[en] The aim of this work is the study, at a molecular scale, of the sorption mechanisms corresponding to UO₂²⁺ /TiO₂ system. The first step of the study is the hydrated solid characterization and the second step is the sorption mechanisms study. The work is performed using a multi-scale approach, which is composed by three parts. At first, we have obtained macroscopic data corresponding to the uranyl retention on the solid. Moreover, we have used CD-MUSIC model in order to calculate the surface acidity constants of the hydrated solid. In a second part, we have performed a multi-spectroscopic study (TRLFS, XPS, DRIFT, SHG) in order to characterize the system at a molecular scale. Finally, all these results were obtained on the solid TiO₂ under different forms (rutile or anatase phases and (110), (001), (111) rutile single crystals). From the structural results, we have determined that there are two sorption sites towards the free uranyl ion onto the titania surface, whatever the studied solid (powder and single crystal). Moreover, these two reactive sorption sites don't present the same reactivity towards free uranyl ion: the more reactive sorption site is assigned to a bridging-bridging oxygen atoms and the less reactive sorption site is assigned to a bridging-top oxygen atoms. Thus, using this methodology, we can conclude that it is possible to understand the powder surface properties using the single crystals surface results. Finally, with the multi-scale approach, we have used the spectroscopic results obtained on UO₂²⁺/TiO₂ system as constraints during macroscopic data simulation using a surface complexation model (CCM). (author)

[en] The solubility of ThO₂(cr) was studied since many years but a large discrepancy in solubility values is noticed in the literature. The present work suggests that this discrepancy is related to differences in the surface properties of thorium oxide. To understand the role of surface properties on solubility values, we conducted experiments with ThO₂(cr) spheres with reproducable surface properties. Batch dissolution experiments were conducted in 0.01 M NaCl solution at pH = 3.0 and 4.0 for periods of time up to 270 days. The solutions were spiked with ²²⁹Th to determine precipitation (sorption) rates of thorium, while dissolution rates were determined by measuring ²³²Th released from ThO₂(cr) spheres. We assume that ²²⁹Th atoms are exchanged only with active sites involved in Th-dissolution. Using ²²⁹Th as local sensor of attachment and detachment processes at the ThO₂(cr) surface under close-to-equilibrium conditions, allows to assess surface reactivity of the solid during solubility experiments. (orig.)

[en] Full text of publication follows. The radiolysis of water is the chemical process which follows the absorption of ionizing radiation in the water. So, this phenomenon occurs during almost all different steps of the radiotherapy. Indeed, from the production of radioisotope, the presence of water in different operations (cooling of solid target system, aqueous phase for separation and purification) leads to its radiolysis. Then, the radiation effects (due to the radioisotope) on the targeting ligands are really important because the radiolysis is involved onto the redox conditions and can perturb the good conjugation between radioisotopes and antibodies. Finally, the free radical and molecular products formed during the radiolysis of water are responsible in the destruction of tumoral cells. In this presentation, we will show how the knowledge of the radiation chemical yields of pure water can help the nuclear medicine scientists to understand and control the behavior of the radioisotope produced for the internal vectorized radiotherapy. Our study is focused onto the production of molecular hydrogen according to the Linear Energy Transfer (LET) during the radiolysis of pure water. Irradiation experiments are performed at the CEMHTI (Orleans, France) cyclotron and the ARRONAX (Saint-Herblain, France) cyclotron facility. We developed a new irradiation cell allowing working in track-segment LET conditions because results obtained can be compare with those of a Monte Carlo simulations program. First comparisons between experiments and simulations are really interesting [Ref.1]. Moreover, we performed irradiation at track-averaged LET. These conditions are closer than to those encountered in radiotherapy. The results show an increase of the radiation chemical yield g(H_2) with regard to the LET (0,23 to 150,0 keV μm"-"1) from 0,03 μmol*J"-"1 to 0,13 μmol*J"-"1 [Ref.2]. However, it is not now possible to compare these experimental results with Monte Carlo simulations due to the too expensive CPU time in order to take into account the high heterogeneity of the energy deposition. References: (1) Crumiere, F.; Study of the TEL effect during water radiolysis: radiolytic yields of molecular hydrogen, Doctorat Thesis, University of Nantes (2012); [2] Crumiere, F., et al., LET effects on the hydrogen production induced by the radiolysis of pure water, Rad. Phys. Chem., Vol. 82, pp. 74-79., (2013). (authors)

[en] Clay host rock and engineered barrier systems are the key elements in the concept adopted by several countries to isolate the high level nuclear waste from the biosphere. Mainly composed by illite, mixed-layer illite-smectite (I/S) and montmorillonite, these clays are characterized by their properties of high retention and low permeability for released radionuclides. After closure of the repository in deep geological formation, groundwater in equilibrium with the host rock, forming the pore water, will saturate the engineered barrier system and come in contact with the nuclear glass waste package. The leaching of the different silicate minerals as well as the uptake of dissolved silicic acid by these phases is an important factor in the dissolution of the nuclear glass waste. To understand how the Si interacts with clay materials, this study aims at evaluating the solubility as well as the dynamics of solid/solution exchange reactions, which control the dissolved silicic acid concentration in solution in contact with Callovo-Oxfordian clay-stone. The results were compared with the dissolution of illite, bentonite and quartz at 25, 50 and 90 deg. C in pore water. The experiments were conducted in batch system and in controlled atmosphere conditions and were followed in continue until the equilibrium between the solid and solution is reached. The results present a stabilization of the pH-values at 8.2 after 211 days for all the samples. The nature of solids and the temperature were not the determining factors on the pH-value but the chemical composition of the pore water and the working atmosphere (here, nitrogen). Dissolution and precipitation rates were calculated from the concentration of Si released from solids and the activity of ³²Si-radiotracer added in solution, respectively. Dissolution rates were in the range of (8.7±0.4) x 10^-12-(6.8±0.3)x10^-11 mol Si/m²/s for quartz, (1.6±0.1) x 10^-13-(6.4±0.3) x 10^-13 mol Si/m²/s for Callovo-Oxfordian clay-stone, (2.4±0.1) x 10^-13-(9.4±0.5) x 10^-13 mol Si/m²/s for illite du Puy and (1.2±0.6) x 10^-12-(9.1±0.5) x 10^-12 mol Si/m²/s for bentonite. Precipitation rates were in the range of (8.4±0.4)x10^-12-(2.0±0.1) x 10^-11 mol Si/m²/s for quartz, (2.0±0.1) x 10^-13-(2.5±0.1) x 10^-12 for Callovo-Oxfordian clay-stone, (2.4±0.1) x 10^-12-(5.2±0.3) x 10^-12 mol Si/m²/s for illite and (1.9±0.1) x 10^-13-(6.9±0.3) x 10^-13 mol Si/m²/s for MX80 bentonite. The plot of dissolution rates versus precipitation rates gave a slope near one indicating a dynamic process of dissolution/precipitation. Through the experiment with the addition of a spike of ³²Si-radiotracer in solution, we have showed that the activity of ³²Si decreases in contact with clay containing at least 6% of Fe and Al, as for example illite, which seems to be a necessary condition. Finally, in the condition of nuclear glass waste disposal, the Callovo-Oxfordian clay-stone would be the phase controlling the solubility of dissolved Si from the nuclear waste and clay with a solubility value of 4.8 x 10^-4 mol/L at 90 deg. C in pore water of Bure site. (authors)

[en] Document available in extended abstract form only. Long term disposal of high-level and long-lived nuclear waste has been extensively investigated for many years. Hydrogen will be generated by corrosion and radiolysis of waste packages with high level radioactive waste in presence of water. Several authors systematically studied the radiolytic decomposition of water and the results obtained were published in an important number of revue papers and books. These studies basically address the kinetic and thermodynamic parameters which quantify radiolytic yields and reaction rates of radiolysis products. Several studies of the water radiolysis in presence of clay and oxide have been reported. The yield of H₂ was found to be greater for the radiolysis studies of water adsorbed on different surface than for bulk water. The present work quantifies the radiolytic production of hydrogen in the presence the argillite sample K-119, nanoparticles of TiO₂ with specific surface area of 253 m²/g, and bentonite FoCa₇ using ⁶⁰Co source. The dose rate was 112, 400 and 1144 Gy/h as determined using the Fricke's dosimetry. The G(H₂) radiolytic yield, calculated for the argillite sample K119 containing 3% water is 32.3 10^-7 mol.j^-1 which is 100 times higher than that obtained from argillite water sample of to that 3.2 10^-8 mol.j^-1. This augmentation in yield appears to be due to electron and (.OH) radical interaction with the argillite sample K119. The G(H₂)-values for hydrogen gas production from titanium oxide anatase-type with specific surface area of 253 m²/g dispersed in water are is 0.43 10^-7 mol/j. This value two times higher than that obtained from?-radiolysis of deaerated, aerated water. The radiolytic yield for hydrogen, calculated for solid FoCa₇-clay containing 11% water is 5.22 10^-9 mol.j^-1 which is much lower than the G(H₂) value of 3.6 10^-7 mol.j^-1 obtained with the same material dispersed in water. The latter yield is similar to the one found in experiments studying the radiolysis of water confined in nanoscale pores of well-characterised porous silica glasses and meso-porous molecular sieves (MCM-41) reported by P. Rotureau et al. (authors)

[en] Full text of publication follows: Radionuclides retention mechanisms onto mineral phases is of primary importance for nuclear waste management. The aim of the presented study is to demonstrate that it is possible to predict the retention properties of a methodological powdery substrate from the study of its natural crystallographic orientations. Among the radionuclides of interest, U(VI) can be seen as a model of the radionuclides oxo-cations. The substrate under study is the titanium oxide (TiO₂). In fact, rutile can be found as powder and also as manufactured single crystal which allows to study the retention processes on perfectly known crystallographic planes. Since the repartition of the different crystallographic orientations are known for the powder, the results obtained for the single crystals can directly be used to account for the powder retention properties. By using combined spectroscopic techniques such as TRLFS, XPS, DRIFT and SHG, it is possible to determine the nature of the reactive surface sites and also the surface species. XPS and TRLFS measurements allowed to determine that two same uranyl surface species were formed on titania (110) and (001). Only, the relative intensities of these species vary with the surface coverage. Atomic Force Microscopy was carried out to verify that no surface precipitation occurs for the higher surface coverages. Moreover, these analysis have also evidenced that the U(VI) sorption is homogeneous. These observations were corroborated by SHG experiments (mainly for (001)) which have also shown that the sorption occurs, in a first step, onto preferential surface symmetry axis. For rutile powder, the preferential crystallographic orientations are (110), (100) and (101) in the ratio 60/20/20. TRLFS and XPS experiments have shown that two uranyl surface species are formed whatever the pH value ranged from 1 to 5. The spectroscopic characteristics of these species are the same as the ones observed on (110) and (001) planes. DRIFT experiments have been shown that the reactive oxygens remained protonated after uranyl sorption. For powders, potentiometric titration experiments were performed and were successfully modelled (Constant Capacitance Model) using CD-Music and the natural repartition of the crystallographic planes of titania. Thus, only the inner-capacitance value remained as an adjustable parameter. The sorption edges defined for powders were then fitted on the basis of the constraints brought by the spectroscopic investigation which allowed to perform the fit with a minimal number of adjustable parameters and then allowed a more accurate determination of the sorption constants values. (authors)

[en] Document available in extended abstract form only. The controls of silica concentrations in argillite pore water have significant bearing on potential mineralogical transformations with temperature and due to interaction with engineered barrier materials and waste matrices. A particularly important issue are Si concentration controls with respect to the dissolution of nuclear waste glass. A detailed study of Si solubility of various constituting minerals and of Callovo-Oxfordian (COX) clay stone samples has been conducted in the frame of ANDRA's laboratory consortium on interactions in the clay/steel/glass system. Solubility experiments were conducted at pH 7-8 for 400 d at 25, 50 and 90 deg. C for quartz, illite, MX80 bentonite and two COX samples of rather similar mineralogical composition from the C2b1 horizon. Reversibility at solubility equilibrium was checked by 32Si addition after achievement of solubility equilibrium. Time to reach apparent equilibrium is about 100 days. Si concentrations obtained in batch experiments at low clay/water ratios (between 10 and 20 g/ l) were similar to those obtained for the compact clay pore water collected by high pressure percolation experiments both at 25 and 90 deg. C. The obtained data were also similar to Si concentrations reported for field conditions. It is interesting to notice that apparent equilibrium concentrations of Si for clay rock samples vary by as much as a factor of 5 for rock samples which are less than a meter apart. This indicates that Si concentrations are controlled by local solubility equilibria. Experimental data can be described both by sorption constants and by solubility equilibrium. Rationalizing the data as sorption processes, both the number of exchanging surface sites and Kd values were obtained from isotopic exchange experiments. Such data are used in the literature to assess by modeling the effect of clay rock on nuclear waste glass dissolution. However, relative large parts (50 to 100%) of crystallographic Si surface sites participate in the exchange equilibrium for all minerals at 25 deg. C and hence, it is more likely that the observed dynamic exchange involves a whole monolayer on the surface not only on quartz, but as well on the edge sites of illite and bentonite clay. At high temperature (90 deg. C) the extent of Si isotopic exchange remains similar for bentonite but it increases for illite and quartz by a factor of 3 and for COX samples by as much as a factor of 10. For clay samples this cannot be described by a simple sorption constant but by solubility constraints involving surface precipitation and/or surface restructuring at edge sites. Solution concentrations of Si in case of clay rock samples are lower but close to quartz solubility at all temperatures. (authors)

[en] For various countries, the direct disposal of high level nuclear fuel wastes is a key option for the back-end of the fuel cycle. For HTR/VHTR reactors this is assumed for the introductory phase of this reactor system. However, closed fuel cycles or a separation of spent coated-particles from the graphite moderator and specific treatment, conditioning and disposal of these waste streams are also possible. In the European Community project 'RAPHAEL', fuel waste performance is going to be studied in depth, including post-irradiation fuel characterization, analysis of the stability and failure mechanism of coatings and of fuel kernels and overall performance of waste packages with compact fuel and/or only with fuel particles in geological disposal environments. Different confinement matrices for separated fuel particles (vitrification, SiC, ZrO₂) have been adapted to limit release of radionuclides into groundwater at low temperatures over geological time spans. The investigations are limited to Low-Enriched Uranium (LEU) fuel with uranium oxide and uranium oxycarbide kernels that will allow higher burn-up, but may be more susceptible to leaching. (authors)

[en] Full text of publication follows: One of the most important processes affecting safety during the storage of nuclear waste in underground disposal is the migration of radio-toxic elements through the geosphere. The major factor involved in the retention of these elements, in the case of water infiltration, would be their sorption at the water/mineral interface. In order to predict if a disposal will be safe, retardation process must be correctly understood. Among the radionuclides of interest, an experimental team of our laboratory is studying the interaction between uranyl ions (UO₂²⁺) and the TiO₂ (110) surface. Under experimental conditions, using XPS and TRLFS measurements two uranyl surface complexes were identified on the TiO₂ (110) surface and it was shown that only the free aquo uranyl ion reacts with the surface sites. The relative intensities of these two structures change with the surface coverage indicating two different stabilities. The purpose of this work is to provide a theoretical support to the experimental investigation. With this aim in view, we report first principles density functional calculations on the adsorption of water and uranyl ions on the TiO₂ (110) surface. Two different approaches were used. First, TiO₂ (110) surface was modeled with periodic slabs in order to investigate clean surface relaxations, water adsorption, and finally interaction of uranyl ions with the mineral surface. The goal of this first step was to identify the most probable UO₂²⁺ adsorption sites and to establish their relative energy stabilities as a function of the surface coverage. Then, starting from the previous TiO₂ (110) relaxed structure, (TiO₂)_n clusters were used to simulate the (110) rutile surface. This second step was done in order to study, from a local point of view, the interaction of one uranyl ion with the TiO₂ (110) surface for the different possible adsorption sites. Each theoretical result is directly compared with the experimental data. This original theoretical study correlated to the experimental investigation leads to an acute representation of the retention processes of radionuclides on mineral surfaces. (authors)

[en] In the frame of the former French 1991-law on waste management, which was extended in 2006-law, hollandite ceramic was studied as a potential specific conditioning matrix for caesium isotopes (long-life radionuclide ¹³⁵Cs and the strong heat generating radionuclide ¹³⁷Cs). In this general study of Cs-containment in a ceramic matrix, the chemical durability was pointed out as a key property. Leaching experiments in static mode were conducted during 240 days at various pH-values from acidic to alkaline range. The initial leaching rates between 0 and 45 days are faster for Cs than for Ba and the average for the caesium are (1.4 ± 0.1) × 10⁻⁴ g/m²/d (pH 2.5), (6.4 ± 0.9) × 10⁻⁵ g/m²/d (pH 4.4) and (3.1 ± 0.6) × 10⁻⁵ g/m²/d (pH 8.6), and for the barium (6 ± 1) × 10⁻⁵ g/m²/d (pH 2.5), (2.8 ± 0.3) × 10⁻⁵ g/m²/d (pH 4.4), and (2 ± 2) × 10⁻⁶ g/m²/d (pH 8.6). At the equilibrium between 45 and 240 days, the normalised mass losses average for caesium are (8.2 ± 0.3) × 10⁻³ g/m² (pH 2.5), (5.2 ± 0.4) × 10⁻³ g/m² (pH 4.4) and (4.1 ± 0.2) × 10⁻³ g/m² (pH 8.6), and for barium (3.7 ± 0.4) × 10⁻³ g/m² (pH 2.5), (2.0 ± 0.1) × 10⁻³ g/m² (pH 4.4) and (4 ± 2) × 10⁻⁴ g/m² (pH 8.6). Caesium and barium are incongruently released in solution with a correlation slope close to 0.5 at pH 2.5 and pH 4.4 and very low (near 0.02) in alkaline solution. Sorption experiments with radioactive isotopes (¹³⁷Cs and ¹³³Ba) were conducted on hollandite pre-leached in aqueous solutions. Caesium and barium release is controlled by the surface reactions. Leaching experiments and isotopic addition experiments (¹³⁷Cs- and ¹³³Ba-radiotracer) indicate that caesium behaviour is independent on pH-values, whereas barium behaviour is strongly dependent. Additional experiments in the presence of gamma irradiation (⁶⁰Co source) did not show any significant effect on hollandite leaching process.