The anodic dissolution of UO2 has been studied at 60°C and the results compared to previous observations at 22°C. The rate of oxidation / dissolution was determined electrochemically at constant potentials in the range -500 mV to 500 mV (vs. SCE). The composition of the electrochemically oxidized surface was determined by X-Ray Photoelectron Spectroscopy (XPS). The onset of oxidation (UO2 → UO2+x) occurred at approximately the same potential (-400 mV) at both temperatures. However, the conversion of UV to UVI, and hence to soluble UO22+, was accelerated by temperature. This acceleration of dissolution caused the development of acidity at localized sites on the fuel surface at lower (less oxidizing) potentials (≥ 100 mV) at 60°C than at 22°C (≥ 350 mV).

A comprehensive sensitivity analysis method has been developed with the aim of providing quantitative information in an efficient manner. This methodology is composed of the following two components: (1) a statistical method with random sampling of independent parameters, which identifies important parameters and extracts threshold values of parameters and/or combinations yielding a 'successful condition' where maximum dose does not exceed a target value, (2) A nuclide migration model that as far as possible incorporates a comprehensive set of phenomena occurring within the repository. This approach was applied as part of a safety assessment of the geological disposal of TRU waste in Japan (TRU-2). It was shown that the concept of TRU waste disposal is robust from the point of view of safety.

REE concentrations have been determined in groundwaters, granite and fracture fillings in a restored uranium mine.

The granitoids normalized REE patterns of groundwaters show HREE-enrichment and positive Eu anomalies. This suggests that the REE are fractionated during leaching from the source rocks by groundwaters. Preferential leaching of HREE would be consistent with the greater stability of their aqueous complexes compared to those of the LREE, together with the dissolution of certain fracture filling minerals, dissolution/alteration of phyllosilicates and colloidal transport.

Recent studies have shown that in the presence of H2 overpressure, which forms due to the corrosion of the Fe based container, the dissolution rate of the spent fuel matrix is slowed down by a factor of about 10, associated with a distinct decrease of concentrations of important ra-dionuclides. However, in a natural salt environment as well as in geological formations with chloride rich groundwater the presence of radiation chemically active impurities such as bro-mide must be taken in consideration. Bromide is known to react with β/γ radiolysis products, thus counteracting the protective H2 effect. In the present experiments using high burnup spent fuel it is observed that during 212 days the matrix dissolution rate was enhanced by a factor of about10 in the presence of up to 10-3 M bromide and 3.2 bar H2 overpressure. However, concen-trations of matrix bound actinides were found at the same level or below as found under identical conditions, but in the absence of bromide. In the long-term it is expected that the effect of bro-mide becomes less important, because the decrease of β/γ-activity results in a decrease of oxida-tive radicals, which react with bromide, while α activity will dominate the radiation field.

There are two compelling reasons for understanding source term and near-field processes in a radioactive waste geologic repository. First, almost all of the radioactivity is initially in the waste form, mainly in the spent nuclear fuel (SNF) or nuclear waste glass. Second, over long periods, after the engineered barriers are degraded, the waste form is a primary control on the release of radioactivity. Thus, it is essential to know the physical and chemical state of the waste form after hundreds of thousands of years. The United States Department of Energy's Yucca Mountain Repository Program has initiated a long-term program to develop a basic understanding of the fundamental mechanisms of radionuclide release and a quantification of the release as repository conditions evolve over time. Specifically, the research program addresses four critical areas: a) SNF dissolution mechanisms and rates; b) formation and properties of U6+-secondary phases; c) waste form–waste package interactions in the near-field; and d) integration of in-package chemical and physical processes. The ultimate goal is to integrate the scientific results into a larger scale model of source term and near-field processes. This integrated model will be used to provide a basis for understanding the behavior of the source term over long time periods (greater than 105 years). Such a fundamental and integrated experimental and modeling approach to source term processes can also be readily applied to development of advanced waste forms as part of a closed nuclear fuel cycle. Specifically, a fundamental understanding of candidate waste form materials stability in high temperature/high radiation environments and near-field geochemical/hydrologic processes could enable development of advanced waste forms “tailored” to specific geologic settings.

The sorption and diffusion of Cs and HTO in compacted bentonite was investigated at temperatures from 30 to 60°C. The apparent (Da) and effective (De) diffusion coefficients were determined by in-diffusion and through-diffusion experiments with a constant boundary concentration maintained. The temperature dependence of De and Da obeyed an Arrhenius-type equation, allowing determination of the activation energy for diffusion of Cs and HTO. The De value of Cs was three times the De of HTO, which is considered to be a result of surface-excess diffusion. Cs may be concentrated near the surface of the negatively charged clay, thus giving a large diffusive flux. The activation energies for Cs diffusion were 21.4×2.8 kJ/mol and 37.3×1.5 kJ/mol as determined based on De and Da, respectively. This difference was due to the temperature dependence of the distribution coefficient Kd of Cs.

At the end of fifteen years of researchs defined by the French act of December 30, 1991 on radwaste management, Andra gave a report, “Dossier Argile 2005”, which concluded with the feasibility of a reversible disposal in the argillaceous Callovo-Oxfordien formation studied by means of an underground research laboratory at Meuse/Haute-Marne site. Starting from source data like the characteristics of the geological medium and the waste inventory, the process followed by Andra to achieve at this conclusion is of type sequential and iterative between concept design, scientific knowledge, in particular that of the phenomenological evolution of the reposiroty and its geological environment from operating period to long term, and Safety assessment. The “Dossier Argile 2005” covers a broad radwaste inventory, ILLW, HLW and Spent Fuel, so that it makes it possible to cover whole of the technological, scientific and safety topics. This article will give an overview of the geological disposal studies in France and draw the main conclusion of the Dossier 2005 Argile. It will be focused on the near field (Engineering components and near field host rock), while considering if necessary its integration within the whole system. After a short description of the concepts (incl. waste inventory and the characteristics of the Meuse/Haute the Marne site) and the functions of the components of repository and geological medium, one will describe successively the broad outline of the phenomenological evolution of repository and the geological medium in near field, by in particular releasing the time scales of processes and uncertainties of knowledge. On this basis, one will indicate the safety scenarios which were considered and the broad outline of performance and dose calculations. Lessons learn from the Dossier 2005 Argile will be discussed and perspective and priority for future will be indicated.

First results of combining X-ray microcomputer tomography (µCT), confocal laser-scanning microscopy (CLSM) and 14C-polymethylmethacrylate (14C-PMMA) impregnation techniques in the study of granitic rock samples are reported. Combining results of µCT and CLSM with those of the 14C-PMMA technique, the mineral-specific porosity and morphology of the open pore space, as well as its connectivity, could be analyzed from a micrometer up to a decimeter scale.

Three different types of granite were studied. In two cases part of the micro-fissure and pore apertures were found to be in a micrometer scale, but in one case all grain-boundary openings were below the detection limit. Micrometer-scale apertures could be analyzed by CLSM and µCT. The benefit of µCT is that it can also provide the heterogeneous distribution of minerals in 3D. The 2D porosity distributions in the mineral phases, consisting of nanometer-scale pores, could be measured by the 14C-PMMA method together with the micro-fissures. This method does not, however, give the exact pore apertures. The limitations and applicability of the methods are discussed.

A geological repository for high-level radioactive waste (HLWR) consists on a multi-barrier system, emplaced hundred meters deep in a geological medium. In most of the repository concepts, the waste would be located in metal canisters surrounded by a layer of compacted clay, i.e. bentonite. To guarantee the long-term safety of a repository, all mechanisms that could affect the radionuclide (RN) migration rate must be well defined and quantified. The particular interest of this work lies on the possible contribution of bentonite colloids to RN transport. The first parameter necessary to assess the colloid-mediated transport is the quantification of the bentonite colloid source term. Secondly, it is necessary to define if colloids remain stable in the geochemical conditions of the medium.

Several mechanisms that are basically related to the hydration of the clay can lead to bentonite colloid generation. In the present work the colloid generation is evaluated at laboratory scale under “realistic” conditions, considering static hydration (no flow). To do so, two experimental set-ups were designed with the aim of quantifying the bentonite colloid generation rates. The experimental cells were designed to study the colloid formation in a confined system by introducing compacted bentonite, at different compactation densities, in stainless steel porous filters. The bentonite hydration is facilitated by immersing the confined cells in different electrolytes, from the most favorable conditions (lowest ionic strength) to different groundwaters of interest as aqueous phase. The concentration of bentonite colloids and the average particle size are evaluated as function of time by Photon Correlation Spectroscopy measurements in the aqueous phase.

Preliminary results showed that all the bentonite particles generated have average size in the colloid range, equivalent to that of bentonite colloids prepared in the laboratory, despite the filter porous sizes were hundred times higher. The experimental set up allows performing stability evaluation at the same time and that after months the colloids generated in the lower strength electrolytes remain stable. The configuration allows quantification of the colloid generation rates. The mechanisms responsible of colloid generation are discussed according to the obtained results in different experimental conditions.

Geopolymers should be serious waste form candidates for intermediate level waste (ILW), insofar as they are more durable than Portland cement and can pass the PCT-B test for high-level waste. Thus an alkaline ILW could be considered to be satisfactorily immobilised in a geopolymer formulation. However a simulated Hanford tank waste was found to fail the PCT-B criterion even for a waste loading as low as 5 wt%, very probably due to the formation of a soluble sodium phosphate compound(s). This suggests that it could be worth developing a “mixed” GP waste form in which the amorphous material can immobilise cations and a zeolitic component to immobilise anions. The PCT -B test is demonstrably subject to significant saturation effects, especially for relatively soluble waste forms.

Once-through fuel cycles directly dispose of one primary waste form: spent nuclear fuel (SNF). In contrast, the reprocessed SNF yields multiple waste streams with different chemical, physical, and radionuclide characteristics. These different characteristics of each waste stream imply that there are potential cost and performance benefits to developing different disposal sites that best match the disposal requirements of each waste stream. Disposal sites as defined herein may be located in different geologies or in a single repository containing multiple sections, each with different design characteristics. The paper describes the results of a series of studies on disposal options for specific wastes and the potential for a waste management system that better couples various reprocessing plant wastes with disposal facilities.

Calcined high-level radioactive waste (HLW) stored at the Idaho National Laboratory (INL) will eventually be immobilised in a suitable wasteform before disposal. A tailored glass-ceramic wasteform, produced by hot isostatic pressing in stainless steel (SS) cans, has been developed at ANSTO as a cost-saving alternative to glass which would improve waste loading and density, and reduce waste volume. We have studied the SS/wasteform interactions under HIPing conditions to understand whether such interactions would have any detrimental effect on long-term wasteform stability. This has been demonstrated by carrying out aqueous durability tests, under near-neutral and alkaline conditions, on the wasteform at the interaction layer, and on the wasteform distal to this reaction edge. Reaction during HIPing resulted in Cr diffusion from the can wall into the wasteform, however without any detectable detrimental impact on the HIP can or the aqueous durability of the wasteform.

The GM2003 model extends the r(t) glass dissolution model with water diffusion through the diffusion layer. Boron and alkali diffusion through the diffusion layer is described by introducing a retention factor Kd,i between boron/alkali and water in the diffusion layer. Introducing a boron/alkali diffusion coefficient, the BRAG model describes boron/alkali diffusion in the diffusion layer as well. It is shown that both models are consistent with each other and an expression is derived for the boron/alkali diffusion coefficient (BRAG) as a function of both parameters of GM2003: the retention factor Kd,i and the water diffusion coefficient DH2O in the diffusion layer. From dissolution data only, it is possible to fit the value for the boron/alkali diffusion coefficient in the diffusion layer but due to correlations the individual values of both parameters Kd,i and DH2O of GM2003 cannot be determined. From theoretical considerations follows that the Kd,i value for boron/alkali should be slightly larger than 0.1 kg/liter. A user friendly code for the BRAG model allows automatic fits of glass dissolution data in water.

Phosphate matrices are considered as potential candidates for the specific immobilization of tri- and tetravalent actinides. One of the main properties of interest of these matrices concerns their chemical durability. It was studied through their behavior regarding to the retention of actinides and developped using under- and over-saturation experiments. In this field, neoformed phases were precipitated. Lanthanides were used as surrogates for trivalent actinides while uranium and thorium were used for tetravalent actinides. These phases were extensively characterized through several analytical and spectroscopic techniques (SEM, EPMA, XRD, µ - Raman, TRLFS…).

New phosphate form such as Nd1-2xCaxThx-yUy(PO4,F). ½ H2O rhabdophane were identified during the dissolution of britholites. The precipitation of Nd1-2xCaxThxPO4. ½ H2O rhabdophane was obtained and it appeared that there was a segregation between neodymium, calcium and thorium after a few days to several months depending on the thorium weight loadings and the precipitation time, thorium precipitating as TPHPH and neodymium as NdPO4  ½ H2O. The determination of the solubility constants showed very low values which means that these phases exhibit some interesting properties for the retention of radionuclides.

Technetium-99 is a long-lived product of nuclear fission. In the ground-waters of European disposal sites Tc is expected to be in the tetravalent state since the geochemical conditions are reducing due to the presence of redox couples such as Fe(II)/Fe(III), and the presence of reducing minerals such as pyrite and siderite (FeCO3). Siderite is expected to be formed as well as container corrosion product. Experiments of coprecipitation of Fe2+ and Tc(IV) were conducted in the laboratory under an inert atmosphere at room temperature. High resolution transmission electron microscopy shows that the coprecipitate is around 3 nm sized particles. Analytical electron microscopy indicates that the precipitate contains 30 atom % of Tc and 70 atom % Fe. The Tc was homogeneously distributed. Selected Area Electron Diffraction doesn't reveal the formation of siderite, which suggests that the presence of technetium in high concentration inhibit the formation of this phase.

Spent Nuclear Fuel (SNF) source terms are used to define the release rate of radionuclides (RN) in a direct disposal and to assess the performance of this waste form. They classically distinguish between two contributions: (i) the Instant Release Fraction (IRF) of RN which are directly leached when water contacts the fuel, (ii) the slow and long term release of RN which are embedded within the fuel matrix. Recent experimental results bring significant input in our understanding and assessment of both contributions. However, they have not yet been integrated in the definition of the SNF source term. This paper will present the impact on the RN source term of the latest results on the SNF long term evolution and the key remaining scientific issues.

During aqueous alteration of glass an alteration layer appears on the glass surface. The properties of this alteration layer are of great importance for understanding and predicting the long-term behavior of high-level radioactive waste glasses. Numerical modeling can be very useful for understanding the impact of the glass composition on its aqueous reactivity and long-term properties but it is quite difficult to model these complex glasses. In order to identify the effect of the calcium content on glass alteration, seven oxide glass compositions (57SiO2 17B2O3 (22−x)Na2O xCaO 4ZrO2; 0 < x < 11) were investigated and a Monte Carlo model was developed to describe their leaching behavior. The specimens were altered at constant temperature (T = 90°C) at a glass-surface-area-to-solution-volume (SA/V) ratio of 15 cm 1 in a buffered solution (pH 9.2). Under these conditions all the variations observed in the leaching behavior are attributable to composition effects. Increasing the calcium content in the glass appears to be responsible for a sharp drop in the final leached boron fraction. In parallel with this experimental work, a Monte Carlo model was used to investigate the effect of calcium content on the leaching behavior. The glass structure was built with a random distribution of silicon and boron on the lattice points of a diamond lattice with sodium atoms located in interstitial position. The model was upgraded to take into account the effect of zirconium and calcium. Zirconium atoms are placed in interstitial position (with cations acting as charge compensators) and interact with six lattice atoms (Si or B) by decreasing their dissolution probability. Acting like sodium but with less effect, calcium atoms increase the dissolution probability of atoms placed in their neighborhood. Monte Carlo simulations performed with this model are in good agreement with the experimental results. The dependence of the dissolution rate on the calcium content can be described by a quadratic function: fitting the simulated points gives a minimum alteration rate at about 7.4 mol% calcium. This value is consistent with the figure of 8.0 mol% obtained from the experimental work. The model was also used to investigate the role of calcium in the glass structure and it pointed out that calcium act preferentially as a network modifier rather than a charge compensator in this kind of glasses.

The global energy context pleads in favour of a sustainable development of nuclear energy since the demand for energy will likely increase, whereas resources will tend to get scarcer and the prospect of global warming will drive down the consumption of fossil fuel sources.

How we deal with radioactive waste is crucial in this context. The production of nuclear energy in France has been associated, since its inception, with the optimization of radioactive waste management, including the partitioning and the recycling of recoverable energetic materials. The public's concern regarding the long-term waste management made the French Government to prepare and pass the December 1991 Law, requesting in particular the study for fifteen years of solutions for still minimizing the quantity and the hazardousness of final waste, via partitioning and transmutation.

At the end of these fifteen years of research, it is considered that partitioning techniques which have been validated on real solutions are at disposal. Indeed, aqueous process for separation of minor actinides from the PUREX raffinate has been brought to a point where there is reasonable assurance that industrial deployment can be successful. A key experiment has been the kilogram scale successful trials in the CEA-Marcoule Atalante facility in 2005 and this result, together with the results obtained in the frame of the successive European projects, constitutes a considerable step forward. For transmutation, CEA has conducted programmes proving the feasibility of the elimination of minor actinides and fission products: fabrication of specific targets and fuels for transmutation tests in the HFR and Phénix reactors, neutronics and technology studies for critical reactors and ADS developments. Scenario studies have also allowed assessing the feasibility, at the level of cycle and fuel facilities, and the efficiency of transmutation in terms of the quantitative reduction of the final waste inventory depending of the reactor fleet (PWR-FR-ADS).

Important results are now available concerning the possibility of significantly reducing the quantity and the radiotoxicity of long-lived waste in association with a sustainable development of nuclear energy. As France has confirmed its long-term approach to nuclear energy, the most effective implementation of P and T of minor actinides relies on the fast neutron GEN IV systems which are designed to recycle and manage their own actinides. The perspective to deploy a first series of such systems around 2040 supports the idea that progress is being made: the long-term waste would only be made up of fission products, with very low amounts of minor actinides.

In this sense, the new waste management Law passed by the French Parliament on June 28, 2006, demands that P and T research continues in strong connection to GEN IV systems and ADS development and allowing to assess the industrial perspectives of such systems in 2012 and to put into operation a transmutation demo facility in 2020.

Copper of the quality oxygen free and high conductivity, doped with phosphorus (Cu OFP) has been chosen as the material for the fabrication of high level nuclear waste containers in Sweden. This material will be the corrosion barrier for spent fuel in the environment of a deep geological repository. It is planned that the service life of this container will be 100,000 years. During this time water with high concentration of chloride ions might come in contact with the copper surface. If pH conditions are appropriate, this might cause pitting corrosion. This work reports a study of the susceptibility of Cu OFP to corrosion when chloride ions are present, both deionized water (DW) and in standard synthetic underground water (SUW). The techniques used were electrochemical methods such as corrosion potential evolution and Tafel curves. In addition, this system was studied with Electrochemical Impedance Spectroscopy (EIS). We also used characterization techniques such as Scanning Electronic Microscopy (SEM), Energy Disperse Spectroscopy (EDS) . The main conclusions are that copper is more susceptible to corrosion at high chloride ion concentration. Additionally, when the chloride concentration is low, it is possible to form copper chloride crystals, but at the highest concentration, copper chloride complexes are formed, leaving the copper surface without deposits. When the chloride concentration is low (<0.1 M) the corrosion process is mainly controlled by diffusion, while at higher concentrations (0.1M to 1M) corrosion is controlled by transport processes. At low concentration of sulfide ( <3*10-5M), copper corrosion in the presence of chloride is controlled by diffusional processes.

For improving the accuracy of the performance assessment studies related to the spent fuel safety under storage conditions it is necessary to develop a new matrix alteration model. These models must be based on laboratory experiences and they should be capable to extrapolate to storing environmental conditions. Most of recently models developed included the oxidation and dissolution process of the spent fuel matrix, but the influence of a possible process of secondary phase formation over the spent fuel surface is not taken into account yet. This is a key process that could produce a reduction of the matrix dissolution rate, radiation shielding behaviour; however, the surface precipitation of the secondary phase could induce a localized corrosion process, which in this case dissolution rate of the spent fuel would be increased. This paper is focussed on microstructural characterization of secondary phases formed in coprecipitation experiments performed under anoxic conditions in granitic-bentonitic simulated groundwater. In order to simulate the influence of the container material, the coprecipitation experiments were performed in absence and presence of iron powder. The solid phases formed were characterized using the following techniques: XRD; SEM-EDX and TEM-EDX. The XRD diffraction pattern showed that under anoxic conditions a mixture of phases were obtained (sodium and potassium uranate and schoepite), whereas uranate phases were detected when only iron was present. The characterization study indicates that the U secondary phase formed (under reducing conditions and presence of iron powder) growth from iron surface. The crystal size of the secondary phase is independent of the presence of iron powder (and it is always less than 3 μm). Furthermore, the microstructural study showed the growing of U phases over iron powder.