[en] One of the main processes that can control radionuclides migration is mineral phase precipitation, known as secondary phases. The formation of one of these phases more stable than UO₂ at repository conditions, could act as a barrier between nuclear waste and groundwater. This modifies the radiation that arrives to the dissolution, blocking dissolution of UO₂ matrix and affecting to radionuclide release. So, is important to know the possible secondary phases to precipitate during SNF (spent nuclear fuel) alteration and its stability at repository conditions. Several experiments of SNF dissolution in groundwater have observed the formation of uranium secondary phases. Nevertheless, these experiments have been developed in specific conditions and they haven't arrived to study the effect of several parameters, such as complexions as phosphate. The rol of phosphate on to dissolution of UO₂ and uranium-phosphate phase formation is necessary to know repository assessment. Uranyl peroxides have been found also in several studies about lixiviation in presence of hydrogen peroxide, which is the expected oxidant for med from the water radiolysis. In this work we performed a study about the stability of these phases. The phases obtained have been characterized with X ray diffraction (XRD). The particle size, shape and chemical composition have been studied by scanning electron microscopy (SEM). A topographic analysis of UO₂ surface in contact with ground water and phosphate media have been performed by means of atomic force microscopy (AFM). The uranium concentration evolution in solution have been followed with ICP-MS. Stability in relation to radiation of uranyl peroxide have been developed with electron transmission microscopy (TEM), and thermal stability with a thermo gravimetry device (TG) and Differential scanning calorimetry (DSC). (Author)

[en] In order to study the dissolution rates for the different radionuclides, the effect of water radiolysis and to elucidate the dissolution mechanisms of the different radionuclides contained in the spent fuel matrix, irradiated spent fuel and UO₂ was used. This study is performed as a part of the collaboration programme ENRESA-CIEMAT-ITU (EC DG/JRC) to provide a source term for use in a performance assessment calculation. For the determination of the dissolution rates a continuous flow through reactor specially designed for hot cell handling was built. This reactor allows the control in situ of different important parameters for leaching experiments such as, redox potential, pH and temperature. These leaching experiments reported the effects of four important parameters (redox potential, pH, carbonate concentration and temperature) on the dissolution kinetics of the spent fuel matrix phase. The kinetic of dissolution of irradiated UO₂ fuel has been studied in synthetic granite groundwater under oxidizing conditions at room temperature. preliminary results indicate that for spent fuel, dissolution rate depends on the burnup, being the dissolution rate calculated for the UO₂ LWR fuel with a burnup of 53 MWd/kg U of 2.66 10''-10 mol m''1 and of 6.77 10''11 mol m''2 s''-1 for the spent fuel of 29 MWd/kg U. (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] 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] 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] 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] One of the mechanisms that may increase the retention of different fission products and actinide elements is their sorption onto uranyl-containing secondary solid phases formed onto the surface of the spent nuclear fuel. Among these phases, studtite (UO₄-4H₂O) has received increasing attention because it was identified on the surface of the spent fuel in leaching experiments in deionized water, due to the radiolytic formation of hydrogen peroxide. In this work, we study the sorption of cesium onto studtite through two series of experiments: (1) kinetics of cesium sorption, and (2) sorption isotherms. The results obtained have shown that cesium sorption is a very fast process, cesium in solution (an almost 40% of a 10^-5 mol.dm^-3 solution) is sorbed in less than one hour at pH around 5. On the other hand, the cesium sorption as a function of initial cesium in solution has been studied between initial cesium concentrations of 7.6.10^-9 mol.dm^-3 and 1.0.10^-3 mol.dm^-3. The data have been modelled considering a Langmuir isotherm and a Freundlich isotherm. The best results have been obtained with a Freundlich isotherm with KF and n values of 10±1, and 1.4±0.1, respectively, with r²=0.998. The results show that the Freundlich isotherm is a good model for the cesium sorption onto studtite (χ²=8.10^-3), especially at cesium concentrations in the range between 7.6.10^-9 mol.dm^-3, and 10^-4 mol.dm^-3. Langmuir isotherm would better fit the experimental data at cesium concentrations at equilibrium higher than 5.10^-4 mol.dm^-3. (authors)

[en] This work has been carried out to analyse different alternatives related to the inner material selection of the Spanish high level waste container for deep geological repository. Its preliminary design considers granitic or clay formations, compacted bentonite sealing, steel corrosion controlled outer wall and glass bed filling of the container. This filling is as relevant as its main role, which is to prevent repository criticality under any foreseen conditions. The present report covers, in first place, the most relevant advances on deep geological storage all around the world, paying special attention on container design solutions. Secondly, having studied carefully the general features of ENRESA preliminary design, the waste forms and all other disposal requirements, a complete and detailed objectives definition is carried out as a selection criterion for candidate materials evaluation and selection. It should be noted that this compilation of demands is significantly deeper and more exhaustive than any other that had been found in literature, including over 20 requirements additionally to another dozen general aspects that could involve improvements in repository performance. Afterwards, eight materials or materials families had been chosen for their potentially interesting properties for geologic disposal. These materials are cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, hematite, phosphates and olivine. Each one of these candidates (under their possible physical forms) had been examined in detail, using available literature and group experience, and evaluated under each of the previously defined objectives. Finally, some relevant conclusions about candidates suitability are extracted from the previous analysis and all the objectives evaluations for each material are summarized in the form of a few matrices to help in decision making. Some other important aspects related to performance improvement, costs, logistics and manufacturing have also been taken into account and summarized in the same manner. As a short conclusion can be said that either cast iron, borosilicate glass, spinel or depleted uranium look quite promising for the mentioned purpose, that further investigation should be needed on hematite and olivine to be able clarify its suitability for this purpose, specially on certain properties related to its behaviour under irradiation, and that neither dehydrated zeolites nor phosphates are expected to fulfil the desired requirements for this application. (Author)

[en] The dependence of the solubility of a microcrystalline uranium dioxide on the chloride concentration has been studied at 25deg C under reducing conditions. The concentration of uranium in solution has been found to be some orders of magnitude lower than in perchlorate media. Possible changes of both the morphology and the composition of the solid phase have been investigated by means of Energy Dispersive X-ray Analysis (EDX) and X-ray Powder Difraction (XPD). The formation of a secondary solid phase as a reason for the decrease of the solubility has been postulated. (orig.)

[en] The dissolution of unirradiated UO₂ (s) has been studied in NaCl and MgCl₂ brines under both reducing and oxidizing conditions. The initial uranium release under reducing conditions has been attributed to the dissolution of an initial oxidized layer. The final uranium concentrations have been modeled by using the PHRQPITZ computer program giving the solubility of the solid phase UO₂ (s). Under oxidizing conditions, the initial release is the sum of the oxidized layer dissolution and the oxidation/dissolution of the UO₂. The release rates calculated are 1.4 x 10^-5 mol/m² in NaCl-brine and 3.6 x 10^-5 mol/m² in MgCl₂-brine. After the initial release, uranium concentration in the NaCl-brine reaches a constant value, which has been attributed to the formation of a secondary solid phase. In MgCl₂-brine, the uranium concentration increases slowly indicating, in this case, no control by secondary phase formation