The pH-dependence of glass corrosion rates has a well-known U-shaped form with minima for near-neutral solutions. This paper analyses the change of U-shaped form with time and reveals that the pH dependence evolves even for solutions that have pH not affected by glass corrosion mathematically corresponding to a zero surface to volume ratio. The U(t) dependence is due to changes of concentration profiles of elements in the near-surface layers of glasses in contact with water and is most evident within the initial stages of glass corrosion at relatively low temperatures. Numerical examples are given for the nuclear waste borosilicate glass K-26 which is experimentally characterised by an effective diffusion coefficient of caesium DCs = 4.5 10-12 cm2/day and by a rate of glass hydrolysis in non-saturated groundwater as high as rh = 100 nm/year The changes of U-shaped form need to be accounted when assessing the performance of glasses in contact with water solutions.

The interface with reactive transport models used in performance assessment calculations is described to identify aspects of the glass waste form degradation model important to long-term predictions. These are primarily the conditions that trigger the change from the residual rate to the Stage 3 rate and the values of those rates. Although the processes triggering the change and controlling the Stage 3 rate are not yet understood mechanistically, neither appears related to an intrinsic property of the glass. The sudden and usually significant increase in the glass dissolution rate suggests the processes that trigger the increase are different than the processes controlling glass dissolution prior to that change. Application of a simple expression that was derived for mineral transformation to represent the kinetics of coupled glass dissolution and secondary phase precipitation reactions is shown to be consistent with experimental observations of Stage 3 and useful for modeling long-term glass dissolution in a complex disposal environment.

The necessity to a priori predict the durability of high level nuclear waste (HLW) glasses on extended time scales has led to a variety of modeling approaches based primarily on solution (leachate) concentrations. The glass composition and structure control the leachate and the gel compositions which in turn control what reaction products form: the leached layer is a hydrogel and reacts with the solution (leachate) to form secondary phases some of which cause accelerated glass dissolution which is undesirable. Glasses with molar excess alkali that is not bound to glass forming (Al,Fe,B)O4 structural groups in the glass resume accelerated leaching. The hydrogels of the glasses that resume accelerated leaching at long times contain excess alkali and the leachates contain excess strong base, [SB]ex. The [SB]ex further accelerates aluminosilicate gel aging into analcime with time. Glasses with no excess molar structural alkali do not resume accelerated leaching: the glass generates weak acids, [WA], in the leachate favoring hydrogel aging into clays. These data indicate that the gel layer transforms to secondary phases in situ in response to interactions with the chemistry of a continuously evolving leachate.

This study investigates the dissolution of CeO2, an isostructural analogue for UO2 and ThO2, which was synthesized to closely approximate the microstructure of a spent nuclear fuel matrix. Dissolution of CeO2 particles was performed in simplified solutions representative of saline, near-neutral and alkaline ground waters that may be encountered in geological disposal scenarios, and in acidic medium for comparison. The normalized mass loss of cerium was found to be significantly influenced by the formation of colloidal particles, especially in the near-neutral and alkaline solutions investigated. The normalized dissolution rate, RL(Ce), k (g m-2 d-1), in these two solutions was found to be similar, but significantly lower than in a nitric acid medium. The activation energies based on the normalized release rate of cerium, at 40°C, 70°C and 90°C in each solution, were in the range of 24 ± 3 kJ mol-1 to 27 ± 7 kJ mol-1, indicative of a surface-mediated dissolution mechanism. The mechanism of dissolution was postulated to be similar in each of the solutions investigated, and further work is proposed to investigate the role of carbonate on the CeO2 dissolution mechanism.

Geological disposal of HLW and spent nuclear fuel (SNF) in very deep boreholes is a concept whose time has come. The alternative – disposal in a mined, engineered repository is beset with difficulties not least of which are the constraints placed upon the engineered barriers by the high thermal loading. The deep borehole concept offers a potentially safer, faster and more cost-effective solution. Despite this, international interest has been slow to materialize, largely due to perceived problems with retrievability and uncertainty about the ability to drill accurate vertical holes with diameters greater than 0.5 m to a depth of 4-5 km. The closure of Yucca Mountain and the subsequent recommendations of the Blue Ribbon Commission have lead to a renewed interest in deep borehole disposal (DBD) and the US DoE has commissioned Sandia National Labs, working with industrial and academic partners (including the University of Sheffield), to undertake a program of R&D leading to a demonstration borehole being drilled somewhere in the continental USA by 2016.

In this paper, we focus on some of the key safety and engineering features of DBD including methods of sealing the boreholes, sealing and support matrices for the waste packages. Numerical modeling has, and continues to play, a significant role in expanding and validating the DBD concept. We report on progress in the use of modeling in the above contexts, paying particular attention to constraints on the engineering materials resulting from high heat loading.

Deep borehole disposal (or DBD) is now seen as a viable alternative to the (comparatively shallow) geologically repository concept for disposal of high level waste and spent nuclear fuel. Based on existing oil and geothermal well technologies, we report details of investigations into cementitious grouts as sealing/support matrices (SSMs) for waste disposal scenarios in the DBD process where temperatures at the waste package surface do not exceed ∼190ºC. Grouts based on Class G oil well cements, partially replaced with silica flour, are being developed, and the use of retarding admixtures is being investigated experimentally. Sodium gluconate appears to provide sufficient retardation and setting characteristics to be considered for this application and also provides an increase in grout fluidity. The quantity of sodium gluconate required in the grout to ensure fluidity for 4 hours at 90, 120 and 140°C is 0.05, 0.25 and 0.25 % by weight of cement respectively. A phosphonate admixture only appears to provide desirable retardation properties at 90°C. The presence of either retarder does not affect the composition of the hardened cement paste over 14 days curing and the phases formed are durable under conditions of high temperature and pressure.

Since 1994, the KHNP has developed a vitrification technology to treat the LILW generated from Korean nuclear power plant. To vitrify the LILW including combustible Dry Active Waste (DAW) and Ion Exchange Resin (IER) containing Zeolite, two borosilicate glasses are formulated. One of the formulated glass, DG2, is for the DAW vitrification solely and the other one, AG8W1, is for the blended wastes (DAW & IER) vitrification in a commercial vitrification facility in HanUl (former Ulchin) nuclear power plant. The physicochemical properties of the two glasses have been evaluated. To evaluate the processability of the glasses, the viscosities and electrical conductivities of the glass melts were measured in the laboratory within a temperature range between 950 and 1,350 degrees C, respectively. The liquidus temperatures of the glasses were evaluated using a gradient furnace for DG2 and data from heat treatment for AG8W1. The Mössbauer spectroscopy for AG8W1 was employed to evaluate the relations between the redox equilibria of iron. In addition, to verify the waste acceptance criteria for the final disposal of the vitrified forms, the compressive strengths of the vitrified forms were tested after an immersion test, a thermal cycling test, and an irradiation test. To verify the chemical durability of the glasses, several tests such as PCT, ISO, VHT, Soxhlet, MCC-1, and ANS16.1 were carried out. The PCT showed leach rates of B, Na, Li and Si were much less than those of the benchmark glass. The ISO test was performed at 90 degrees C for 1,022 days and Cumulative Fraction Leached of all elements in the glasses were analyzed. According to the VHT, the glasses had an outstanding chemical resistance under humid environment at 200 degrees C for 7 days. The Soxhlet leaching was performed on rectangular glass samples at 98 degrees C for 30 days. To analyze the forward dissolution rates of major glass elements, the MCC-1 was conducted at temperatures of 40, 70, and 90 degrees C for three weeks in pH buffer solutions ranging from pH 4 to 11. The processability of the glasses was in the desired ranges. And the product quality of the glasses met all regulatory guidelines. Using two glasses, the CCIM commissioning tests in the UVF were successfully performed and they showed good workability.

In a future Swedish deep repository for spent nuclear fuel, irradiated control rods from PWR nuclear reactors are planned to be stored together with the spent fuel. The control rod absorber consists of an 80% Ag, 5% Cd, 15% In alloy with a steel cladding. Upon in-reactor irradiation 108Ag is produced by neutron capture. Release of 108Ag has been identified as a potential source term for release of radioactive substances from the deep repository.

Under reducing deep repository conditions, the Ag corrosion rate is however expected to be low which would imply that the release rate of 108Ag should be low under these conditions. The aim of this study is to investigate the dissolution of PWR control rod absorber material under conditions relevant to a future deep repository for spent nuclear fuel. The experiments include tests using irradiated control rod absorber material from Ringhals 2, Sweden. Furthermore, un-irradiated control rod absorber alloy has been tested for comparison. The experiments indicate that the release of Ag from the alloy when exposed to water is strongly dependent on the redox conditions. Under aerated conditions Ag is released at a significant rate whereas no release could be measured after 133 days during leaching under H2.

An extension of the Swedish final repository for short-lived radioactive waste (SFR) is planned and a safety assessment has been performed as part of the licensing process. Within this work, steps have been taken to advance the modelling environment to better integrate its individual parts. It is desirable that an integrating modelling environment provides the framework to set up and solve a consistent hierarchy of models on different scales. As a consequence, the consistent connection between software tools and models needs to be considered, related to the full assessment domain. It should also be possible to include the associated geometry and material descriptions, minimizing simplifications of conceptual understanding.

The usefulness of the analysis software Comsol Multiphysics as component of an integrating modelling environment has been tested. Here, we present two examples of hierarchical models. Consistent properties and boundary conditions have been extracted form regional hydrogeology and surface hydrology models when setting up repository scale models. CAD models of the repository have been imported into the analysis software, representing tunnel systems and storage vaults with engineered structures and barriers. Data from geographic information systems such as digital elevation models and geological formations have been also directly implemented into model geometries.

The repository scale hydrology models have provided a basis for further developments focussed on the modelling of coupled processes. An interface between Comsol Multiphysics and the geochemical simulator Phreeqc has been developed to support reactive solute transport studies. An important test case concerns radionuclide transport in a 3D, near-surface model of a catchment area. The dynamic surface hydrology has been simulated with MIKE SHE and connected to Comsol Multiphysics and Phreeqc for detailed hydro-geo-chemical modelling of radionuclide migration through soils and sediments.

Integrated safety assessment methodology that analyzes radionuclide migration reflecting the spatial and temporal changes of disposal systems was developed for a geological disposal site with uplift and denudation, and then some case analyses for an assumed site were carried out. The combination of uniform uplift and denudation has the largest effect on the radionuclide migration because the ground water flow velocity increases with decreasing depth from the ground surface. In the case without denudation, tilted uplift has more effect than uniform uplift because flow velocity in tilted uplift increases with increasing hydraulic gradient. The long-term change of the geological structures including the uplift and denudation, the hydraulic conditions, and the recharge and outlet of the ground water around a candidate site should be carefully investigated to determine the appropriate the place, depth and layout of the repository.

In order to better understand the micrometer-scale structure of rock and its transport properties which arise from it, seven monomineral samples from two sites (Olkiluoto and Sievi, Finland) were studied with micro- and nanotomography and scanning electron microscopy. From the veined gneiss of Olkiluoto we studied biotite, potassium feldspar, plagioclase (composition of oligoclase) and cordierite, and from Sievi tonalite biotite and two grains of plagioclase (albite). These minerals were the main minerals of these samples. Samples were carefully separated and selected using heavy liquid separation and stereomicroscopy, their three dimensional structure was imaged using X-ray tomography, and their precise mineral composition was determined using scanning electron microscopy and energy-dispersive X-ray spectroscopy (SEM/EDS). The micrometer-scale mineral structure of these samples was observed together with their pores and fissures, and alteration effects were identified whenever applicable. Nanotomography combined with SEM analysis was concluded to be a good tool for analyzing effects of alteration in monomineral samples.