Two recent ongoing major projects at the Grimsel Test Site (GTS) (www.grimsel.com) that were initiated to simulate the long-term behaviour of radionuclides in the repository near-field and the surrounding host rock are presented: the Colloid Formation and Migration (CFM) project, which focuses on colloid generation and migration from a bentonite source doped with radionuclides and the Long-Term Diffusion (LTD) project, which aims at in-situ verification and understanding of the processes that control the long-term diffusion of repository-relevant radionuclides. So far, the CFM project has principally involved: development and implementation of a state-of-the-art sealing concept to control hydraulic gradients in a shear zone to imitate repository-relevant conditions; extensive laboratory studies to examine bentonite erosion and colloid formation in a shear zone; and, development of models to estimate colloid formation and migration. The next stage will be to assess the behavior of bentonite colloids generated from a radionuclide spiked bentonite source-term emplaced into the controlled flow field of the shear zone. This will be coupled with further extensive laboratory studies in order to refine and evaluate the colloid models currently used in performance assessments. The LTD project consists of: a monopole diffusion experiment where weakly sorbing and non-sorbing radionuclides (3H, 22Na, 131I, 134Cs) have been circulating and diffusing into undisturbed rock matrix since June 2007; experiments to characterise pore space geometry, including determination of in-situ porosity with 14C doped MMA resin for comparison with laboratory derived data; a study of natural tracers to elucidate evidence of long-term diffusion processes; and, an investigation of the in-situ matrix diffusion paths in core material from earlier GTS experiments. Future experiments will focus on diffusion processes starting from a water-conducting feature under realistic boundary conditions.

The behavior of actinides in a deeply buried radioactive waste cannot be predicted from evidence of their movement in geological environments in the geological past because these elements do not occur naturally in appreciable concentration. A useful means of reducing the uncertainty is to observe the behavior, both past and present, of chemical analogues in geological environment. In this sense, the rare earth elements (REE) have been used to predict the behaviour of actinide series elements in water-rock systems on account of the similar valence and ionic radii and high similarity in its electronic structure.

This paper describes the factors that control REE signature in the fracture fillings of an ancient uranium mine (dug granite) located in the Central-Iberian Zone of the Hesperian Massif (Caceres, Spain).

The study of REE distribution in the fracture fillings provides interesting information about the REE distribution in the hydrothermal fluids that flow through the system. In order to obtain this information, it is necessary to evaluate the control of different minerals in the total REE content of the fracture fillings.

The fracture fillings display a large mineralogical variety dominated by K-feldspars, quartz, plagioclase, micas, monazites, xenotimes, apatites, uraninites, inherited of the granites, and carbonates, quartz, apatites, phyllosilicates, sulphides and Fe-oxihydroxydes, neoformed in successive hydrothermal stages.

In general, the fracture fillings present different kind of REE patterns, just as positive or negative Eu anomalies. The REE final distribution in the fracture fillings is controlled by the presence of REE-minerals, both inherited and neoformed, and of carbonates, with great capacity to fractionate HREE in the system. On the other hand, the chemical composition of the carbonates and accessory minerals is responsible, as a last resort, for the Eu anomalie (positive or negative) observed in the fracture fillings.

In Japan, uplift/erosion scenarios must be analysed even if they occur far in the future, as no assessment cut-off times have yet been defined. For this purpose, an argumentation method is developed to allow sensible scenarios to be constructed. The consequences of erosion of the repository may be better estimated in terms of radionuclide fluxes and these compared with those of naturally occurring radionuclides. This paper discusses procedures to derive relevant conceptual models and resultant analyses in a credible manner, which illustrates the effectiveness and robustness of the HLW disposal system.

A quantitative risk assessment methodology is outlined for the source term for the proposed geologic repository for high-level nuclear waste at Yucca Mountain, Nevada. The methodology involves construction of a logical event tree to identify scenarios for the success or failure of the waste isolation system. Uncertainties in the intervening events between initial conditions and consequences can be quantified using probability distributions for steps along each scenario path. Likelihood is quantified according to the frequencies for each split fraction in the event tree. The result is a calculated probability of frequency curve for the release rate(s) of the modeled radionuclide(s).

According to recent seismic observation records, there are some cases where unexpectedly large seismic motion was observed deep underground and that was larger than at the surface. The factors influencing such phenomena are assumed to be deep geological structures with topographic irregularity, velocity structure and non-linearity of subsurface layers. These factors should be taken into account in the earthquake-resistant design of a geological repository. The influence of a deep underground geological structure with topographic irregularity on ground motion has been studied and it has been confirmed that such a structure have a significant impact on ground motion and the constructive interference of waves may result in strong earthquake ground motion in the vicinity of a structural boundary deep underground.

Japan is located at the converging plate boundaries and is one of the most earthquake-prone zones in the world. In order to ensure the safety of a geological repository against earthquakes, a site with a high possibility of direct destruction by active faults should be excluded, and the relationship between the characteristics of seismic ground motion, subsurface structures, geological disposal system and the propagation characteristics of seismic ground motion should be fully investigated. Earthquake-resistant design based on the latest technology is also very important for ensuring the safety of a geological repository.

Following rapid improvement of seismic observation networks after the Hyogoken Nanbu earthquake in 1995, numerous seismic observation records have been obtained in the vicinity of large earthquakes. According to these seismic observations, some phenomena that might affect the safety of a geological disposal system have occurred. Some earthquakes occurred in the areas where active faults had not been identified, while some records showed that seismic motion in the deep underground environment was greater than that at the surface. We have identified the implications from the latest information concerning large earthquakes for the geological disposal program. This study made it clear that detailed investigation incorporating state-of-the-art technologies could reduce the likelihood of missing active faults to an extremely low level and a more practical analysis of seismic ground motion could be achieved by taking the latest information into account.

As a part of the optimization study for achieving the highest possible Hanford and Savannah River Site waste loading into acceptable borosilicate glasses, thirty glass compositions were selected for testing at KRI. These thirty test matrix glasses were designed to augment existing data within the compositional regions of interest with relatively high concentration of Al2O3 between 10 and 20 wt%.

This paper reports experimental data on liquidus temperature (TL) and crystallization behavior of all synthesized glasses as well as durability of quenched and heat-treated glasses. The results of this study will be used to develop glass formulations for specific DOE waste streams to maintain or meet waste loading and/or waste throughput expectations while satisfying critical process and product performance related constraints.

A new type of flow-through test method using micro-reactor was developed and applied to measurement of the dissolution/alteration kinetics for a Japanese type of simulated HLW glass, P0798. In this test method, a face of coupon shaped glass specimen (30mm × 10mm × 4mm size) is in contact with a micro-channel (20mm length, 2mm width, 0.16mm depth) constructed on a PTFE (Teflon®) plate, and a solution is injected into the inlet of micro-channel at a constant rate. The injected solution, which flows through the micro-channel reacting with the glass to the outlet, is retrieved at certain intervals to be analyzed for determination of the glass dissolution/alteration rate. After the test, the glass specimen removed from the micro-reactor is subjected to surface analyses. This test method has major features as follows, 1) any controlled solution condition can be provided over the test duration, 2) a relatively high S/V ratio can be provided by use of micro-reactor in spite of using coupon shaped glass specimen, which results in precise and consistent analyses of both the solution and the reacted glass surface, 3) the test apparatus is simple with compact size and easy operation, which allows a flexible setup of test conditions. By use of this test method the dissolution/alteration rate for P0798 glass was measured as a function of pH, temperature, and time, and the results indicated that this test method is applicable and suitable for evaluation of the dissolution/alteration kinetics.

The objective of this study was to experimentally measure the properties and performance of a series of glasses with compositions that could represent high level waste Sludge Batch 5 (SB5) as vitrified at the Savannah River Site Defense Waste Processing Facility. These data were used to guide frit optimization efforts as the SB5 composition was finalized. Glass compositions for this study were developed by combining a series of SB5 composition projections with a group of candidate frits. The study glasses were fabricated using depleted uranium and their chemical compositions, crystalline contents and chemical durabilities were characterized. Trevorite was the only crystalline phase that was identified in a few of the study glasses after slow cooling, and is not of concern as spinels have been shown to have little impact on the durability of high level waste glasses. Chemical durability was quantified using the Product Consistency Test (PCT). All of the glasses had very acceptable durability performance. The results of this study indicate that a frit composition can be identified that will provide a processable and durable glass when combined with SB5.

The predictability of models describing long-term nuclear glass behavior in a geological repository can be tested by means of natural or archaeological analogs. This study covers fractured archaeological glass blocks from a shipwreck discovered near the Mediterranean island of Embiez (France). The blocks were examined mainly because of their morphological analogy with nuclear glasses. Fractured after production (as in the case of nuclear glass), these blocks had been leached for 1800 years in seawater. The laboratory investigation led to the development and subsequent validation on archaeological objects of a geochemical model capable of accurately simulating the coupling between chemistry and transport to account for the alteration state of the cracks according to their geometric characteristics. Laboratory experiments allowed us to determine the kinetic and thermodynamic parameters for modeling glass alteration. The model was then tested against short-term experiments before simulating the crack alteration over 1800 years. We show that cracks in the outer regions of the block are the most severely altered because of rapid solution renewal, whereas internal cracks are very slightly altered because of a rate-limiting effect of water transport due to the formation of secondary phases. This study also establishes a direct link between data obtained at lab scale and the long-term evolution of a complex system in a natural environment, indicating that the key phenomena have been identified experimentally. The analogous behavior of archaeological and nuclear glass during leaching experiments and the similarities in their crack networks allow us to consider applying the model to nuclear glasses under geological repository conditions. This study clearly shows that the internal crack network does not play a major role in the overall long-term alteration of archeological glass blocks. The issue of the transposition studies will be to determine whether this conclusion can be generalized to nuclear glasses.

In 2008, AREVA NC Industrial Vitrification of High-Level Liquid Waste blows out its 30th candle, with always two main objectives during all the time: containment of the long lived fission products and reduction of the final volume of waste. During all this time AREVA with the French Atomic Energy Commission (CEA) developed and use in their industrial installations a selection of borosilicate glass that have been demonstrated as the most suitable containment matrix for waste from spent nuclear fuel. Consistent and long-term R&D programs associated to industrial feed back from operation have enabled continuous improvement of the process: throughput and waste loading factor enhancement. The Vitrification Process used and currently implemented in the AREVA facilities will be described.

Lanthanide-Borosilicate (LaBS) glass capable to dissolve up to ∼10 wt.% PuO2 is designed for immobilization of plutonium-bearing wastes. The sample of LaBS glass with a target chemical composition (wt %): 9.0 Al2O3, 11.8 B2O3, 12.2 Gd2O3, 6.3 HfO2, 17.2 La2O3, 13.6 Nd2O3, 9.5 PuO2, 18.1 SiO2, 2.3 SrO was prepared from PuO2 powder and mechanically activated mixture of chemicals at 1500 °C. The obtained product was visually homogeneous. Xraydiffraction of the as-prepared glass showed that it mostly consists of a vitreous phase withsmall amounts of crystalline PuO2 (or PuO2-HfO2 solid solution with minor HfO2), britholite andan oxide with a fluorite structure and a composition close to GdHfO1.75. The crystalline fractionincreased after storage for ∼1 year. Magnitude of the FT of EXAFS spectrum at Pu LIII edgeshows that the peak due to first coordination sphere is much more intense than that of the secondshell. This indicates that though some Pu entered crystalline phases (mainly distorted PuO2), itsmajor fraction remained in the vitreous phase. Fit of the spectrum (R-factor = 0.02) gives thefollowing distances: R (Pu-O1)1 = 1.98 (σ = 0.04), R (Pu-O1)2 = 2.18 (σ = 0.04), R (Pu-O1)3= 2.35, R (Pu-Pu) = 3.72 (σ = 0.04), R (Pu-O2) = 4.4 (σ = 0.06). Oxygen environment ofthe Pu4+ ions in the vitreous phase resembles axially squeezed tetragonal pyramid with acoordination number ∼5. The distances ∼2.35 and ∼4.4 correspond to the pairs Pu-O in thefirst and the second shells in the crystalline PuO2. The distance ∼3.72 corresponds to the Pu-Pu and/or Pu-M (M = Ln, Hf) distances. EXAFS spectra of Hf show that Hf is present mostly in thevitreous phase with major neighbors at 2.17 and ∼3.2 A.

The Sellafield Waste Vitrification Plant (WVP) immobilises highly active liquors produced during reprocessing of spent nuclear fuel by bonding the fission products as metal oxides into a borosilicate glass matrix. This provides a stable and durable waste form suitable for safe long term storage and ultimate disposal.

WVP was commissioned with feed from reprocessing of Magnox reactor fuel. This material is relatively low in fission product content per tonne of fuel, but contains significant Al and Mg from fuel cladding. WVP also routinely treats a blended feed made from a mixture of Magnox and Oxide reprocessing products. The Oxide fuel from Light Water Reactor (LWR) and Advanced Gas Cooled (AGR) power stations is of higher burnup and contains more fission products per tonne of fuel, also Gd and other process additives. Blending allows 25% incorporation of waste oxides by weight in glass to be achieved routinely.

Recent programmes of development work in WVP have been aimed at increasing incorporation rates for these feeds, to reduce the number of waste containers produced for disposal. Work has also focussed on increasing the throughput of WVP, to more rapidly treat current stocks of liquid reprocessing waste, both by increasing the feed rate and by improving the lifetime of key components to improve plant availability.

Future challenges for WVP include flowsheet changes to treat historic stocks of reprocessing wastes containing high U, Fe and Cr. Washout of solids from the base of waste storage tanks in preparation for decommissioning is also likely to give high Mo feeds. Development of flowsheet and glass formulation to accept these changes in feed composition will be a key objective of future work.

The effect of intensive electron radiation on viscous flow in silicate glasses is analysed and shown that it can result in a many orders of magnitude decrease of viscosity and stepwise decrease of activation energy of flow. Fluidisation or quasi-melting of glasses on intensive electron irradiation is caused by bond breaking via the radiation-chemical reaction ≡Si-O-Si≡ + e− Si-O + Si + (e−)′ which weakens the silicate glass network and leads tofive-fold coordination of oxygens around the silicon. An explicit equation of viscosity wasobtained for irradiated glasses as well as an equation for glass transition temperature. Theassessments of temperature increase by electron radiation show that radiation-inducedfluidisation of glasses can occur at minimal thermal effects. Radiation-induced fluidisation ofglasses can result in nanoscale patterning effects caused by surface tension forces. Changes inthe viscous flow behaviour are also important in conditions of long-term irradiation for glassesused in nuclear installations as well as for nuclear waste glasses.

The Sellafield Waste Vitrification Plant (WVP) immobilises highly active liquid waste (HAL) arising from the reprocessing of spent nuclear fuel in the UK. In order to optimise WVP operations a full scale working replica of a WVP processing line, the Vitrification Test Rig (VTR), was constructed to processes non-active HAL simulants. Recently the VTR has been used to determine an operational envelope for the vitrification of HAL from Magnox reprocessing at a waste oxide incorporation rate in glass of up to 35wt% (compared to a “standard” incorporation rate of 25wt%). This paper discusses the differences in operating conditions necessary to achieve acceptable waste throughput at the increased incorporation rate. The chemical durability of the resulting vitrified product is also discussed, along with the formation of secondary phases, and a comparison is drawn between 35wt% incorporation glasses and products made at the standard 25wt% incorporation.

Sellafield Ltd operates a Waste Vitrification Plant (WVP) to immobilise the arisings from the reprocessing of spent nuclear fuel. Washout of solids from the base of waste storage tanks in preparation for decommissioning is likely to produce feeds enriched in molybdenum to the WVP. Vitrification of such feeds in the borosilicate glass formulation currently used by the WVP for vitrification of reprocessing waste has been investigated to determine the maximum achievable loading of MoO3.

The vitrification of molybdenum in the absence and presence of reprocessing waste was studied. A number of glasses were manufactured in the laboratory containing various waste loadings. The resultant glasses were examined both visually and under the scanning electron microscope for the presence of any phase separation. Additional aluminium was added to the glasses manufactured in the absence of reprocessing waste to improve the durability of the glass. In borosilicate glass containing 3.5 wt% Al2O3 the onset of a molybdenum phase separation was observed in glasses containing 2.6 wt% MoO3. In the presence of Magnox reprocessing waste, phase separation was observed when the product contained >3.8 wt% MoO3. Soxhlet durability testing of a selection of the glasses manufactured was carried out. The Soxhlet durability of glasses in the absence of phase separation was good.

Static leach tests were conducted for simulated low-level radioactive waste (LLW) glass in deionized water at 90 °C for up to one year to investigate the dissolution mechanism of LLW glass. Widely studied leaching behavior of high-level radioactive waste (HLW) glass is referred in discussing the dissolution mechanism. LLW glass is characterized by higher sodium (Na) and aluminum (Al) contents than HLW glass, about twice as high as R7T7, with its SiO2 content close to HLW glass. Powdered simulated LLW glass of three different chemical compositions was tested with the glass-surface-to-water-volume ratio of 2,000 m−1. The release rates of boron (B), widely used as an indicator of dissolution for HLW glass, decreased with time during leaching, as commonly observed in similar tests for HLW glass. The pH of the leachate was stable around 11.3 - 11.6, which is higher than those in similar tests for HLW glass by one pH unit or more. The concentrations of Al in the leachates were higher compared to data for HLW glass by two orders of magnitude. The high concentration seems to be caused by higher pH. In the leachate condition of the present tests, a zeolitic mineral (analcime) is thermodynamically more stable than amorphous silica (SiO2(am)) which is known to control the concentration of dissolved silica (Si) with respect to HLW glass. The present results imply that dissolution of the LLW glass is accompanied with formation of analcime under virtually closed systems such as geological repository where the groundwater flow rate is quite low.

We applied a new type of flow-through test method using micro-reactor consisting of a simple test apparatus with compact size to measurement of the dissolution rate of a Japanese type of simulated waste glass (P0798 glass). In this test method, a solution flows through a micro-channel (20 mm length, 2 mm width, 0.16 mm depth) in contact with a face of coupon shaped glass specimen, and the output solution is retrieved at certain intervals to be analyzed for determination of the glass dissolution rate. By using this test method the initial dissolution rate of glass matrix or forward dissolution rate was measured as a function of pH (3 to 11) and temperature (25°C to 90°C). The present test results indicated that the initial dissolution rate has ‘V-shaped’ pH dependence, and the effect of pH on the dissolution rate decreases with increase in temperature similar to the results measured by using the Single-pass flow-through (SPFT) method. The present test results also indicated that the dissolution of B is controlled by diffusion process and that of Si is controlled by surface reaction process.

Glass matrices were selected to immobilize fission products because glass is capable of chemically incorporating a wide spectrum of elements within a single matrix. Some of these elements can be found at different oxidation states. The redox equilibrium constants of multivalent species can be used to develop thermodynamic models for a better description of nuclear glasses. Some of the multivalent species loaded in nuclear glass, such as iron and sulfur, have already been a subject of investigation by conventional glassmakers or geochemists in the earth sciences. Other redox species more specifically related to nuclear glass, including cerium and ruthenium, have also been investigated. These studies have demonstrated the advantages of using electrochemical techniques, voltammetry and potentiometry, to determine the equilibrium constants. Oxygen potential measurements are also particularly suitable for characterizing the redox state of the multivalent dissolved species in molten glass.

The metallurgical industry of ferrous and non-ferrous metals produce huge amount of wastes. In Slovakia, the factory for nickel production was closed in 1993, but around 5.5 kt of wastes remain in a dump. This waste was used as a model sorbent of heavy metals (Cu, Cd, Co) from wastewater treatments. The TCLP (Toxicity Characteristic Leaching Procedure) test of precipitated heavy metals on waste materials has confirmed the necessity of waste stabilization. The microwave vitrification was applied because of a high content of iron in waste. After vitrification, the chemical durability and microhardness by indentation have been tested in the glassy and glassceramics obtained.