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International audience; In the frame of Research Reactors Spent Fuel (RRSF) treatment by hydrometallurgy, the dissolution in nitric acid of irradiated U-Al, is a key issue because of the low solubility of molybdenum fission product in presence of high concentration of aluminium. In this study, the values of molybdenum solubility have been accurately measured in different operating conditions. Studies have carried out with non-active materials. To be more representative of metallic fuel, uranium-molybdenum alloy powder and molybdenum metal have been dissolved in aluminium nitrate solutions at high temperature. In order to be sure that molybdenum solubility has been reached, experiments have been carried out with an excess of molybdenum metal. In spite of this excess addition, metallic elements have been dissolved completely after stirring time of thirty minutes with a magnetic stirrer. Shortly after this total dissolution, a slow molybdenum precipitation has been observed for almost 15 hours. An experimental protocol has been developed to properly wash precipitates in order to determine their elemental composition. No uranium has been detected in the washed precipitate by ICP-AES measurements performed after redissolution of solids in aluminium free nitric acid solutions. Further analyses by Scanning Electron Microscope have shown a needle-like morphology. Energy-dispersive X-ray spectroscopy analyses on several selected areas have confirmed the absence of uranium in precipitates. EDX semi-quantification has been carried out on ionically polished particles. They are composed of 75% oxygen and 25% molybdenum, suggesting MoO$_3$ compounds. X-ray diffraction spectra of powders have confirmed this result: all samples matched the crystallographic form of MoO$_3$ .

International audience; Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectrometry, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality-safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remaining after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction in stainless steel containers. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations n°1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station n°1 allows determining the burn-up of the irradiated fuel by gamma-ray spectrometry with HP Ge (high purity germanium) detectors. The burn-up is correlated to the $^{137}$Cs and $^{134}$Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station n°3 is dedicated to the control of the correct fuel dissolution, which is performed with a $^{137}$Cs gamma ray measurement with a HP Ge detector. Station n°7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional $^3$He detectors. So far, large campaigns of reprocessing of the UOX fuels with a burn-up rate up to 60 GWd/t have been performed at AREVA/La Hague. This paper presents a brief overview of the current status of the nuclear measurement stations

International audience; A broad variety of methods for measurement error (ME) correction have been developed, but these methods have rarely been applied possibly because their ability to correct ME is poorly understood. We carried out a simulation study to assess the performance of three error-correction methods two variants of regression calibration (the substitution method and the estimation calibration method) and the simulation extrapolation (SIMEX) method. Features of the simulated cohorts were borrowed from the French Uranium Miners' Cohort in which exposure to radon had been documented from 1946 to 1999. In the absence of ME correction, we observed a severe attenuation of the true effect of radon exposure, with a negative relative bias of the order of 60% on the excess relative risk of lung cancer death. In the main scenario considered, that is, when ME characteristics previously determined as most plausible from the French Uranium Miners' Cohort were used both to generate exposure data and to correct for ME at the analysis stage, all three error-correction methods showed a noticeable but partial reduction of the attenuation bias, with a slight advantage for the SIMEX method. However, the performance of the three correction methods highly depended on the accurate determination of the characteristics of ME. In particular, we encountered severe overestimation in some scenarios with the SIMEX method, and we observed lack of correction with the three methods in some other scenarios. For illustration, we also applied and compared the proposed methods on the real data set from the French Uranium Miners' Cohort study. © 2012 John Wiley and Sons, Ltd.

International audience; Radiation-catalysis is a well-known process leading to H2 production through radiolysis of adsorbed water on oxides. In this article, we show that common, easily accessible, hydroxides can be as much efficient for H2 production as the more efficient oxide identified until now. H2 radiolytic yields were determined from the same nanostructured hydrated samples that differ by their particle size (AlOOH L and AlOOH S for large and small particle size, respectively). The measured yields are of the order of 10−8 mol J−1. It means that such an irradiated material produces more efficiently H2 than an equivalent mass of water. H radicals, trapped electrons (F centers), and related O− centers were identified by electron paramagnetic resonance (EPR), at room and low temperature. Adsorbed water seems to play a role in the secondary processes of radical recombination, allowing a very efficient H2 production in these systems. This raises interesting questions about the energy transfer mechanism underlying this efficient hydrogen production and provide design lines for the design of efficient radiation-catalytic materials for H2 production.

International audience; In this study, we assessed the efficacy of unilamellar 110-nm liposomes encapsulating the chelating agent diethylenetriaminepentaacetic acid (DTPA) in plutonium-exposed rats. Rats were contaminated by intravenous administration of the soluble citrate form of plutonium. The comparative effects of liposomal and free DTPA at similar doses were examined in terms of limitation of alpha activity burden in rats receiving various treatment regimens. Liposomal DTPA given at 1 h after contamination more significantly prevented the accumulation of plutonium in tissues than did free DTPA. Also, when compared to free DTPA, liposome-entrapped DTPA was more efficient when given at late times for mobilization of deposited plutonium. In addition, repeated injections of liposomal DTPA further improved the removal of plutonium compared to single injection. Various possible mechanisms of action for DTPA delivered through liposomes are discussed. The advantage of liposomal DTPA over free DTPA was undoubtedly directly and indirectly due to the better cell penetration of DTPA when loaded within liposomes, mainly in the tissues of the mononuclear phagocytic system. The decorporation induced by liposomal DTPA may result first from intracellular chelation of plutonium deposited in soft tissues, predominantly in the liver. Afterwards, the slow release of free DTPA molecules from these same tissues may enable a sustained action of DTPA, probably mainly by extracellular chelation of plutonium available on bone surfaces. In conclusion, decorporation of plutonium can be significantly improved by liposomal encapsulation of DTPA regardless of the treatment regimen applied.

International audience; The thermal decomposition of uranyl peroxide tetrahydrate, (UO2)O2(H2O)2·2H2O, was studied by combining high temperature powder X-ray diffraction, scanning electron microscopy, thermal analyses and spectroscopic techniques (Raman, IR and 1H NMR). In situ analyses reveal that intermediates and final uranium oxides obtained upon heating are different from that obtained after cooling at room temperature and that the uranyl precursor used to synthesize (UO2)O2(H2O)2·2H2O, sulfate or nitrate, has a strong influence on the peroxide thermal behavior and morphology. The decomposition of (UO2)O2(H2O)2·2H2O ex sulfate is pseudomorphic and leads to needle-like shaped particles of metastudtite, (UO2)O2(H2O)2, and UO3-x(OH)2x·zH2O, an amorphous phase found in air in the following of (UO2)O2(H2O)2 dehydration. (UO2)O2(H2O)2·2H2O and the compounds resulting from its thermal decomposition are very reactive towards hydrofluorination as long as their needle-like morphology is kept.

International audience; Uranium dioxide containing 3 mol.% platinum group metals (PGMs) (Ru, Rh, Pd) was synthesized by hydroxide precipitation. The powders were converted to oxides, pelletized and sintered to prepare dense pellets of UO 2 incorporating PGMs particles. The characterization techniques performed revealed a microstructure similar to that of spent nuclear fuel (SNF). Dissolution tests in nitric acid demonstrated that in the presence of PGMs, the uranium dissolution rate was increased and the induction period was shortened. We used a new method based on the acquisition of Environmental Scanning Electron Microscopy (ESEM) micrographs recorded at three tilt angles and at different dissolution times. This method allowed the reconstruction of the topography of the solid/liquid interface. By monitoring the evolution of the solid/liquid interface during dissolution by means of 3D reconstructions, we were able to observe preferential dissolution zones in the vicinity of the PGMs particles and to determine microscopic dissolution rates for several regions of interest. PGMs particles were found mainly at the grain boundaries. In 0.1 M HNO 3 solution at 60˚C, the normalized dissolution rate for uranium at the grain boundaries reached R L (U) = (7 ± 1)×10-2 g.m-2 .d-1 , a value similar to the normalized dissolution rate determined for the whole image over the first 30 days of the experiment. This result showed that the dissolution occurred mainly at the UO 2 grain boundaries in the vicinity of PGMs particles. Furthermore, the 3D reconstructions of the solid/liquid interface were used to determine the evolution of the surface area of the pellet. By combining the weight losses determined at the macroscopic scale using the uranium concentrations in solution with the reactive surface area values, it was possible to estimate an effective normalized dissolution rate for the whole pellet.

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Samples of butanal oxime in aqueous solution have been irradiated with the helion ( 4 He 2+ ) beam of the ARRONAX (Nantes) and the CEMHTI (Orleans) cyclotrons. The consumption yield of butanal oxime has been measured by gas-chromatography coupled with mass spectrometry. Yields of gaseous products (mainly H 2 ) have also been measured by micro-gas-chromatography. Butanal oxime can react with H ∙ radicals by abstraction mechanism to enhance H 2 production. Yields of liquid phase products (hydrogen peroxide and nitrite ion) have been measured by colorimetric methods. Butanal oxime acts as a scavenger of OH ∙ radical to inhibit the production of H 2 O 2 . The observation of the radiolytic products allows then to discuss a degradation mechanism of butanal oxime in aqueous solutions.

International audience; External lubrication is often used to complete compaction process of powder materials. The main goal of this method is generally to reduce the amount of admixed internal lubricant (Zinc stearate powder) within the raw material. The application of external lubricants enhances the density uniformity and the mechanical strength of the resulting compacts. This study investigates the effects of the external lubricant amount for UO$_2$ powder compaction and the properties of the corresponding green pellets (corresponding to the compacts before sintering) without any admixed lubricant in the raw powder in order to evaluate the feasibility of this route in the case of nuclear powder. Results show that there is a quantity or number of layers from which the external lubricant applied on the die wall becomes detrimental to the friction index and the ejection force measured during the pressing cycle. The quality (surface defects, mechanical strength) of the green pellets can also be affected by the amount of lubricant. Thus the quantity and the thickness of the die wall lubricant must be optimized in order to assure an efficient mixed lubrication mode corresponding to the better lubrication mode in our study case.

International audience; The present article reviews a selection of results obtained in the AREVA/CNRS/UCA joint research laboratory. It focuses on interfaces formed by uranium hexafluoride (UF6) with chemical filter (purification), carbon (UF6 storage), and metallic substrate (corrosion). As a matter of fact, along the nuclear fuel cycle, metallic surfaces of the fluorination reactors, cooling systems (for the liquefaction of UF6), and storage containers are in contact with UF6, either in the gas or in the liquid phase. For the removal of volatile impurities before the enrichment, surface of chemical filters with a high specific surface area must be enhanced for both selectivity and efficiency. To store depleted UF6 (238U), graphite intercalation compounds are proposed and preliminary results are presented.

International audience; In order to optimize decommissioning waste management and better anticipate related issues, French nuclear waste producers CEA and AREVA, and the French Agency in charge of radioactive waste disposal (ANDRA) have decided to collectively develop technologies mastering costs, schedules and dose uptake as well as optimizing waste volumes and repository safety. In this context, several innovative projects have emerged over the past few years to develop specific thermal treatment processes. This paper focus on four new processes, presenting their goals, technical descriptions and first experimental results.

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International audience; UO2 pellets were prepared by densification of oxides obtained from the conversion of the oxalate precursor. Then characterized in order to perform a multiparametric study of the dissolution in nitric acid medium. In this frame, for each sample, the densification rate, the grain size and the specific surface area of the prepared pellets were determined prior to the final dissolution experiments. By varying the concentration of the nitric acid solution and temperature, three different and successive steps were identified during the dissolution. Under the less aggressive conditions considered, a first transient step corresponding to the dissolution of the most reactive phases was observed at the solid/solution interface. Then, for all the tested conditions, a steady state step was established during which the normalised dissolution rate was found to be constant. It was followed by a third step characterized by a strong and continuous increase of the normalised dissolution rate. The duration of the steady state, also called “induction period”, was found to vary drastically as a function of the HNO3 concentration and temperature. However, independently of the conditions, this steady state step stopped at almost similar dissolved material weight loss and dissolved uranium concentration. During the induction period, no important evolution of the topology of the solid/liquid interface was evidenced authorizing the use of the starting reactive specific surface area to evaluate the normalised dissolution rates thus the chemical durability of the sintered pellets. From the multiparametric study of UO2 dissolution proposed, oxidation of U(IV) to U(VI) by nitrate ions at the solid/liquid interface constitutes the limiting step in the overall dissolution mechanism associated to this induction period.

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International audience; Phosphate rocks are widely exploited for the manufacturing of phosphoric acid and fertilizers but they contain uranium (30-300 ppm). Therefore, recovering this uranium would enable the decontamination of phosphoric acid while valorizing uranium for the nuclear industry. Following a renewal of interest for the recovery of uranium from phosphate ores in the past few years, new extractant molecules were investigated in order to develop a new solvent extraction process. An amidophosphonate, the butyl-1-[N,N-bis-2-ethylhexyl)carbamoyl]nonyl phosphonic acid (DEHCNPB), showed good uranium extraction efficiency while meeting U/Fe decontamination requirements (as demonstrated during pilot scale trials). Afterwards, DEHCNPB was also used for the extraction of uranium from conventional resources (from sulfuric lixiviation media). However, pilot scale trials showed poorer performances, as uranium leaks in raffinates were higher than expected. The objective here is to study uranium and iron extraction mechanisms from those two different media (phosphoric and sulfuric). Thermodynamical data were acquired such as extraction isotherms, slope analysis, phosphates/ sulfates and water extraction. These data showed different behaviors depending on the initial medium. Spectroscopic techniques such as FTIR, NMR, ESI-MS and EXAFS were also investigated to study uranium-DEHCNPB complexes formed in the organic phase, enabling the determination of stoichiometries and coordination modes

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International audience; Modelling of precipitation reactions requires the knowledge of flow pattern inside the reactor, the thermodynamic properties as well as the kinetic laws of the different mechanisms (nucleation, crystal growth, agglomeration). Thanks to this information, population balance can be solved in support of design and operating of industrial process.Precipitations performed in fluidized bed reactor are commonly used, as fluidized bed leads to solid particles of excellent properties, in terms of morphology, granulometry or density. In this study we propose a new simple model to describe precipitations carried out in continuous fluidized bed reactor equipped with a mechanical stirring the Mixed Suspension Separated Product Removal (MSSPR). Inspired by the classical Mixed Suspension Mixed Product Removal (MSMPR), this model was applied to the precipitation of uranium peroxide. This latter is involved in the processing of uranium ores to recover uranium from purified leach solution as yellow cake

Samples of butanal oxime in aqueous nitric acid solutions have been irradiated with the helium ion (4He2+) beam of the CEMHTI (Orleans, France) cyclotrons. The consumption yield of butanal oxime has been measured by gas-chromatography coupled with mass spectrometry. Gaseous products (mainly H2 and N2O) have also been monitored by micro gas-chromatography. Yields of liquid phase products (hydrogen peroxide and nitrous acid) have been determined by colorimetric methods. The influence of nitric acid on the radiation chemical behavior of butanal oxime depends on the nitric acid concentration. For a low concentration (≤0.5 mol L-1) butanal oxime is protected by the nitrate ions which can efficiently scavenge the water radiolysis radicals. For higher concentrations, nitrous acid can accumulate in the medium, therefore leading to a strong increase of the butanal oxime degradation. The associated mechanism is an auto-catalytic oxidation of butanal oxime by HNO2.

International audience; In the scope of fast reactor spent fuel recycling, head-end research and development are performed at the Atalante facility in Marcoule. Considering the initial plutonium content, quantitative plutonium recovery is one of the main objectives for the dissolution process. In addition, the quantities of undissolved residue increase with the burn up and can impact the waste conditioning downstream. A silver (II) oxidizing digestion step was studied to assess its application to the treatment of undissolved residues containing eventually some plutonium. This process was first optimised on dioxide plutonium powders, and then tested on irradiated LWR MOX fuel residues. More recently it was applied to the solid residue obtained after dissolving in nitric media a MOX fuel irradiated in the Phenix Sodium Fast Reactor (SFR). In light of past experimentations and in order to obtain new basic data, not the whole spin was dissolved but only separate sections linked to known local burn-ups (BU). The first objective was to better correlate the quantity/composition of the dissolution residues with the local BU and Pu content of the initial irradiated material. Then the digestion step was applied on each dissolution residue obtained from each fuel pin part (bottom, medium i.e. full-flux zone, upper) with a view to evaluating the complementary Pu recovery, studying key parameters and characterising secondary residues.The digestion step permitted to recover up to 99% of residual plutonium with some slight differences depending on the position of the pin part the dissolution residue was obtained from. Oxidation conditions, local burn-up and chemical composition were found to be influential. Quantity of residues after digestion was significantly reduced thanks to this digestion treatment. The ultimate residues consisted mainly of metallic compound like ruthenium, molybdenum, rhodium or palladium.

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International audience; Transuranic wastes are produced by the nuclear fuel cycle management, essentially from MOX fuel production units. These waste, contaminated with Uranium and Plutonium, have large composition variability and enclose, in current temporary conditioning, mixed metals and organics. Metals are mainly stainless steel, aluminum or copper and organics are composed of different kind of plastics, highly chlorinated for some.On purpose, a new treatment process is being studied to incinerate the organics, melt the metallic fraction and incorporate the mineral residue fraction in a glass matrix. This paper presents the outlines of the process feasibility studies waste main characteristics, process specifications and main technical options, RetD strategy, mixed glass / metal package description, basic studies needed to develop the process and the matrix (glass/metal redox reactions, actinides partitioning )

International audience; Calcination is an efficient process for volume reduction and stabilization of nuclear liquid wastes. It is often the first step of waste containment processes such as vitrification or grouting.The CEA (French Alternative Energies and Atomic Energy Commission) and AREVA have acquired a high-level experience in the field of calcination through more than forty years of R&D and industrial operation. During this period, a broad range of liquid wastes have been studied and treated.R&D work and associated engineering studies allowed defining the design and sizing of the equipment and the process control parameters to be applied for the calcination of all kinds of solutions.The calcination process has been qualified for the following effluents amongst others:Liquid waste from UOx fuel reprocessingLiquid waste from UMo fuel reprocessing (high molybdenum and phosphorus contents)Hanford HLWLiquid waste from D&D operation (sodium bearing waste)This paper presents several studies led by the Joint Vitrification Laboratory (L.C.V), a common research laboratory between CEA and AREVA in charge of qualifying new processes and matrices for waste containment. These studies allowed defining the calcination rules, in particular for very hard to process solutions such as effluents with high boron, molybdenum and sodium contents.The feedback from industrial operation of the calcination process with fission products and effluents from D&D operations, highly loaded with sodium, is also presented.

International audience; The analysis of the DOUBLON experiment has provided C/E's for several isotopic ratios in the first and second rows of the radial blanket after an irradiation in the PHENIX Sodium Fast Reactor. We have used this as a basis for the validation of the DARWIN-2.3 (Tsilanizara et al., 2000) package for the depletion calculation in the radial blankets of SFR's (Lebrat et al., 2015). Unfortunately, no measurements have been performed in the upper and lower axial blankets of the fissile assemblies. Of course we expect our calculations to be reliable in these zones as well, as the initial compositions, nuclear reactions and neutron spectra are very similar. The representativity theory appeared to be the appropriate tool to correlate the available experimental information in the radial blanket to our calculations in the axial zones. Indeed, the representativity factors that we calculate between the final $^{239}$Pu/$^{238}$U ratio in the radial and axial fertile zones are very close to 1. Hence, the measurements performed in the radial zone can help us ''correct" the calculation in the axial blanket and reduce its a-priori uncertainty due to nuclear data. The estimated uncertainty reduction can reach a factor 5, since the experimental uncertainty is low. The $^{240}$Pu/$^{238}$U ratio is studied with a similar method.

Ruthenium extraction by tri-n-butylphosphate (TBP) from nitric acid was studied and modelled in the conditions of the PUREX process. Experimental distribution ratios obtained for water, nitric acid and ruthenium were described with a physicochemical model based on the application of the mass action law on each extraction equilibrium and by taking into account deviations from thermodynamic ideal behaviour both in aqueous and organic phase using Mikulin and Sergievskii-Dannus equations. The best agreement between experimental and calculated ruthenium extraction isotherms was obtained by considering the formation of two complexes TBP 2 RuNO ( N O 3 ) 3 ( H 2 O ) 2 and TBP 3 RuNO ( N O 3 ) 3 ( H 2 O ) 2 in organic phase. This model was implemented in the CEA-AREVA PAREX process simulation code and used to simulate ruthenium behaviour in a counter-current PUREX hot test performed in mixer-settlers.

Magnesium alloys (Mg-0.5%Zr and Mg-1.2%Mn) and graphite from spent nuclear fuel that has been used in the former French gas cooled reactors, have been stored together in AREVA La Hague plant. The recovery and packaging of these wastes is currently studied and several solutions are under consideration. One of the developed solutions would be to mix these wastes in a grout composed of industrially available cement, e.g. OPC, OPC blended with blast furnace slag or aluminous cement. Within the alkaline pore solution of these matrixes, magnesium alloys are imperfectly protected by a layer of magnesium hydroxide (Mg(OH)2, Brucite) resulting in a slow process of corrosion releasing hydrogen. As the production of this gas must be considered for the storage safety, and the quality of wasteform, it is important to select a cement matrix capable of lowering the corrosion kinetics of magnesium alloys. This is especially true when magnesium alloys are conditioned together with graphite wastes. Indeed, galvanic coupling phenomena may increase early age corrosion of the mixed wastes, as magnesium and graphite will be found in electrical contact in the same electrolyte. Many types of cements have been tested and most of them have caused strong hydrogen production when magnesium alloys and graphite are conditioned together into such cement pastes. Exceptions are geopolymer binder which is already known for that and another binder based on alkali hydroxide activated ground granulated blast furnace slag (AHABFS) which is presented in the present article. First are presented hydrogen production experiments that demonstrate the efficiency of AHABFS towards reduction of corrosion of Mg alloys embedded. In a second part, a formulation of fluid mortar based on this binder is proposed.