[en] In this work we have investigated the interaction of magnetite with cesium, strontium, molybdenum and selenium, in the frame of radionuclide retention by canister corrosion products. For each radionuclide, the retention on magnetite has been studied as a function of pH and the mass/ volume ratio. The experimental results have been modeled by means of Surface Complexation Models (SCM), that constitute a tool that allows an approach to sorption mechanisms in a wide range of experimental conditions taking into account electrostatic interactions at the mineral-water interface.(Author)

[en] The behaviour of the spent nuclear fuel in the conditions expected in a deep geological repository depends on some basic processes such as the formation of oxidizing species due to the radiolysis of water, the effect of these species on the spent nuclear fuel matrix oxidation and dissolution, and the precipitation of secondary solid phases. In this work, some of those processes have been studied in detail. In particular, both uranium dioxide oxidation and dissolution rates have been determined in the presence of oxygen an hydrogen per-oxide (the most important molecular species formed in the radiolysis of water). In addition, the precipitation of secondary phases on the UO₂ surfaces in the presence of hydrogen peroxide has also been fool owed by means of the Atomic Force Microscope allowing the identification of studies (UO₄.H₂O). On the other hand, the radiolytical models of the dissolution of spent nuclear fuel need to estimate the surface site densities of the solid. In this sense, we have studied surface site densities of three different uranium oxides: UO₂, U₃O₈, and UO₃.4H₂O. (Author) 42 refs

[en] In this work we made a review on the different models and mechanisms that have been developed by different authors to explain the dissolution of spent nuclear fuel under oxic conditions. In most cases the oxidizing reagent used has been the molecular oxygen, but also some works with hydrogen peroxide or even with hypochloric acid can be found. Leaching experiments have been carried out with different types of spent nuclear fuel as well as with either chemical or natural analogues such as non irradiated uranium dioxide or natural uraninites, respectively. In oxygen and in the absence of bicarbonate ion, the data found in literature can be fitted considering the two-step oxidative dissolution mechanism developed by Torrero et al. (1998). This mechanism is able to explain the different reaction orders for pH oxygen concentration obtained depending on the experimental conditions. In the presence of bicarbonate, the data can be fitted considering the mechanism described de Pablo et al. (1999), which consists on two different steps: (1) oxidation of the surface of the solid and (2) surface co-ordination of the bicarbonate ion and dissolution of the complex formed. This model allows to explain different reaction orders for bicarbonate and oxygen concentration obtained by different authors. The development of a mechanism of UO₂ oxidation and dissolution in the presence of hydrogen peroxides is much more complied than in the case of oxygen because of the decomposition of the hydrogen peroxide, which is probably catalysed by the UO₂(s). At present, more work is being directed to the elucidation of this mechanism, including the study of the influence of some radicals such as OH on the UO₂ dissolution. (Author)

[en] This presentations is mainly based on the electrochemical studies carried out by the Canadian team and the research group of the Berlin University. Electrochemical studies allow to study separately both the anodic reaction which corresponds-sources on UO₂-electrodes response is one of to the UO₂ dissolution and the cathodic reaction that is the reduction of the oxidants. By using intensity current-potential plots a mechanisms of UO₂ corrosion has been established. At-300 mV (vs SCE), irreversible oxidation of UO₂ takes place and dissolution begins. In the absence of complexing agents like carbonate, an oxidised layer is formed at 100 mV a stoichiometry close to UO₂. In carbonate medium, the oxidized layer is not formed because the U(VI) formed is rapidly dissolved. Results in terms of dissolution rates obtained by electrochemical measurements are similar to the ones obtained in dissolution experiments by using flow through reactors and similar kinetic laws are obtained. The effect of external α and γ-sources on UO₂-electrodes response is one of the few available data on the effects of radiolysis on the UO₂ dissolution rate and can offer a complementary knowledge to the spent fuel and α-doped pellets dissolution experiments. (Author)

[en] The dissolution rate of unirradiated uranium dioxide was studied in batch experiments as a function of hydrogen peroxide concentration (from 10^-5 to 10^-3 mol dm^-3). Unirradiated UO₂(s) was used in order to differentiate surface chemical processes from radiolytic effects. Dissolution rates were determined from both uranium release and hydrogen peroxide consumption. Results showed that H₂O₂ consumption rate was higher than UO₂ dissolution rate. This observation may indicate that the overall UO₂ oxidative dissolution process would be controlled by the dissolution of the oxidized solid surface. The calculated hydrogen peroxide reaction order was 1 in the H₂O₂ concentration range from 10^-5 to 10^-4 mol dm^-3, while at higher concentrations no clear dependence was observed. Copyright (2001) Material Research Society

[en] Full text of publication follows: Hydrogen peroxide is one of the main molecular oxidants produced by water radiolysis by the spent nuclear fuel radiation. In previous works, we have studied the effect of this oxidising agent on the UO₂ in order to develop an oxidation/dissolution mechanism under conditions relatively close to the ones expected in a deep geological repository. However, the effect of parameters such as temperature and pressure, which may have a significant effect in the whole mechanism, has not been studied. In the present work we have developed and used a continuously stirred tank flow-through reactor to determine the kinetics of dissolution of UO₂ in the presence of H₂O₂ (from 1 x 10^-5 to 8 x 10^-5 mol dm^-3) and as a function of hydrostatic pressure (from 1 to 32 bar) and temperature (from 20 to 50 deg. C). In the output flow we measured both uranium and hydrogen peroxide concentrations until the system reached a steady state condition. At that point, we used the uranium concentrations, the flow rate and the total area of solid inside the reactor to calculate the rates of dissolution (mol x m^-2 x s^-1). The results obtained have been evaluated using a multi-parametric analysis. The equation obtained, fitted reasonably well the data obtained at different experimental conditions. In addition, the dissolution rates predicted by that equation have been compared with some values obtained in our laboratory in both continuous an batch studies. When the experimental conditions of those studies were entered in the multi-parametric equation, the predicted values calculated were in good agreement with the experimental ones. Finally, an activation energy of 35 ± 9 kJ x mol^-1 was determined for the temperature range studied. This value falls in the range of energies found in the literature for spent nuclear fuel, non-irradiated uranium dioxide and uraninites (20-60 kJ x mol^-1). (authors)

[en] Full text of publication follows: In order to improve the knowledge of the influence of phosphate on the spent nuclear fuel alteration, Scanning Force Microscopy (SFM) has been used in this work to detect the formation of secondary solid phases on UO₂ in contact with phosphate-containing solutions. In this sense, two different experiments were carried out using a batch methodology. Discs of non irradiated UO₂ (0.77 g and 2.3 x 10^-4 m² geometric surface area) were put in contact with leaching solutions with two different phosphate concentrations (10^-4 and 10^-6 mol x dm^-3) at pH near 7. The experiments were carried out in the presence of oxygen at room temperature. Periodically, the discs were taken off of the reactors and analysed by SFM using a Nanoscope IIIa apparatus. In addition, uranium concentration in solution was determined by ICP-MS. The initial UO₂ dissolution rates have been determined from the variation of the uranium concentration in solution with time and considering the specific surface area and the weight of solid. The values are 2.4 x 10^-10 and 1.2 x 10^-9 mol x m^-2 x s^-1 in the experiments with 10^-4 M and 10^-6 M total phosphate concentration, respectively. Time-Resolved Laser Fluorescence Spectroscopy (TRLFS) has been used to establish the uranium speciation in solution. With this technique, two different uranyl-phosphate complexes have been found in solution: UO₂PO₄ - and UO₂HPO₄(aq). SFM determinations of the UO₂ surface have shown the formation of a secondary solid phase in both experiments. This phase is detected after 3 days in the experiment with the highest phosphate concentration and after 20 days in the experiment with the lowest. In addition, the roughness of the surface changes in both experiments corroborating the precipitation of a secondary solid phase. Following the methodology of Clarens et al. [1], the precipitation rate of this secondary phase is determined. [1] Clarens, F.; de Pablo, J.; Diez-Perez, I.; Casas, I.; Gimenez, J.; Rovira, M. (2004) The formation of studtite during the oxidative dissolution of UO₂ by hydrogen peroxide: A SFM study. Environ. Sci. Technol. 38, 6656-6661. (authors)

[en] In this report the results of the experimental work carried out within the collaboration project between ITU-ENRESA-UPC/CTM on spent fuel (SF) covering the period 2005-2007 were presented. Studies on both RN release (Fast Release Fraction and matrix dissolution rate) and secondary phase formation were carried out by static and flow through experiments. Experiments were focussed on the study of the effect of BU with two PWR SF irradiated in commercial reactors with mean burn-ups of 48 and 60 MWd/KgU and; the effect of High Burn-up Structure (HBS) using powdered samples prepared from different radial positions. Additionally, two synthetic leaching solutions, bicarbonate and granitic bentonite ground wa ter were used. Higher releases were determined for RN from SF samples prepared from the center in comparison with the fuel from the periphery. However, within the studied range, no BU effect was observed. After one year of contact time, secondary phases were observed in batch experiments, covering the SF surface. Part of the work was performed for the Project NF-PRO of the European Commission 6th Framework Programme under contract no 2389. (Author)

[en] If a failure of the spent nuclear fuel container is assumed to occur during the first 1000 years after the final disposal, ss radiation has to be considered in the conceptual model. However, the influence of ss radiation separated from α or γ radiation on the dissolution of the spent nuclear fuel has not been extensively studied. The study of the effect of ss radiation on the UO₂ dissolution will help to understand the results obtained in spent fuel leaching experiments and might be useful to improve the databases of the radiolytical models. For this reason, we have studied the dissolution of non irradiated uranium dioxide, as a chemical analogue of the spent nuclear fuel, under external ss irradiation using a sequential methodology. (orig.)

[en] The SFS (Spent Fuel Stability) project that is being carried out as part of the European Union's 5th Framework Programme has a dual objective, technical and sociologic. The technical objectives consists of developing a model of the behaviour of irradiated fuel and of the radionuclides contained therein, under the conditions of a deep geological disposal facility, incorporating the experimental part of this and previous. European projects. The sociological objectives is to develop a common scientific and technical opinion throughout the European Union, for consensus to be reached regarding the evolution of a deep geological disposal facility for high level wastes. With a view to achieving this dual objective, and as a project activity, a Seminar was organised in Avila in June 2002 (the presentations made of this Seminar will be the subject of another publication), the aim being to establish the bases for a new spent fuel behaviour model with and ample experimental basis and the consensus of the European countries participating in the project (France, Switzerland, Germany, Sweden, Belgium and Spain. (Author)