The sorption capacity of hydrotalcite (HT) and its calcined product (CHT) was evaluated for 131I− sorption from water solution and it was determined as a function of the calcining temperature. The radionuclide content was determined by γ-spectrometry. Solids were characterized by thermal analysis, X-ray diffraction (XRD), electron microscopy and Brunauer-Emmett-Teller (BET) analysis. For 0.1 M Na I solution, labeled with 131I−, sorption capacity was found to be 0.24 meq g−1 (7.2% of the anion exchange capacity, AEC). But, if the sample was previously calcined at 773 K and the HT structure destroyed, the sorption of I− increased considerably, up to 2.08 meq g−1 (63% of the AEC) and the HT structure was reconstructed. The 131I− sorption at very low concentrations (10−14M) was 0.04 × 10−14 meq of 131I− g−1 in the noncalcined HT, but for calcined samples at 773 K, the sorption increased to circa 0.97 × 10−14 meq g−1. Calcination temperature determines the surface area of the resulting mixed oxides, and that property seems to be the most important factor controlling the I− sorption. If the calcination temperature was increased to 873 K, the specific surface area of the oxide mixture increased and I− sorption increased as well, whereas calcination of HT at 973–1073 K resulted in a low surface area and a low I− retention.

The potential 14C (carbon-14, radiocarbon) flux from disposal of 14C containing waste into air is compared with the natural 14C emanation rate from soil in order to put the 14C hazard potential from disposal of this waste in perspective with the 14C exposure from cosmogenic origin. Chemical corrosion of neutron irradiated metals, steel and Zircaloy, is bounded by diffusion of water through a thermodynamically stable metal-oxide layer and dissolution of this metal-oxide in a nuclear plant. Many countries process radioactive waste for disposal using cementitious materials, an acknowledged end-point management technique for this waste. The metal-oxides are also stable when these waste forms are embedded in cementitious materials. The 14C release rate from this Zircaloy at these alkaline and reducing conditions is comparable to the natural 14C emanation rate from soil into air. Neutron irradiated graphite and spent ion exchange resins are chemically inert and therefore other release mechanisms need to be assumed. Radiolytic corrosion is used to determine the 14C release rate from this graphite. Moreover, ion exchange—with ingressing anionic species that have a higher affinity than contained anionic 14C—is proposed as a release mechanism for these resins.

A preliminary evaluation of gaseous radiocarbon (14C) behavior under geological repository conditions for Italian radioactive high level waste-long-lived and intermediate level waste disposal has been performed. Although in Italy there is still no defined project for a geological disposal facility, current work may support future safety assessment studies for a hypothetical future repository in deep salt rock. In the Italian context of radioactive waste, the percentage of 14C bearing waste to be disposed in a possible geological repository is low; irradiated graphite is the most important radiological source. Data about the radiological inventory has been collected to simulate production and migration of gaseous 14C in a hypothetical geological repository. Three different conceptual models have been developed and simulated. The first model has considered a preliminary evaluation of the radiological impact referred to the whole inventory; the second and third model have evaluated the impact only due to the irradiated graphite. A preliminary sensitivity analysis has been carried out, highlighting the importance of geometry and of distribution coefficients (Kd) in materials used to seal the disposal underground facility. Results show the possibility to correlate the Kd values, the volume and the location of the sealing materials to the amount of 14C migrating toward the surface.

Understanding the environmental and biogeochemical behaviour of radionuclides is essential for managing our nuclear legacy safely. Remediation efforts and the concept of geological disposal of nuclear waste focus on immobilizing radionuclides within the subsurface. Here we review recent developments in the understanding of solid-phase capture processes of Cs, Sr, Tc, U, Pu and Np. Abiotic interactions between minerals and these radionuclides (including sorption, reductive precipitation and co-precipitation) have been studied in various conditions. Microbially driven processes are much less well characterized, for example the effects of microbial reduction on the structure and reactivity of existing minerals, or their role in the formation of new minerals. Metabolites released by bacteria can play a role in both mineral dissolution and formation, and better understanding their release and role in mineralization has great potential in the development of solid-phase capture processes for radionuclides.

With the aid of a map of the research landscape covered by this review (created using a cluster-analysis tool, a self-organizing map), we highlight the most promising sequestration processes for specific radionuclides. However, radionuclides exhibit highly species-specific behaviour in their interactions with minerals and microorganisms. More research is required to characterize the role mineral surfaces play in bioreductive immobilization of Pu and Np, the reduction products formed, and their relative stability. Further studies should concentrate on more environmentally relevant experiments that include bacteria, minerals and radionuclides.

Most countries intend to dispose of their high-level radioactive wastes by converting them into a solidified wasteform which is to be buried within the earth. SYNROC is a titanate ceramic wasteform which has been designed for this purpose on the basis of geochemical principles. It comprises essentially rutile TiO2, ‘hollandite’ Ba(Al,Ti)Ti6O16, zirconolite CaZrTi2O7, and perovskite CaTiO3. The latter three phases have the capacity to accept the great majority of radioactive elements occurring in high-level wastes into their crystal lattice sites. These minerals (or their close relatives) also occur in nature, where they have demonstrated their capacity to survive for many millions of years in a wide range of geological environments. The properties of SYNROC and the crystal chemistry of its constituent minerals are reviewed in some detail and current formulations of SYNROC are summarized. A notable property of SYNROC it its extremely high resistance to leaching by groundwaters, particularly above 100°C. In addition, it can be shown that the capacity of SYNROC minerals to immobilize high-level waste elements is not markedly impaired by high levels of radiation damage. Current investigations are focused on developing a satisfactory production technology for SYNROC and progress towards this objective is described. The high leach resistance of SYNROC at elevated temperatures increases the range of geological environments in which the waste may be finally interred; in particular, SYNROC is well adapted for disposal in deep drill-holes, both in continental and marine environments. The fact that SYNROC is comprised of minerals which have demonstrated long-term geological stability is significant in establishing public confidence in the ability of the nuclear industry to immobilize high-level wastes for the very long periods required.

The significance of the potential impacts of microbial activity on the transport properties of host rocks for geological repositories is an area of active research. Most recent work has focused on granitic environments. This paper describes pilot studies investigating changes in transport properties that are produced by microbial activity in sedimentary rock environments in northern Japan. For the first time, these short experiments (39 days maximum) have shown that the denitrifying bacteria, Pseudomonas denitrificans, can survive and thrive when injected into flow-through column experiments containing fractured diatomaceous mudstone and synthetic groundwater under pressurized conditions. Although there were few significant changes in the fluid chemistry, changes in the permeability of the biotic column, which can be explained by the observed biofilm formation, were quantitatively monitored. These same methodologies could also be adapted to obtain information from cores originating from a variety of geological environments including oil reservoirs, aquifers and toxic waste disposal sites to provide an understanding of the impact of microbial activity on the transport of a range of solutes, such as groundwater contaminants and gases (e.g. injected carbon dioxide).

One of the most important isotopes to be considered for disposal in the proposed UK Geological Disposal Facility (GDF) for higher-activity radioactive wastes will be 99Tc, due to its long half-life, high fission yield and ability to migrate through the geosphere as the pertechnetate ion. Much of the Tc is likely to be in the lower Tc(IV) oxidation state due to the low Eh in the near field. Batch Tc(IV) sorption experiments have been performed from pH 3–13, using 95mTc as a spike, in the presence of quartz, hematite, goethite, plagioclase feldspar, sand and shale. Solutions containing Tc(IV) were prepared at trace concentrations to avoid precipitation of TcO2. Values for the partition coefficient (Rd) were found to range from ∽4 up to ∽2 × 104 cm3 g–1. Rd was heavily dependent on pH in all cases, with the highest values being found in the circumneutral area. These data will inform the performance assessment for the near-field behaviour of technetium in the UK's planned higher-activity wastes GDF. Surface complexation modelling has been performed, fitting the data using both monodentate and bidentate binding models.

Understanding flow along fractures and faults is of importance to the performance assessment (PA) of a geological disposal facility (GDF) for radioactive waste. Flow can occur along pre-existing fractures in the host-rock or along fractures created during the construction of the GDF within the excavation damage zone (EDZ). The complex fracture network will have a range of orientations and will exist within a complex stress regime. Critical stress theory suggests that fractures close to localized shear failure are critically stressed and therefore most conductive to fluid flow. Analysis of fault geometry and stress conditions at Sellafield has revealed that no features were found to be, or even close to being, classified as critically stressed, despite some being conductive. In order to understand the underlying reasons why non-critically stressed fractures were conductive a series of laboratory experiments were performed. A bespoke angled shear rig (ASR) was built in order to study the relationship between fluid flow (water and gas) through a fracture surface as a function of normal load. Fluid flow reduced with an increase in normal load, as expected. During unloading considerable hysteresis was seen in flow and shear stress. Fracture flow was only partially recovered for water injection, whereas gas flow increased remarkably during unloading. The ratio of shear stress to normal stress seems to control the fluid flow properties during the unloading stage of the experiment demonstrating its significance in fracture flow. The exhumation of the Sellafield area during the Palaeogene–Neogene resulted in considerable stress relaxation and in fractures becoming non-critically stressed. The hysteresis in shear stress during uplift has resulted in faults remaining, or becoming, conductive. The field and laboratory observations illustrate that understanding the stress-history of a fractured rock mass is essential, and a mere understanding of the current stress regime is insufficient to estimate the flow characteristics of present-day fractures.

Disposal in deep geological formations aims to provide isolation of long-lived radioactive waste for hundreds of thousands of years. This raises the question of the long-term governance of the repository throughout its lifetime. In the operational phase the repository is under active regulatory control. Once closed, there will be a phase of passive management control or indirect oversight. This will be followed, at some time in the future, by a period in which there is no oversight. This may be a result of a decision to cease management control or it may occur through loss of records or a change in priorities. The importance of the main exposure scenarios (exposure as a result of the gradual transport of radionuclides in groundwater, transport of radionuclides in gas, and exposure arising from inadvertent intrusion into the repository) are discussed with reference to these different phases. An interesting question is 'How do we minimize the risk of inadvertent intrusion in the far future?' Perhaps it is better to ensure that the repository is forgotten and should we try to plan for this? The different approaches are discussed and the importance of deciding on a strategy at an early stage is emphasized.

To understand the fate and impact of gas produced within a repository for radioactive waste, a series of laboratory and field scale experiments have been performed on the Callovo-Oxfordian claystone (COx), the proposed host rock for the French repository. Results show the movement of gas is through a localized network of pathways, whose properties vary temporarily and spatially within the claystone. Significant evidence exists from detailed laboratory studies for the movement of gas along highly unstable pathways, whose aperture and geometry vary as a function of local stress, gas and porewater pressures. The coupling of these parameters results in the development of significant time-dependent effects, impacting on all aspects of COx behaviour, from gas breakthrough time, to the control of deformation processes. Variations in gas entry, breakthrough and steady-state pressures are indicative of microstructural heterogeneity which exerts an important control on the movement of gas. The localization of gas flow is also evident in preliminary results from the large scale gas injection test (PGZ) where gas flow is initially focussed within the excavation damaged zone (EDZ), which acts as a preferential pathway for gas. Numerical models based on conventional two-phase flow theory are unable to adequately describe the detailed observations from laboratory tests.

The Nuclear Decommissioning Authority (NDA) is developing a safety case for the long-term management of higher activity wastes. This includes safety assessments of transport to and operations at the repository. One of the main faults and hazards to be considered is waste package response to impact accidents.

The criteria of impact performance for waste packages are based upon activity release of particulates generated from the break up of the waste form during impact. The NDA approach to impact performance is based upon waste package response from finite element modelling in combination with break-up tests.

Previous break up research commissioned by the NDA has concentrated on commercial graphite and glass samples. These extended studies, undertaken by the National Nuclear Laboratory in collaboration with the Department of Aerosol Technology of the Fraunhofer Institute of Toxicology and Experimental Medicine, provide break-up data specific to nuclear facilities and waste materials. These include archived unirrradiated graphite used to construct Magnox reactor cores and reflectors, simulant high level waste glass, selected grout formulations and selected metal-in-grout formulations.

Diffusion into the rock matrix is potentially an important retardation mechanism for nuclides leached from an underground radioactive waste repository in a fractured hard rock. A technique for measuring the intrinsic diffusion coefficient and rock capacity factor is briefly described. Simple solutions to migration model equations are used, together with diffusion results and typical hydrogeological parameters, to estimate the impact of matrix diffusion on radionuclide migration. It is shown that retardation factors in excess of 100 and reductions in the peak concentration by 3–4 orders of magnitude are possible for non-sorbed ions, which would otherwise be carried by the flow and not retarded at all.

The sorption behaviour of I−, Cs+, Ni2+, Eu3+, Th4+ and UO2+ 2on NRVB (Nirex reference vault backfill) a possible vault backfill, at pH 12.8 was studied. Sorption isotherms generated were compared to results obtained in the presence of cellulose degradation products (CDP). Whereas Cs was not affected by the presence of the organic compounds, a notable reduction in the sorption of Th and Eu to cement was observed. The results also indicated limited removal of Ni from solution (with or without an organic ligand) by sorption, the concentration in solution seemingly being determined solely by solubility processes. In the case of uranium, the presence of CDP increased the sorption to cement by almost one order of magnitude. Further studies into the uptake of CDP by cement are being undertaken to identify the mechanism(s) responsible.

Radioactive particulate matter (identified as hot particles) is present in the effluent discharged by the British Nuclear Fuels Ltd (BNFL) uranium reprocessing plant at Sellafield, Cumbria, UK. There is very little information on the abundance or chemical and physical forms of solid matter in the effluent; even less is known of the significance of particulate debris in relation to the uptake of radionuclides for non-occupationally exposed people as a result of transfer along marine foodchains. Some observations on the occurrence and abundance of hot particles in the vicinity of Sellafield are reported, with special reference to those that contain transuranic radionuclides (Pu,Am,Cm). Some of the uncertainties are discussed in an evaluation of the significance of hot particles, albeit aggregates of colloids for the smallest particles, and exposure to man from ionizing radiation. There is no evidence that hot particles derived from BNFL and subsequently dispersed into the marine environment represent a hazard to man. However, further studies are required in order to determine whether or not the pathways followed by the particles are significant, or different to those of other radionuclides through which the radiation exposure of man within this region of Cumbria is assessed.

The 2008 UK government White Paper, published as part of the Managing Radioactive Waste Safety programme, identified benefits to disposing of all of the UK's higher activity wastes at the same site. That is, a single geological disposal facility (GDF) could be constructed that consists of a module for low- and intermediate-level waste, and a module for high-level waste and spent fuel.

A safety case for a co-located GDF will have to consider the extent to which evolving thermo-hydro-mechanical-chemical and gas (THMCG) conditions in and around one module may affect conditions in the other module, including the extent to which barrier performance and radionuclide migration behaviour could be altered. Several research projects have been undertaken on behalf of Radioactive Waste Management Directorate aimed at understanding and evaluating the THMCG interactions that might occur during the disposal facility operational and post-closure phases.

This paper describes research on THMCG interactions between disposal modules based on illustrative GDF designs for different host rock environments. Interactions were evaluated using simple analytical solutions and detailed three-dimensional models. The analyses demonstrated that interactions can be controlled by design constraints.

Several European waste management organizations have started work on creating a technology platform to accelerate the implementation of deep geological disposal of radioactive waste in Europe. There is an increasing consensus in the international community about geological disposal as the preferred option for solving the long-term management of spent fuel, high-level waste and other longlived radioactive wastes. At the same time, European citizens have a widespread desire for a permanent solution for high-level radioactive waste disposal. A majority of European countries with nuclear power have active waste-management programmes, but the current status and the main challenges of those programmes vary. The most advanced waste management programmes in Europe (i.e. those in Sweden, Finland and France) are prepared to start the licensing process of deep geological disposal facilities within the next decade. Despite the differences between the timing and the challenges of the different programmes, there is a joint awareness that cooperation on the scientific, technical and social challenges related to geological disposal is needed, and that it is beneficial for the timely and safe implementation of the first geological disposal facilities. Such a demonstration of a viable solution for the management of high-level radioactive waste will enhance stakeholder confidence in Europe. It is envisaged that a technology platform would enhance European cooperation in this area. The platform will provide a tool for reducing overlapping work, to produce savings in total costs of research and implementation, and to make better use of existing competence and research infrastructures. From 2008, SKB (Sweden) and Posiva (Finland) led the preparation work to set up the implementing geological disposal of radioactive waste technology platform (IGD-TP). Since then other implementers from France, Germany, Switzerland, United Kingdom, Spain and Belgium have joined the project. To date a strategic research agenda for the platform has been prepared and consulted upon, which is now the basis for taking the platform into a deployment phase.

At the post-closure stage of a geological disposal facility for higher activity radioactive waste several species of gas are likely to be generated in the near-field environment. These could alter the sealing and chemical properties of the bentonite buffer and the local geochemical environment significantly. The authors' attempt to simulate multicomponent gas flow through variably saturated porous media is presented. Governing equations have been developed for a reactive gas-flow model to simulate the thermo-hydro-gas-chemical-mechanical behaviour, with specific reference to the performance of highly compacted bentonite buffer subjected to repository gas generation and migration. The developed equations have been included in the bespoke numerical model COMPASS and some generic simulations are also presented. The model presented extends current capability to assess buffer performance.

The implementation of a geological disposal facility requires the demonstration of confidence that such a facility would be safe during both the operational period and in the long-term after the closure of such a facility. The generic environmental safety case described in this paper is the vehicle used to demonstrate an understanding of environmental safety. It will be used to prepare a site-specific environmental safety case in due course. The approach taken will be consistent with a staged development and approval process, as advocated by the environmental regulators.

Cementitious matrices are being assessed for immobilization of radioactive wastes. This paper discusses some mineralogical aspects of cement chemistry and the uses of siliceous minerals as selective sorbants to enhance immobilization potential.

Studies of sorption and leaching of caesium from pulverized fuel ash (PFA), blast furnace slag, tobermorite, xonotlite, and clinoptilolite are reported. The role of incorporation of these additives in cement and the effect on the nature of the composite matrix on caesium behaviour has been investigated. Specific mechanisms of the interaction of additives with highly alkaline cement environment are described. While slags, PFA, and clinoptilolite undergo reaction at different rates, tobermorite and xonotlite appear to be stable in cement.