The groundwater corrosion of three archaeological model glass compositions, a replica Roman glass (B1), a medieval glass (B2) and a 17th century glass (B3), recovered from the Ballidon burial site, were investigated by field emission scanning electron microscopy. Given the equivalent burial conditions for all three glasses, the different corrosion mechanisms of these glasses are dependent principally upon composition, and in particular the dominant alkali and network former content. Composition B1 was found to the most durable glass and showed progressive Na leaching coupled with dissolution of the silicate matrix in the outermost 5μm after 32 years of burial. The B2 and B3 compositions were found to be less durable and corroded via congruent dissolution of the silicate matrix and in situ condensation, resulting in the formation of a discrete 5-15μm thick layer after only 1 year of burial. The formation of a protective Ca-rich surface layer was observed for B2, hindering successive inward growth of the dissolution reaction and favouring localised pitting of the extant layer. The protective surface phenomenon was not observed for B3, and a large increase in layer thickness was observed after nine years of burial.

A number of possible options have been proposed for the encapsulation and immobilisation of long lived actinide (Act) fractions in nuclear waste. Ceramics offer superior durability against chemical migration and the ability to be tailored to accommodate a variety ofdifferent waste streams. Research on the fabrication of dense, durable crystalline matrices for the safe disposal of fissile plutonium is ongoing and this study reports quantitative chemical, structural and spectroscopic analysis on fluorite based host phases.

Ceramics based on the fluorite structure are known to be able to incorporate a variety of actinides and in this work two candidate ceramic matrices were investigated: a pyrochlore, Gd2Zr1.60Ce0.20Hf0.20O7; and a zirconolite, (Ca0.90Gd0.10)(Zr0.50Ce0.20Hf0.20Gd0.10)Ti2O7. The chemical compositions of the two major phases observed in the ‘zirconolite’ sample were consistent with the 2M and 4M zirconolite polytypes and the presence of the 4M structure was confirmed by Electron Diffraction (ED). The major phase in the ‘pyrochlore’ ceramic was confirmed by ED to have the pyrochlore structure. Electron Energy Loss Spectroscopy (EELS) data indicated the presence of both Ce3+ and Ce4+ in all the samples.

In a dry air storage facility, the oxidation of irradiated nuclear fuel is considered as the main risk induced by a leakage that would put the nuclear ceramics in contact with air. The oxidation of UO2 in U3O8 leads indeed to a relative volume swelling of more than 30% which could cause the ruin of the fuel rod. In order to fill the lack of data on high burn-up fuel, some oxidation experiments are currently conducted in CEA with UO2 and MOX fuel. Fragments as well as slices of fuel rod, with a high burn-up, have been oxidized in air at 200°C for more than 4000 hours. The associated weight gain curves are presented together with the characterization of some samples picked up at different oxidation times. It is shown that the weight gain curve of high burn-up UOX is not parabolic-shaped, as predicted in literature, but sigmoid-shaped. Moreover it is also shown that the grain boundary oxidation is not uniform inside a fragment. This leads to a new description of the oxidation mechanism of irradiated fuel.

In this paper, an evaluation of the alpha dose rate profile at the UO2/water interface is presented. The calculation is performed in spherical geometry by considering the position of alpha emitters in the material. Moreover it proceeds to an iterative way. The fuel matrix and water is divided in spherical layers of 0.1 μm of thickness. The alpha emitters have an average energy of 5.3 MeV corresponding to a maximum distance of 13.2 μm in UO2 matrix and 42 μm in water. The direction of emission and the energy deposition, described by the Bragg curve, in the nuclear fuel are also taken into account. It appears that the alpha dose rate strongly decreases towards the calculated maximum penetration depth in water namely 42 μm. The concentrations of oxidising and reducing species are largely higher at the UO2/water interface than in solution. This alpha dose profile is used as input file in a model coupling a radiolysis kinetic model and a diffusion model. This kind of calculation is particular suitable for HTR kernels which have a spherical form of few hundreds microns of diameters (200-500 μm).

Immobilization of adsorbed Cs+ and Sr2+ on a molybdenum-doped, hexagonal tungsten bronze (HTB)-polyacrylonitrile (PAN) composite adsorbent can be achieved by heating in air at temperatures in the range 600 - 1200 °C. Thermal treatment of the parent composite material at 800 – 1000 °C undergoes a ca. 60% reduction in volume and retains its spherical morphology. For materials prepared at 800 – 1200 °C the full complement of Cs+ and the majority of Sr2+ partition into HTB phases (A∼0.16-0.3MO3; A = Cs, Sr, Na; M = Mo,W), along with sodium cations. The presence of high concentrations of Na+ relative to either Cs+ or Sr2+ does not appear to interfere with the formation of the HTB phase. The fraction (f) of Cs+ and Sr2+ leached from the tungstate phase assemblages is better or comparable with cesium hollandite (Cs0.8Ba0.4Ti8O18; f = ca. 8 × 10−5; rate = <1.2 × 10−4 g m−2d−1) and strontium titanate (SrTiO3; f = 3.1 × 10−3; rate = 2.63 × 10−4 g m−2day−1), respectively, using a modified PCT test. Furthermore, where aggressive leaching conditions are employed (0.1M HNO3; 150 °C; 4 days), the tungstate phase assemblages display leach resistance orders of magnitude better than the reference phases (Cs+ - f = ca. 5 × 10−3; rate = ca. 1.4 × 10−3 g m−2day−1; Sr2+ - f = ca. 8 × 10−2; rate = ca. 2.5 × 10−2 g m−2day−1).

In this study we experimentally investigated the interaction of Mo(VI) with magnetite, which is a corrosion product of steel. The work was conducted with commercial magnetite, and we studied the influence of pH, pe and solid/liquid ratio on Mo sorption. A Surface Complexation Model (SCM) has been applied tothe experimental data, allowing to explain the results using the Diffuse Layer Model (DLM) and by considering the formation of the monodentate complex: >FeOMo(OH)5. At pH 2, experimental data were satisfactorily fitted to a a Langmuir isotherm.

Using a variety of electrochemical and surface analytical techniques, the mechanism and kinetics of Cu corrosion in anoxic, aqueous, sulphide-containing environments are being investigated. Under these conditions ([S]total = 10−4 to 3 ×10−3 mol/L), the anodicgrowth ofa film (XRD identifies Cu2S/Cu1.8S as major/minor phases, respectively) is supported by the cathodic reduction of water thereby destabilizing the copper surface. For more oxidizing conditions, the subsequent growth of a partially passivating film is observed. Electrochemical impedance spectroscopy (EIS) and cyclic voltammetry at rotating disc electrodes show film growth occurringwith negligible dissolution and under partial SH- transport control. Current-potentialrelationships as a function of [S]total give Tafel slopes of ∼ (40 mV)−1 suggesting reaction occurs via a 2step process: Aninitial rapid adsorption of SH- leading to an equilibrium surface concentration, followed by a rate determining electron transfer to form a sulphide film. It is proposed thatfilm growth propagates via transport of CuI through the film to the solution interface. The primary goal of this research is the development of a mathematical model which can be used to assess the performance of copper nuclear waste canisters in granitic repositories.

Structural studies of single phase or nearly single phase zirconolite ceramic samples have been conducted using electron microscopy and microanalysis, X-ray diffraction, neutron diffraction, and spectroscopic methods. We show that it is possible to produce a complete series of zirconolite 2M samples with substitution of 2Ti by Nb+Fe in the HTB layer. The samples are single phase up to about 80% Nb +Fe substitution, with the appearance of a minor perovskite phase at higher Nb+Fe levels. Electron probe microanalysis reveals that the samples are homogeneous and close to their nominal compositions, except for those containing perovskite, which have a slight excess of Zr and a deficiency in the Fe content. The lattice parameters and the positions of certain Raman bands are non-linear as a function of composition, suggesting the possibility of cation ordering over the three available Ti sites within the HTB layer. Rietveld refinement of Synchrotron X-ray powder data for the Nb+Fe end-member have been conducted for the disordered case and for six trial models each with a different ordering scheme. Results of this exercise indicate that Fe preferentially occupies the Ti2 (split) site with partial ordering of Nb and the remaining Fe over the Ti1 and Ti2 octahedra. The preference of Fe for the five coordinated Ti2 site has been confirmed by 57Fe Mossbauer spectroscopy.

The potential repository at Yucca Mountain in Nevada is in an arid area, in unsaturated volcanic tuff with oxidising conditions, and may be kept open and ventilated for hundreds of years. Existing US work on archaeological material and structural analogues in such conditions has been extended to seek further well-preserved objects in such conditions. The objectives of the study are partly scientific and partly aimed at the public. A wide range of exceptionally wellpreserved glass, metal and organic materials is found in the archaeological record from sites that have been characterised by arid conditions for many thousands of years. Underground preservation environments include those either continuously open (caves) or sealed but not ‘backfilled’ (tombs). Preservation of materials in such openings provides a useful analogue reference point for considering the operation and evolution of the Yucca Mountain repository.

A biogeochemical model was developed with an aim to illustrate the transport and reaction processes that may be involved in copper canister corrosion by sulphide in a deep repository for nuclear waste. Sulphide concentrations in the groundwaters in Sweden are relatively low and are generally considered to be of little importance for corrosion of the copper canisters. Sulphate, however, is present in relatively large amounts (50-700 mg/L) and may be reduced to sulphide by reaction with organic matter. Mediated by microbes, the reduction rate may be increased considerably compared to abiotic reduction. Microbially mediated sulphate reduction occurs in many natural environments, provided that reactive organic matter is available. Groundwater analyses indicate relatively high concentrations of dissolved methane (up to 16 mg/L) which thermodynamically is a suitable reducing agent. This processes could occur in fractures inthe bedrock provided that substrates are continuously supplied and that there is a sink for the reaction products. In the repository, the copper canisters may provide a sink for sulphide that yields a favourable environment for the microbe population. The model domain includes a canister with surrounding backfill material intersected by a hydraulically conductive fracture. Transport of substrates and reaction products in the fracture and backfillas well as microbially mediated reaction at the fracture opening is included in the model. Results indicate that microbially produced sulphide could potentially contribute to canister corrosion.

The results of 3H, 36Cl and 137Cs diffusion experiments through compacted bentonite using a new design of diffusion cell and a new methodology of diffusion coefficients evaluation are presented. The diffusion cell was made from the stainless steel and enables to connect it directly to the input and/or output reservoirs without any tubing. The evaluation of diffusion coefficients utilizes a compartmental model developed in the environment of the GoldSim transport code. It enables to determine diffusion coefficients for various types of boundary conditions, including also input and output filters. The influence of the diffusion through filters on the determined values of both effective (De) and apparent (Da) diffusion coefficients was numerically demonstrated for the through diffusion method. This effect is most important for Da, the value of which would be underestimated using standard ways of evaluation for neutral and positively charged species, mainly in the case of high effective porosity.

The comparison of standard and the newly developed method of evaluation of diffusion coefficients showed a significant influence of diffusion in filters for HTO. Contrary to the standard method of evaluation, the evaluation taking into account filters showed here no difference between total and effective porosity. The effect of filter resistance was negligible for Cl-, especially at high dry density of compacted bentonite, due to the anion exclusion effect. The numerical model developed enabled to determine Da values of Cs+ from the concentration change in the inlet reservoir.

Eu migration in a Callovo-Oxfordian argilite sample was studied using the micro LIBS technique. Quantitative elemental mapping were made by micro LIBS that showed the actual distribution of the various micro areas observed on the Callovo-Oxfordian sample's surface. Calcite, dolomite, alumino-silicates, quartz, pyrite and iron oxides were identified and their statistical distribution was determined. Experimental Eu profiles observed are consistent with diffusion process accompanied by heterogeneous sorption on alumino-silicate surface.

Britholite, Ca-REE silicate with apatite structure, is an actinide host phase occurred in vitreous borosilicate waste forms. Such glass-ceramics are considered as potential host phases for immobilization of actinide-containing high-level waste. Crystalline phases have to be radiation resistant for this application. Radiation stability of the britholites was mainly studied by either heavy ions irradiation or incorporation of Cm-244 or Pu-238 and 240. A wide range of critical doses (0.15 - 0.6 dpa at 25°C) and temperatures have been obtained depending on the compositions of the samples. Natural analogue study of the waste forms allows to predict the behavior of actinide host phases for long periods after disposal. Britholites with age from 320 to 2600 millions years, ThO2+UO2 content from 1.0 to 12 wt.%, and cumulative doses from 0.6×1019 to 7.7×1019 α-decays/g have been studied. The britholite becomes amorphous at a dose of 1 dpa (0.9×1019 α-decays/g) and higher. Critical doses for natural minerals are higher than those for synthetic samples, most likely due to re-crystallization during annealing.

The proposed waste package design for the disposal of high-level radioactive waste at the potential repository at Yucca Mountain, Nevada, consists of an outer container made of Alloy 22, a corrosion resistant Ni-Cr-Mo-W alloy, surrounding an inner container made of Type 316 nuclear grade stainless steel. Models to assess the influence of waste package degradation modes on the overall system performanceconsider uniform and localized corrosion processes. Based on measurements of passive anodic current density, the uniform aqueous corrosion rate of Alloy 22 is estimated to be slow, and long waste package lifetimes are projected in the absence of conditions that promote accelerated degradation processes. The initiation of localized corrosion is possible in chloride-containing waters with low concentrations of inhibiting oxyanions such as nitrate. Although propagation rates for localized corrosion are typically orders of magnitude greater than the passive uniform corrosion rates, the maximum penetration depth of localized attack may be limited to depths significantly less than the container thickness as a result of stifling andarrest of localized corrosion.

The applicability of a process, originally developed to immobilize a chloride-rich intermediate level waste (ILW) arising from the pyrochemical reprocessing of plutonium, to alternative ILW waste-streams containing a large variety of cations, present as oxides together with KCl and CaF2, has been investigated. Non-radioactive studies, using simulated wastes containing hafnium and samarium as surrogates for the actinides, have been performed to establish the optimum processing parameters for the manufacture of wasteforms suitable for long-term storage/disposal. It has been demonstrated that the wastes can be immobilized by calcining with either Ca3(PO4)2 or CaHPO4 as the host. In excess of 97.5 % of the chloride present in calcined wastes, containing up to 3.2 wt% halides as a mixture of F- and Cl-, is immobilized when using this process. Whereas in the original calcined waste, the Cl- (about 12 wt%) was immobilized as chlorapatite, Ca5(PO4)3Cl, and spodiosite, Ca2(PO4)Cl, in these current wastes fluorapatite is the predominant halide containing phase. Conversion of the immobilized waste powder into a monolithic wasteform was achieved by sintering the powder with the addition of a sodium aluminium phosphate based glass. Both the calcined powder and the sintered monolithic wasteforms showed excellent durability in aqueous solution.

In the framework of the contract “Advanced Melter Technology Application to the Defense Waste Processing Facility (DWPF) –Cold Crucible Induction Heated Melter (CCIM)”, vitrification tests with Savannah River Site defense waste surrogate were performed at the SIA Radon facility. Cold crucible melters with inner diameter of 216 mm and 418 mm were used in the testing. Commercially available (USA) frits 200 and 320 were used as glass-forming additives. In three different test campaigns, waste additive mixtures were fed as slurries with ∼60 wt.%, ∼30 wt.%, and 45 wt.% water content. Maximum slurry capacity and glass productivity under steady-state conditions were 35.4 kg/h and 16.2 kg/h, respectively. Specific glass productivity reached up to ∼3000 kg/(m2×day). The average melt process temperature was 1250- 1350 °C. Waste loadings in glass were 45 wt.% in tests 1 and 2 and 50 wt.% in test 3. The glasses produced were found to be homogeneous but contained a magnetite-type phase with the spinel structure due to high iron and manganese content in waste. Spinel was observed in the glassy matrix as individual regular crystals and their aggregates. All the waste uranium entered the vitreous phase. Infra-red spectra consist of strong absorption bands due to bridging Si-O-Si and non-bridging Si-O- bonds, some weak bands due to B-O bonds, and a number of narrow bands due to occurrence of the crystalline phase. The glassy products demonstrate high leach resistance. Normalized release of major glass elements (Na, Li, B, Si) is by 10 to 50 times lower than the values required for repository disposition by EPA.

Existing knowledge on glass dissolution and silica diffusion in Boom Clay is validated by experiments where both phenomena could be studied simultaneously. SON68 glass coupons, doped with radioactive 32Si, were sandwiched between two cores of fresh humid Boom Clay and heated to 30°C. At the end of the experiment, the system was dismantled, the mass loss of the glass coupon was measured, and the clay core was sliced to determine the diffusion profile of the 32Si dissolved from the glass. These data were completed with analyses of the clay water and surface analyses for analogous tests with undoped glass. The results are interpreted by assuming congruent glass dissolution at a constant rate, with a glass silica saturation concentration between 14 and 20 mg/l, a forward glass dissolution rate (at zero silica concentration) of 0.028 g.m−2day−1, an apparent silica diffusion coefficient in the clay of 1.4 10−12 m2sec−1, and a distribution coefficient for silica on Boom Clay between 0.010 and 0.075 m3kg−1. These parameter values are close to the range found in literature. It was not necessary to consider diffusion through the gel, precipitation or detailed geochemical reactions. The modeling exercise shows that the existing knowledge about the subsystems glass and clay can succesfully be integrated to describe the coupled processes in the whole system.

Stress corrosion cracking is a potential failure mechanism of the copper canister proposed to be the main corrosion barrier in the Scandinavian design of the final nuclear waste disposal vault. The main target in SCC-oriented research is to determine the limiting values for the critical concentration of harmful species, the potential range in which these species cause SCC and the threshold stress level required for SCC. The present paper attempts to redefine the criteria for SCC susceptibility of copper in acetate solutions through analysis of potentiodynamic curves using both rotating disc and thin plate electrodes insulated on one side. A potential range in which the current densities at a slow sweep rate are lower than those at a fast sweep rate exists for Cu in 0.001 M - 0.1 M acetate. However, the values of the current densities within this range are significantly smaller than those postulated for active anodic dissolution, implying dissolution mediated by a surface film. New criteria for the SCC susceptibility of pure copper in acetate solutions have been formulated through Stern-Geary analysis of the polarisation curves. Several electrochemical characteristics, such as the current density at the positive crossover potential, the ratio between the corrosion current densities determined from the slow and fast sweeps, as well as the anodic current density in a fast sweep show a clear jump for concentrations between 0.01 M and 0.05 M. This, however, cannot yet be considered as an indication of SCC susceptibility at higher acetate concentrations, as evidenced by complementary SSRT tests.

The dissolution of irradiated HTR UO2-ThO2 fuel particle under reducing conditions was studied using a continuous flow-through reactor and compared to the dissolution of unirradiated fuel particle in the same condition. The irradiated fuel particle was leached for more than eight months. A fast 137Cs release was observed, corresponding to a labile fraction (“instant release fraction”). Then, a congruent leaching of 137Cs and 90Sr was measured corresponding to a matrix dissolution rate equal to 1.7 mg/m2/d. A slower release of 238Pu probably due to sorption phenomena is observed. Dissolution rates are 100-2000 times higher than for unirradiated material. We can conclude that alpha-radiolysis is responsible for this increase due to local oxic conditions. However we cannot exclude that due to irradiation the accessible surface area has been increased as well.

The coating of this irradiated particle was equally studied to determine the presence of different radionuclides and their leaching properties of leaching

Sample of natural Zr-silicate gel containing up to 13 wt.% U was characterized using scanning electron microscopy (SEM), X-ray powder diffraction (XRD), high resolution transmission electron microscopy (HRTEM) and electron microprobe analysis (EPMA) method. It was found that gel matrix is amorphous in general; however, it contains non-identified nanocrystallites. No separated oxide phases of U, Zr or Si were observed in a gel matrix. After sintering in air at 1400°C for 1 hour gel transformed largely into crystalline zircon, (Zr,U)SiO4. Uranium was not found in any other phases besides zircon. It was assumed that high chemical durability of natural Zr-P-U-Ti-silicate gel is caused by two competing processes which exist under self-irradiation conditions: 1) crystallization of the gel and 2) metamictization of the crystallized zircon and other phases.

Synthetic samples of Zr-silicate gel doped with Ce, U, Pu and Am were obtained and studied in comparison with natural samples. It was suggested to use artificial solid Zr-silicate gels for durable fixation of actinides for the goal of long-term or intermediate storage.