasis has to be put on mechanisms which promote FPs release and fuel relocation. One of the most useful ways to do that is to perform appropriated annealing treatments with representative thermal transients in order to measure the absolute level and kinetics of the released FPs. To understand the promoting mechanisms, theses FPs release measurements have to be coupled with the corresponding fuel micro-structural changes resulting from these thermal transient.To this end, since the last decade, CEA has set up two complementary research axes, aiming at reproducing conditions representative of nuclear severe accidents, using both high burn-up irradiated fuel samples and model materials. The first axis corresponds to the VERDON program and deals with commercial UO2 and MOX fuels irradiated in French PWR. Model materials (often called SIMFUELS) consist in natural UO2 doped with stable isotopes of FP in concentrations that match a targeted burn-up. Therefore, SIMFUELS are representative of irradiated nuclear fuels but without their radioactivity. The importance of such materials lies in the possibility of using powerful characterization techniques, such as X-ray Absorption Spectroscopy, which today are unavailable for large samples of irradiated nuclear fuels.The present paper, organized in four main parts, presents successively the experimental facilities available at the CEA Cadarache and Marcoule centers together with the corresponding RetD axes SA experimental VERDON laboratory and associated annealing test device as MERARG, Analytical and micro analysis laboratories by which all the pre- and post-test fuels examinations, supported by analytical development on simulated corium samples, are performedUse of SIMFUELS methodology.The last part of the paper focuses on results obtained with this general approach, with special emphasis on VERDON-1 test.

International audience; The development of realistic models for fission products behavior during a severe accident requires experimental data on fission products speciation into the fuel (as irradiated and during the transient) together with thermodynamic predictions. Due to the limitations in terms of experiments and characterization techniques available to study fission products speciation in nuclear irradiated fuels, simulant materials (SIMFUEL manufactured thanks to conventional process) are often used. However, these materials do not allow the study of volatile fission products because they are totally released during the sintering stage. In the case of SIMFUEL with implanted fission products surrogates, the chemical state and the corresponding evolution of the phases during the thermal sequence are not entirely representative of the real case. An innovative method to study volatile fission products speciation in SIMFUEL samples has thus been developed. Spark plasma sintering was used to synthesize dense SIMFUEL samples containing cesium in collaboration with the JRC-Karlsruhe within the frame of the GENTLE program. Thermodynamic calculations made using the TAF-ID v.6 associated with the FACTSage software enabled to determine the conditions in which to perform thermal treatments. In-temperature XAS experiments in conditions representative of a sevre accident are developed in collaboration with FAME-UHD beamline at the ESRF (Grenoble, France). Additional thermal treatments will also be performed to enable further characterizations (ceramography, SEM-EDX...).

International audience

International audience

International audience; Up to now, plutonium dioxide is the major end product of nuclear fuel reprocessing. Due to the radioactive properties of plutonium, handling and management both of plutonium dioxide and of contaminated wastes involve to take into account the radiolysis phenomena plutonium can induce. For instance, plutonium is known to be a highly hygroscopic compound. Kept under a humid atmosphere, up to 3% wt. water can be sorbed on the surface of plutonium dioxide. The sorbed water then is going to undergo a radiolytic decomposition and this latter will lead to hydrogen evolution. It has been shown through our studies and other reported results that hydrogen formation kinetic depends on the specific surface area of plutonium dioxide, the isotopic composition of plutonium and the relative humidity in the atmosphere. Unfortunately no clear dependences between the rate of hydrogen formation and these parameters can be established. However, it has been shown that in many cases hydrogen amount in the atmosphere reaches a steady-state. Such behaviour is probably linked to surface evolution of plutonium dioxide in presence of moist air and under radiolytic decomposition of water. Studies are still in progress to identify surface changes due to these phenomena. Hydrogen can be produced under radiolysis of some plastic waste contaminated by plutonium dioxide too. Whereas $\gamma$-radiolysis of plastic materials is quite well described, $\alpha$-radiolysis of such compounds, especially the ones in close contact with actinides, is very poorly known. It can be demonstrated that only a small part of the whole energy emitted by plutonium dioxide is going to be deposited in the material and leads to a radiolytic decomposition with hydrogen and others hazardous gaseous products (CH$_4$ for instance) formation. Taking into account the amount of energy transferred to the plastic material, $\alpha$-radiolytic yield of hydrogen can be calculated and then used for waste repository scale-up. It should be noticed that plutonium dioxide can exhibit specific behaviour towards some radiolytic products. For instance, it has been shown that chlorine arising from PVC radiolytic decomposition is going to be retained mainly on plutonium dioxide surface, whereas without plutonium dioxide chlorine is evolved as gaseous HCl. In the same way, $\alpha$-radiolysis induced by plutonium dioxide on several other materials is now studied, such as hydrogen, carbon oxides and short-length alkane production from some lubricating compounds used in fuel pellet manufacturing or from specific waste embedding materials (bitumen, mineral polymers). As the present trend is still to increase fuel burn-up in one hand, and, in the other hand to manufacture new fuel with increasing amount of plutonium, it appears necessary to look closer at the radiolytic phenomena induced by actinide oxide and especially plutonium dioxide to actualize the safety analysis for handling and management of future nuclear fuels and wastes.

International audience; The cladding material envisaged for the Gas-cooled Fast Reactor (GFR) is a SiC f /PyC/SiC composite with a β-SiC coating on the surface that is subjected to temperatures going up to 1500 K in nominal operating condition and up to 2300 K in accidental operating conditions, in helium pressurized at 7 MPa. An experimental approach was made on SiC f /PyC/SiC under active oxidation conditions (1400 ≤ T ≤ 2300 K; 0.2 ≤ pO 2 ≤ 2 Pa). The variation of the mass loss rates according to the temperature showed the existence of three domains. The oxygen partial pressure had a weak impact on the mass loss rate for temperatures lower than 2070 K. Beyond 2070 K, the mass loss rate increased leading to important damage of the material, at lower temperature under pO 2 = 0.2 Pa than under pO 2 = 2 Pa, which was associated to the effect of the oxygen partial pressure on the sublimation of SiC.

International audience; This paper describes an original method to measure the plutonium self-diffusion coefficient in the mixed oxide U0.55Pu0.45O2±x. This method is based on using 242Pu as a tracer atom. A thin film of the tracer was deposited on the well-polished surface of the samples and then diffusion annealings were performed from 1500°C to 1700°C, in an Ar-H2 5% atmosphere. The oxygen potential was fixed at-395 kJ.mol-1. After annealing, the 242Pu self-diffusion profiles were established by means of secondary ion mass spectrometry (SIMS). The 242Pu concentration profiles were determined by assessing the relative U, Am and Pu ionisation yields according to the experimental parameters. The choice of favourable experimental conditions and the relevance of the resulting concentration profiles are discussed at length

International audience; A method has been developed in order to assess small volume interdi ffusion coeffi cients from experimental Electron Probe MicroAnalysis concentration profi les of polycrystalline materials by means of Boltzmann-Matano or den Broeder methods and their complementary Hall method. These methods have been used as tools for the investigation of the quasi-binary UO2/U(1-y)PuyO(2-z) interdi ffusion, for which obtaining a solid solution in the bulk of grains is of major interest. In this paper uncertainties on the interdiff usion coefficient as a function of concentration have been computed for each method. Small volume coeffi cient measurements were enhanced by means of a small angle acquisition pro le line with respect to the interdiff usion interface.

International audience; For Generation IV reactors, and more particularly the Gas Fast Reactor, ternary carbide Ti3SiC2 is an interesting candidate for the application as fuel coating; actually, it has the advantage to combine some properties of metals with those generally attributed to ceramics. Unfortunately, few data are available on its behavior under irradiation. In this study, we attempted to measure and to understand the origin of the swelling induced by nuclear collisions. Thus, it seems that Ti3SiC2 irradiated at room temperature swell less than silicon carbide and that critical amorphization temperature is less than room temperature.

Comme pour tout système nucléaire, le concept du combustible pour les réacteurs de IVème génération consiste en des pastilles de combustible conditionnées dans une matrice qui doit contenir les produits de fission. De par leur excellente réfractarité, les carbures sont pressentis pour constituer cette barrière de confinement. Parmi ceux étudiés, le ternaire Ti3SiC2 se distingue par ses propriétés mécaniques particulières : en effet, sa structure nanolamellaire lui confère une certaine plasticité ainsi qu'une ténacité supérieure aux carbures classiques tels que SiC ou TiC. Cependant, son comportement sous irradiation n'est pas connu. Afin d'appréhender ce comportement, des échantillons ont été irradiés aux ions lourds de différentes énergies puis caractérisés. Le panel d'énergies utilisé a permis de discriminer l'effet des interactions nucléaires – pulvérisation – des interactions électroniques – apparition de monticules et dilatation de la maille cristalline.

International audience; The concept of the fuel for the IVth generation reactors should consist of fuel pellets surrounded with a matrix that must contain fission products. Thanks to their interesting thermo-mechanical properties, carbides are sensed to become this matrix. Among the studied carbides, Ti3SiC2 can be distinguished; actually, its nano-laminated structure confers to it some softness as well as a better toughness than classical carbides like SiC or TiC. However, before to use this remarkable carbide, a study of its behaviour under irradiation must be led. Thus, some characterisations were performed on 75 MeV Kr irradiated specimens. They allowed to underline that TiO2 (formed on the surface of Ti3SiC2 during the surface preparation) seems to be sputtered by irradiation, and that the unit cell of Ti3SiC2 is dilated along c axis.