The obtention of iron aluminium silicate- phosphate glasses including U3O8 oxide has been carried out and the study of crystalline phases formation. A serie of these original glasses have been investigated by thermal cycles at 780? C for promoting the nucleation and growth of crystalline phases of iron aluminium phosphates including U3O8 in the 5.1- 13.71 wt% range. Thermal expansion and DTA/G analysis have been carried out showing behaviour trend to crystallization of these glasses. The glass with higher uranium oxide content shows liquid phase in phase separation which change of microstructure during HREM observations. It seems produced a pulsed order- disorder processing which takes place under the electron beam. This effect first observed in TEM-HREM on phase separation droplets is discussed.

Phase formation mechanisms associated with the vitrification of Savannah River Site (SRS) Sludge Batch 4 (SB4) high level waste surrogate with high iron and aluminum contents were studied by infrared spectroscopy (IRS), X-ray diffraction (XRD) and optical microscopy. Two mixtures at 50 wt.% SB4 waste loading were prepared as slurries with a water content of ∼50 wt% using a waste surrogate and commercially available Frit 503-R4 (Li2O – 8 wt%, B2O3 – 16 wt%, SiO2 – 76 wt%) or mixture of chemicals (LiOH·H2O, H3BO3, SiO2). The mixtures were air-dried at a temperature of 115 °C and heat-treated at 500, 700, 900, 1000, 1100, 1200, and 1300 °C for 1 hr at each temperature. IR spectra and XRD patterns of the products heat-treated at each temperature were recorded. In both the mixtures phase formation reactions started at low temperatures and yielded intermediate phases (sodalite, pyroxene-type, nepheline), and the reactions were mostly completed within the temperature range between 1000 and 1100 °C. The glassy materials prepared at 1200 and 1300 °C were composed of vitreous phase and magnetite/trevorite type spinel.

Solidification of low-activity wastes with cementitious materials is a widely accepted technique that contains and isolates waste from the hydrologic environment. The radionuclides I-129, Se-75, Tc-99, and U-238 are identified as long-term dose contributors. The anionic nature of these radionuclides in aqueous solutions allows them to readily leach into the subsurface environment. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages via mass flow and/or diffusion into the surrounding subsurface environment. Assessing the long-term performance of waste grouts for encasement of radionuclides requires understanding the: 1) speciation and interaction of the radionuclides within the concrete wasteform, 2) diffusion of radionuclide species when contacted with vadose zone porewater or groundwater under environmentally relevant conditions, and 3) long-term durability and weathering of concrete waste forms. An improved understanding of the interactions of long-lived radionuclides in cementitious matrices will improve predictions of the long-term fate of these sequestered contaminants. An integrated laboratory investigation has been conducted including a: 1) multifaceted spectroscopic investigation to interrogate the speciation and interaction of radionuclides within concrete wasteforms, 2) solubility tests to quantify the stability of solid phases identified as radionuclide-controlling phases, 3) quantify the diffusion of radionuclides from concrete wasteforms into surrounding subsurface sediment under realistic moisture contents (4%, 7%, and 15% by weight moisture content), 4) quantify the long-term durability of concrete waste forms as a function environmental parameters relevant to depository conditions, and 5) identify the formation of secondary phases or processes (microcracking) that influence radionuclide retention. Data obtained from this investigation provides valuable information for understanding the speciation, behavior, and fate of radionuclides immobilized within concrete wasteforms under vadose zone conditions and underscores the necessity for robust, multi-disciplinary performance assessments for concrete waste forms.

Graphite is used as the neutron moderator and reflector in many nuclear reactors. Obsolete graphite nuclear reactors are put out of operation, leading to formation of a large quantity of radioactive graphite waste.

It is proposed that irradiated reactor graphite is processed by high-temperature chemical oxidation in salt melts with an oxidant, which is part of the salt melt, leading to formation of exhaust gases: gaseous compounds of carbon and oxygen (CO2 and CO).

This study deals with carbon oxidation and physical-chemical transformations of radioactive elements during the interaction between graphite waste of the atomic power industry and salt melts. The method of thermodynamic simulation is used. The carbon melt decreases the transfer of radionuclides to the gaseous phase as compared to incineration of graphite in the atmosphere.

A scenario analysis was performed for microbial effect on the performance assessment (PA) of the high-level radioactive waste disposal system. Based on review of recent literature on PA and some results of our experiment, possible scenario of microbial effects on the Japanese high-level radioactive waste disposal system were analyzed. The result of the analysis shows that the microbial effect on groundwater evolution, which is caused by the microbial redox reaction, is one of the most important issues in assessing PA. The evolution not only affects radionuclide migration but also the life-time of metal overpack corrosion.

To assess the microbial effect on groundwater evolution, a numerical estimation code is required, with reliable parameters for analysis of microbial activities used in the code. A biological parameter database has been developed to give the growth and metabolism of the six typical microbial groups to be used in the estimation code. For this purpose, about 573 data from 95 papers were examined to obtain effective data for disposal. The database is composed of significant data like specific growth rate, maximum growth rate, constant decay rate, and experimental condition for model assessment.

Implementation of full thermodynamic models in performance assessment (PA) calculations (large domain and very large timescales) is practically unfeasible due to excessive computational times. The complex competitive sorption processes of radionuclides are often simplified.

In this paper, such simple sorption models (i.e. linear Kd and Langmuir isotherms) were compared with more complex thermodynamic models for a reduced geometry and relatively short timescales (compared to traditional PA calculations).

Within the context of a safety case the value of this study is twofold. Primarily, it provides support information on how to choose adequate parameter values in a consistent simplified analysis for compliance with safety criteria. For the cases studied in this paper it became clear that a linear model is sufficient to represent sorption, but a proper choice of the Kd values is critical. Secondly, the comparison of results from these compliance calculations with more realistic analyses demonstrates quantitatively the safety margin implicitly present in PA calculations through model abstraction.

Low level nuclear waste (LLW) will be stored in a series of modular vaults at the UK's principal LLW repository located near the village of Drigg in west Cumbria. The design of the new vaults is reviewed in the context of likely volume, composition and packaging of future waste streams; planning and regulatory considerations; and the site's geological and hydrogeological setting. Site construction work for the first of the new vaults (Vault 9) commenced in September 2008 and is anticipated to be substantially complete by the end of 2009 with future vaults constructed as required. The vaults are designed as a series of containment cells comprising a composite basal and sidewall lining system with an engineered capping system. Provision has been made for an extensive, deep cut-off wall as part of the longterm containment system for the whole repository site, which includes areas where LLW was historically tumble tipped into a series of trenches. Constraints on the design of the vaults include a complex hydrogeological setting controlled by two groundwater bodies influencing maximum excavation depth; geometrical limitations on the height, level and plan extents to which LLW could be stacked within the vaults; and strict limitations on the movement of construction materials by road.

A sensitivity analysis was conducted to evaluate the impact of silicon migration through buffer material on the lifetime of vitrified waste. The results indicate that the lifetime depends on a combination of the dissolved glass fraction in the non-steady-state phase controlled by the silicon pore diffusion coefficient (Dp) and the silicon distribution coefficient (Kd) in the buffer material and the steady-state dissolution rate defined by Dp and the groundwater flow rate (Q) in the excavation disturbed zone. In the case where the glass dissolution rate reaches the steady-state dissolution rate, the sensitivity of the lifetime to Dp and Q varies according to the magnitude relationship between Dp and Q. We also discuss the impact on the lifetime of glass hydration, which proceeds simultaneously with glass matrix dissolution. The results show that glass hydration is less important for the lifetime than glass matrix dissolution in an open system and it can be concluded that silicon migration through the buffer material will be an important process for estimating the lifetime of the vitrified waste. A preliminary calculation of the long-term waste behavior with realistic assumptions indicates the importance of the silicon migration parameters Kd and Dp, which control the dissolution behavior of the vitrified waste in the non-steady-state phase, for evaluating radionuclide release.

Two fundamental concerns must be addressed when attempting to isolate low-level waste in a disposal facility on land. The first concern is isolating the waste from water, or hydrologic isolation. The second is preventing movement of the radionuclides out of the disposal facility, or radionuclide migration. Particularly, we have investigated here the latter modified scenario.To assess the safety for disposal of radioactive waste-concrete composition, the leakage of 137Cs from a waste composite into a surrounding fluid has been studied. Leakage tests were carried out by original method, developed in Vinca Institute [1,2,3,4,5]. Transport phenomena involved in the leaching of a radioactive material from a cement composite matrix are investigated using three methods based on theoretical equations [6,10]. These are: the diffusion equation for a plane source an equation for diffusion coupled to a first-order equation, and an empirical method employing a polynomial equation‥ The results presented in this paper are from a 25-year mortar and concrete testing project that will influence the design choises for radioactive waste packaging for a future Serbian radioactive waste disposal center.

The consistency of migration parameters obtained by different methods is examined for strontium in Boom Clay. In addition to a previous in-situ percolation experiment and electromigration experiments (using an electrical field as driving force), a lab percolation test and two through-diffusion tests (on clay cores of different lengths) have been performed.

All experiments lead to a robust value for the apparent diffusion coefficient of strontium between 5 ? 10−12 m2/s and 1 ? 10−11 m2/s.

No reliable value is obtained for the product ηR of the diffusion accessible porosity and theretardation factor R. For one of the through-diffusion experiments, the concentration decrease inthe inlet as a function of time was fitted simultaneously with the strontium profile in the clay,leading to a ηR value around 400. Estimating ηR at the inlet/clay interface as the ratio of thebulk concentration and the pore water concentration leads to a similar value. The valuesobtained from the strontium profiles in the percolation and electromigration experiments are inline as well. However, fitting for both through-diffusion tests the outlet concentrations versustime leads to very low and unrealistic values of ηR (0.15 and 0.5).

Modeling shows that the transport of strontium in Boom Clay cannot be satisfactorilydescribed with the current assumptions (like e.g. instantaneous linear sorption equilibrium andboundary conditions), meaning that there is a problem to obtain a reliable value for the product ηR for strontium in Boom Clay.

Hydrogen peroxide is considered as one of the main oxidants formed due to the radiolysis of water. In a spent nuclear fuel repository, it is necessary to establish the interaction of hydrogen peroxide with the elements constituting the repository. The objective of this work is to study the consumption of hydrogen peroxide via reaction with the elements of the canister.

In this sense, two different series of experiments were conducted, with iron steel an magnetite, respectively. Each series consisted on three different experiments that contained a coupon of the solid and different hydrogen peroxide concentrations (10−4 mol·dm−3, 10−5 mol·dm−3 and 10−6 mol·dm−3). Hydrogen concentration in solution was measured at different intervals of time by means of chemiluminescence. At the end of the experiments, the coupons were studied by X-ray Diffraction (XRD) and Scanning Electron Microscopy (SEM) in order to determine the possible secondary solid phases formed on the coupons.

In both series of experiments, a decrease of the hydrogen peroxide concentration in solution with time was observed. The determined consumption rates increased with hydrogen peroxide concentration and were higher in steel than in magnetite. The reaction orders relative to hydrogen peroxide concentration were very close to the unity on both solids.

The study of the carbon steel coupons by SEM at the end of the experiments showed that they were more attacked at higher hydrogen peroxide concentrations. On the other hand, the XRD measurements in the steel coupons showed that lepidocrocite (γ-FeO(OH)), and magnetite (Fe3O4) were formed on the coupon as iron secondary solid phases.

Non-cementitious grouts have been tested in Olkiluoto for the sealing of fractures with the small hydraulic apertures. A promising non-cementitious inorganic grout material for sealing the fractures with the apertures less than 0.05 mm is commercial colloidal silica called silica sol. The potential relevance of colloid-mediated radionuclide transport is highly dependent on their stability in different geochemical environments. The objective of this work was to follow stability of silica sol colloids in low salinity Allard and saline OLSO reference groundwater (pH 7–11) and in deionized milliQ water. Stability of silica sol colloids was followed by measuring particle size distribution, zeta potential, colloidal and reactive silica concentrations. The particle size distributions were determined applying the dynamic light scattering (DLS) method and zeta potential based on dynamic electrophoretic mobility. The colloidal silica concentration was calculated from DLS measurements applying a calibration using a standard series of silica sol. Dissolved reactive silica concentration was determined using the molybdate blue (MoO4) method.

These results confirmed that the stability of silica colloids dependent significantly on groundwater salinity. In deionized water, particle size distribution and zeta potential was rather stable except the most diluted solution. In low salinity Allard, particle size distribution was rather constant and the mean particle diameter remained less than 100 nm. High negative zeta potential values indicated the existence of stable silica colloids. In saline OLSO, particle size distribution was wide from a nanometer scale to thousands of nanometers. The disappearance of large particles, decrease in colloidal particle concentration and zeta potential near zero suggest flocculation or coagulation. Under prevailing saline groundwater conditions in Olkiluoto silica colloids released from silica sol are expected to be instable but the possible influence of low salinity glacial melt water has to be considered.

Mineralogical changes of cement and bentonite accompanied with their interaction wereexperimentally studied by mixing granulated hardened cement paste and bentonite, and aging the mixture for91 days at 50° C. Mineralogical changes of cement and bentonite were identified by XRD. Hydratedcalcium-silicate phases (C-S-H), Ca(OH)2, ettringite and monosulfate were identified in the unalteredhardened cement. While Ca(OH)2 and monosulfate decreased with aging and disappeared after 91 days,calcite and katoite (Ca3Al2(SiO4)(OH)8) were formed concurrently. Montmorillonite, quartz (and/orchalcedony), clinoptilolite, plagioclase, calcite, analcime and pyrite were identified in the unaltered bentonite.The XRD pattern showed that diffraction intensities of these minerals decreased with aging. It seems thatthese primary minerals dissolved in the course of the alteration. C-S-H appeared in bentonite during the agingas secondary phases, indicating the participation of silicon dissolved from the bentonite and calcium from thecement formed the C-S-H. The formation of C-S-H that had been predicted by previous modeling studieswas confirmed by the present experiments.

In addition, diffusivity of tritiated water in mixed specimen with granulated hardened cement andbentonite was determined by a through-diffusion method. The effective diffusivity of tritiated water decreasedwith aging. The result suggests that the mass diffusivity in the interface of cement-bentonite system willdecrease with their interactions. The results of the diffusion experiments are qualitatively consistent with thediffusivity change in cement-bentonite systems predicted by some computational studies.

Carbon steel overpack will corrode by consuming oxygen introduced by repository construction after closure of repository and then will keep the reducing environment in the vicinity of repository. The iron corrosion products can migrate in bentonite as ferrous ion through the interlayer of montmorillonite replacing exchangeable sodium ions in the interlayer. This replacement of sodium with ferrous ion may affect the migration behavior in the altered bentonite not only for redox-sensitive elements but also the other ions. Therefore the authors have carried out electromigration experiments of potassium or rubidium with source of iron ions supplied by anode corrosion of iron coupon in compacted bentonite. Five to fifteen micro liter of tracer solution containing 3.3 M of KCl or 2.2 M of RbCl was spiked on the interface between an iron coupon and bentonite, which dry density was around 1.4 Mg/m3, before assembling. The iron coupon was connected as the working electrode to the potentiostat and was held at a constant supplied potential between - 600 and 300 mV vs. Ag/AgCl reference electrode for up to 8 days. Potassium could migrate faster and deeper in bentonite specimen than iron in each condition. On the other hand rubidium could migrate slower than iron. Migration velocity was a function of applied electrical potential and 8 to 14 nm/s for potassium, 5 to 10 nm/s for iron and 3 to 5 for rubidium, respectively. Dispersion coefficient was also a function of applied potential and 10 to 14 × 10−12 m2/s for potassium, 4 to 8 overv 10−12 m2/s for rubidium and 2 to 4 overv 10−12 m2/s for iron, respectively. Diffusion experiments were also carried out for comparison. Potassium and rubidium might migrate slightly slower in the altered bentonite by iron corrosion than in ordinary compacted bentonite.

Theoretical modeling of the decomposition in bcc Fe-Cu alloys has been performed using a combined approach which includes ab-initio calculations of the effective cluster interactions and statistical-mechanical (Monte Carlo) simulations. We showed that the effective Cu-Cu and Cu-vacancy interactions in the bcc Fe matrix have a strong dependence on the global magnetic state of iron. As a result, all the related thermodynamic properties of the alloys (such as solubility limit and diffusivity) are expected to have a pronounced non-Arrhenius temperature behavior, originated from variation of the global magnetization with temperature. We find that strong Cu-vacancy interactions in the bcc Fe matrix lead to a remarkable effect of vacancies on the Cu precipitation and significantly modify the alloy decomposition kinetics under irradiation.

Kinetics of formation of intrinsic colloids by tetravelent metal ions, namely, Th(IV), Hf(IV) and Pu(IV) have been studied using the dynamic light scattering technique. The milli-molar solutions of Hf(IV) and Th(IV) were prepared at varying pH (2-4) and ionic strength (0-0.01 M) and the samples were subjected to dynamic light scattering measurements at regular intervals for few days. The results showed that the size of intrinsic colloids varies with pH and ionic strength. In the case of Pu(IV) the concentration of aqueous solutions was 4.5×10−5 M.

Chemical conditions and mass transport properties of engineered barrier systems in TRU waste facilities would change with time due to the interaction of cement/bentonite materials. (‘TRU waste’ is one of categories of the radioactive wastes and contains a significant amount of alpha-emitting transuranic nuclides. In some countries, these wastes are classified into the Intermediate Level Waste (ILW).) Previous numerical model analyses to assess the long-term performance of engineered barrier systems in TRU waste repositories predicted to form Calcium Silicate Hydrate (C-S-H) species at the interface between the cementitious and bentonite materials. If C-S-H precipitates in the bentonite side of the boundary, mass transport in the bentonite buffer decreases and mineralogical alterations are expected to be restricted for a long period. The evidence of C-S-H precipitation in the bentonite side, however, still has not been identified in the former experimental studies. To improve the reliability of numerical analyses, immersion experiments were performed using contact samples of cementitious and bentonite materials, and X-ray absorption fine structure (XAFS) analysis was carried out to detect C-S-H precipitation at the contacting interface. Precipitation of C-S-H was confirmed from the obtained XAFS spectra. This result is one of the evidences to show the validity of the current numerical model analyses, which suggests that the bentonite buffer performance as an engineered barrier would be kept over a long period.

The aim of this study is to improve the understanding of the water infiltration within Bituminized Waste Products (BWP) and the associated phenomena such as the development of the porous layer, the matrix swelling or the water uptake kinetics. Two sets of leaching experiments have been performed on synthesized model samples that are constituted by inactive soluble and/or insoluble salts (40% wt) embedded in bitumen. Large samples were used to quantify the water uptake kinetics and to control the macroscopic swelling of the matrixes during experiments. Smaller samples were used to characterize the depth distribution of the infiltrated water and its motion properties by NMR techniques (1H-T1 relaxation times and 1D-NMR imaging). These techniques enabled us to study the influence of the matrix chemical composition on the water advance in depth as a function of the leaching time. Results show that insoluble salts have a significant influence on water transport while soluble salts entail the development of large porosities. Preliminary results obtained with an original method based on the measurement of NMR relaxation times along the water concentration in depth are presented. They illustrate the great potential of the approach to further improve our knowledge on the establishment of the aqueous network.

A coupled hydro-chemical-mechanical constitutive law for the Belgian Eurobitum bituminized waste is being developed by the International Centre for Numerical Methods and Engineering (Polytechnical University of Cataluña, Spain) to contribute to the study of the compatibility of Eurobitum with Boom Clay as a geological disposal environment. A large experimental programme is ongoing at SCK•CEN to support the development of a constitutive law for Eurobitum. Water uptake tests are being performed under different conditions to obtain insights in the parameters that influence the water uptake behaviour of Eurobitum. Furthermore, Environmental Scanning Electron Microscopy and microfocus X-ray Computer Tomography are used to characterize hydrated samples in order to improve the understanding of the water uptake processes.

The salt content, the distribution of the salt crystals, and the membrane efficiency in the Eurobitum samples affect the swelling and pressure increase rate. High membrane efficiencies and a large amount of hygroscopic salts inside the Eurobitum samples result in very high pressures when almost no swelling is allowed. The pressure in small inactive samples with 28 wt.% NaNO3has risen to ∼19 MPa after ∼3 years of hydration in nearly constant volume conditions. Slower pressure increase rates are being measured for samples with 6, 12, 18, and 33 wt.% NaNO3.

Spent nuclear fuel is in Sweden planned to be disposed of by encapsulating in waste packages consisting of a cast iron insert surrounded by a copper canister. The waste package is heavy. Throughout the manufacturing process from the extrusion/pierce-and-draw manufacturing to the final placement in the repository, the copper is subjected to handling which could introduce cold work in the material. It is well known that the creep properties of engineering materials at higher temperatures are affected by cold working.

The study includes creep testing of four series of cold worked, oxygen-free, phosphorus doped copper (Cu-OFP) at 75 °C. The results are compared to reference series for as-received material carried out in a recent study. Two series of copper cold worked in tension (12 and 24 %) and two series cold worked in compression (12 % parallel to creep load axis and 15 % perpendicular to creep load axis) were tested.

The results show that pre-straining in tension of copper leads to prolonged creep life at 75 °C. The creep rate and ductility are reduced. The influence on the creep properties increases with the amount of cold work. Cold work in compression applied along the creep load axis has no effect on the creep life or the creep rate. Nonetheless the ductility is still impaired. However, cold work in compression applied perpendicular to the creep load direction has a positive effect on the creep life. Cold work in both tension and compression results in a pronounced reduction of the initial strain on loading. Yet the high value of the area reduction, 90 %, is unaffected by the degree of cold work.