Geological disposal is the preferred option for the final storage of high-level nuclear waste and spent nuclear fuel in most countries. The selected host rock may be different in individual national programs for radioactive-waste management and the engineered barrier systems that protect and isolate the waste may also differ, but almost all programs are considering an engineered barrier. Clay is used as a buffer that surrounds and protects the individual waste packages and/or as tunnel seal that seals off the disposal galleries from the shafts leading to the surface.

Bentonite and bentonite/sand mixtures are selected primarily because of their low hydraulic permeability in a saturated state. This ensures that diffusion will be the dominant transport mechanism in the barrier. Another key advantage is the swelling pressure, which ensures a self-sealing ability and closes gaps in the installed barrier and the excavation-damaged zone around the emplacement tunnels. Bentonite is a natural geological material that has been stable over timescales of millions of years and this is important as the barriers need to retain their properties for up to 106 y.

In order to be able to license a final repository for high-level radioactive waste, a solid understanding of how the barriers evolve with time is needed. This understanding is based on scientific knowledge about the processes and boundary conditions acting on the barriers in the repository. These are often divided into thermal, hydraulic, mechanical, and (bio)chemical processes. Examples of areas that need to be evaluated are the evolution of temperature in the repository during the early stage due to the decay heat in the waste, re-saturation of the bentonite blocks installed, build-up of swelling pressure on the containers and the surrounding rock, and degradation of the montmorillonite component in the bentonite. Another important area of development is the engineering aspects: how can the barriers be manufactured, subjected to quality control, and installed?

Geological disposal programs for radioactive waste have generated a large body of information on the safety-relevant properties of clays used as engineered barriers. The major relevant findings of the past 35 y are reviewed here.

The Tamusu region in northwest China is a key candidate area for China's clay rock deep geological repositories (DGRs) for high-level radioactive waste (HLRW) as it is composed of a continuous layer of thick lacustrine mudstone. To evaluate this mudstone as a host rock, two special test boreholes were drilled to investigate its spatial distribution and mineralogical and hydraulic characteristics. The southwest boundary and depositional centre of the lake basin were well delineated by boreholes TZK-1 and TZK-2. The continuous single-layer thickness of the target mudstone formation was up to 300 m at a depth of 500–800 m. Three main mineral types were determined, namely carbonates (mainly dolomite and ankerite), analcime and albite, and their abundance was used to distinguish three different facies. Other mineral phases, such as clay minerals (mainly illite and kaolinite), pyrite, hematite, quartz and calcite, were present as admixtures. The presence of carbonates may increase the mechanical strength and analcime may enhance the radionuclide adsorption properties of the mudstone. The self-sealing properties, which may be affected by the small amount of clay minerals, remain to be investigated. The hydraulic conductivity of the mudstone determined via in situ pulse tests ranged from 10–13 to 10–10 m s–1, suggesting that the Tamusu mudstone has ultra-low permeability. The transmissivity of the Tamusu mudstone fluctuated in regions with varying lithologies, but remained relatively constant for consistent lithologies. In summary, these preliminarily results confirm the possible suitability of the target formation as a host rock for DGRs of China's HLRW.

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.

Carbon-14 (radiocarbon, 14C) is a long-lived radionuclide (5730 yr) of interest regarding the safety for the management of intermediate level wastes (ILW). The present study gives an overview of the release of 14C from irradiated Zircaloy cladding in alkaline media. 14C is found either in the alloy part of Zircaloy cladding due to the neutron activation of 14N impurities by 14N(n,p)14C reaction, or in the oxide layer (ZrO2) formed at the metal surface by the neutron activation of 17O from UO2 or (U-Pu)O2 fuel and water from the primary circuit in the reactor by 17O(n,α)14C reaction. Various irradiated and unirradiated Zircaloys have been studied. The total 14C inventory has been determined both experimentally and by calculations. The results seem to be in good agreement. Leaching experiments were conducted in alkaline media for several time durations. 14C was mainly released as carboxylic acids. Further, corrosion measurements were performed by using both hydrogen measurements and electrochemical measurements. The corrosion rate (CR) ranges from a few nm/yr to 100 nm/yr depending on the surface conditions and the method used for measurement. From a safety assessment point of view, the instant release fraction (IRF) was determined on irradiated Zircaloy-2. The results showed that the 14C inventory in the oxide was significantly below the 20% commonly used in safety case assessments.

Radiocarbon (14C or carbon-14, half-life 5730 yr) is a key radionuclide in the assessment of the safety of a geological disposal facility (GDF) for radioactive waste. In particular, the radiological impact of gaseous carbon-14 bearing species has been recognized as a potential issue. Irradiated steels are one of the main sources of carbon-14 in the United Kingdom’s radioactive waste inventory. However, there is considerable uncertainty about the chemical form(s) in which the carbon-14 will be released. The objective of the work was to measure the rate and speciation of carbon-14 release from irradiated 316L(N) stainless steel on leaching under high-pH anoxic conditions, representative of a cement-based near field for low-heat generating wastes. Periodic measurements of carbon-14 releases to both the gas phase and to solution were made in duplicate experiments over a period of up to 417 days. An initial fast release of carbon-14 from the surface of the steel is observed during the first week of leaching, followed by a drop in the rate of release at longer times. Carbon-14 is released primarily to the solution phase with differing fractions released to the gas phase in the two experiments: about 1% of the total release in one and 6% in the other. The predominant dissolved carbon-14 releases are in inorganic form (as 14C-carbonate) but also include organic species. The predominant gas-phase species are hydrocarbons with a smaller fraction of 14CO (which may include some volatile oxygen-containing carbon-species). The experiments are continuing, with final sampling and termination planned after leaching for a total of two years.

The gas release and speciation of carbon species from irradiated and unirradiated Zircaloy-4 samples, representative for the fuel cladding as used in Belgian nuclear power plants, were studied in a saturated Ca(OH)2 solution in anaerobic conditions. This environment is relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. Potentiodynamic polarization curves, recorded prior to the potentiostatic tests, revealed that irradiation seems to induce changes on the Zircaloy-4 corrosion behavior, such as a shift of the corrosion potential. Potentiostatic corrosion tests on unirradiated Zircaloy-4 provided a corrosion rate of ~54 nm/yr over a 7 day-experiment, whilst a corrosion rate of only ~4 nm/yr was calculated for the irradiated sample. Gas chromatography revealed that during simple immersion tests, which lasted 195 days, hydrogen, methane, ethane, and CO2 were produced, with methane being the major compound. Assuming that all carbon released from the metal was transformed into gaseous compounds, this yields to a corrosion rate ranging from 57 to 84 nm/yr for the irradiated sample. However, caution has to be taken on these corrosion rate and more tests should be performed to confirm these results.

Hectorite was synthesized from a Eu(III)-bearing brucite precursor in a multistep procedure. In a separate experiment, Eu(III) ions were adsorbed onto hectorite in suspension. Colloids were extracted from both samples. The size distributions in the colloidal fractions were characterizedby application of the asymmetrical flow field-flow fractionation (AsFlFFF) method and the corresponding elemental compositions were obtained by ICP-MS. Extended X-ray absorption fine structure (EXAFS) spectroscopy was used to characterize the local chemical environment surrounding Eu in thebulk samples and in the colloidal fractions.

The EXAFS results show that Eu is associated with hectorite upon co-precipitation or adsorption. Results from AsFlFFF suggest that Eu is structurally associated with the colloidal fraction extracted from bulk Eu-bearing co-precipitated hectorite.The AsFlFFF data are different for the colloidal fraction containing Eu(III) adsorbed on hectorite; in this sample they are consistent with a surface retention mechanism. These small but significant differences enable surface sorbed Eu to be distinguished from co-precipitated Eu. Eu is verylikely located in a clay-like environment in the co-precipitation experiment, and it forms inner-sphere surface complexes in the adsorption experiment. The results obtained using the different experimental techniques agree, and show the benefits of using multiple methods of analysis.

Trivalent europium was used as non-radioactive chemical homologue for trivalent actinides. Similar retention mechanisms are expected for the trivalent actinides if they are co-precipitating with or adsorbing onto sheet silicates. The present study provides information which can be usefullyadded to the safety assessments required for deeply buried nuclear waste disposal sites.

In the UK, disposal of packaged intermediate-level radioactive waste may involve waste packages being placed in a geological disposal facility (GDF) and surrounded by a cementitious backfill. Cracking of the backfill could occur due to a number of mechanisms, and this could affect the post-closure performance of the GDF.

This work has assessed potential cracking in the backfill during the backfilling and early post-closure period of GDF vaults with an open crown space in a higher strength rock. From the comprehensive range of processes considered, three were identified as potentially causing cracking: (1) during backfilling, plastic settlement under solid horizontal surfaces could result in horizontal gaps beneath waste packages; (2) within days of backfilling, early-age thermal contraction of the backfill could result in primarily vertical cracks; (3) over a number of years, expansion of waste packages could result in large horizontal cracks.

A groundwater flow model incorporating a representation of the cracks was used to calculate flows through a backfilled GDF vault, and through the cracks themselves. Including cracks increased the flow rate significantly. A reactive transport model was used to estimate the evolution of the pore water chemistry as groundwater flows through the cracked backfill. Calcite and brucite were predicted to precipitate, with brucite subsequently dissolving. Calcite build-up could seal some cracks.

The Environment Agency is the environmental regulator for the nuclear industry in England and Wales. We aim to ensure that any future Geological Disposal Facility (GDF) will meet our standards for protecting people and the environment during development, operation and following closure. We shall only grant an environmental permit for such a facility if the GDF developer submits an Environmental Safety Case (ESC) to the Environment Agency demonstrating that the proposed GDF will meet our standards. The ESC should be based on sound science and supported by appropriate research and development. The Environment Agency undertakes its own research programme to inform its regulatory decisions.

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 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.

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.

Programmes for the geological disposal of radioactive wastes are by nature extremely complex. A structured approach for making and documenting varied kinds of decisions is required to support programme design and implementation. At each programme stage, the decision-making process must be able to identify and justify key priorities for work, to reduce uncertainties.

To support structured decision-making evidence support logic (ESL) has been developed and applied to varied complex projects, nationally and internationally, in several industries. Evidence support logic involves breaking down a hypothesis that informs a decision into a hierarchical 'decision tree'. Examples of hypotheses are 'the geology associated with site x will provide sufficient disposal capacity', 'container x will contain waste form y for z years' and 'the engineered barrier system will provide the required safety functions'. Independent evaluations of confidence 'for' and 'against' bottom-level hypotheses allow the level of remaining uncertainty (or conflict) to be recognized explicitly, and the overall confidence (and uncertainty) relevant to the overall decision, and key sensitivities, to be represented clearly and succinctly.

Thus ESL can help (1) break down decisions into a manageable and logical structure, assisting clear presentation; (2) identify key uncertainties and sensitivities to inform prioritization; and (3) test whether the outcomes of specific studies have improved confidence.

Batch experiments have been used to assess the sorption properties of a potential cementitious repository backfill, NRVB (Nirex reference vault backfill), using Eu3+ as a model trivalent radionuclide and ethylenediaminetetraacetic acid (EDTA) as a competing ligand. The NRVB is an effective scavenger of Eu from solution, with most sorbed within minutes onto the crushed material and less than 1.5% remaining in solution after one day (R d values in the range 0.3–2.4 × 104 l kg−1). Ultrafiltration showed that nearly all of this remaining Eu (>94%) is attached to NRVB derived colloids or particulates that are mainly retained by a 100 kDa ultrafilter. High concentrations of EDTA (>0.01 M) reduced the extent of sorption at apparent equilibrium. The addition of EDTA to a pre-equilibrated system of Eu3+ and NRVB resulted in a temporary suspension of some Eu, but this very quickly returned to the solid phase. There is some irreversibility in these systems, with EDTA able to prevent removal of Eu(III) from solution, but unable to bring it back into solution under the same conditions.

The role of the microbial activity on the transport properties of host rocks for geological repositories, particularly in the far-field, is an area of active research. This paper compares results from experiments investigating changes in transport properties caused by microbial activity in sedimentary rocks in Japan (mudstones) and sandstone (UK).

These experiments show that both Pseudomonas denitrificans and Pseudomonas aeruginosa appear to survive and thrive in pressurized flow-through column experiments which utilized host rock materials of relevance to radioactive waste disposal. Indeed, despite there being a difference in the numbers of organisms introduced into both biotic experiments, numbers appear to stabilize at ∼105 ml−1 at their completion. Post experimental imaging has highlighted the distinct differences in biofilm morphology, for the chosen rock types and bacteria, with Pseudomonas aeruginosa derived biofilms completely covering the surface of the sandstone host and Pseudomonas denitrificans forming biofilament structures. Regardless of substrate host or choice of microbe, microbial activity results in measurable changes in permeability. Such activity appears to influence changes in fluid flow and suggests that the transport of radionuclides through the far-field will be complicated by the presence of microbes.

Understanding the behaviour of gas in a geological disposal facility (GDF) is an essential component of analysing the facility evolution and long-term (post-closure) safety performance. This includes the impacts of gas on the physico-chemical evolution of the GDF, and the release and migration of radionuclides in water and gas.

The Nuclear Decommissioning Authority Radioactive Waste Management Directorate is participating in the EC FORGE (fate of repository gases) project (www.forgeproject.org) and conducting independent research. Key research themes are modelling the impacts of different host rocks on facility evolution including coupled processes, and upscaling the effects of small scale features that can significantly influence the evolution of the whole facility.

Recent code developments have enabled coupled processes to be represented more realistically in models. This has significantly advanced understanding of facility evolution, as discussed in this paper, and will improve future assessment models. There is potential to further improve approaches to upscaling the effects of small scale features on strongly coupled processes, within the context of the EC FORGE project.

In the Swedish KBS-3 repository concept, there is potential for gas to be generated from corrosion of ferrous materials under anoxic conditions, combined with the radioactive decay of the waste and radiolysis of water. A full understanding of the probable behaviour of this gas phase within the engineered barrier system (EBS) is therefore required for performance assessment. We demonstrate key features from gas transport experiments on pre-compacted Mx80 bentonite, under laboratory and field conditions, and discuss their implications in terms of a conceptual model for gas migration behaviour. On both scales, major gas entry is seen to occur close to the sum of the porewater and swelling pressures of the bentonite. In addition, gas pressure at breakthrough is profoundly sensitive to the number and location of available sinks for gas escape. Observations of breakthrough can be explained by the creation of dilatational pathways, resulting in localized changes in the monitored porewater pressures and total stresses. These pathways are highly unstable, evolving spatially and temporally, and must consequently influence the gas permeability as their distribution/geometry develops.

Such observations are poorly embodied by conventional concepts of two-phase flow, which do not fully represent the key processes involved. Although dilatancy based models provide a better description of these processes, the paucity of data limits further development and validation of these models at present.

Corrosion, water radiolysis and microbial degradation will result in the generation of gas within repositories designed for the geological disposal of high-level radioactive waste. It is therefore crucial in the design of such facilities that the relevant mechanisms allowing gas migration through repository materials, both engineered barriers and clay-based candidate host rocks, are correctly identified. In Belgium, the Boom Clay represents a candidate host material for which the advective gas breakthrough characteristics and transport properties have been extensively tested and are well defined by numerous studies. The Boom Clay displays a significant capacity for self-sealing and both laboratory and field tests indicate that advective gas transport occurs not by visco-capillary flow, but instead through the formation of pressure-induced dilatant pathways. In this study, we present results from a gas injection test designed to demonstrate the presence of these features by injecting nanoparticulate tracers with helium gas into a sample of Boom Clay. The results provide conclusive evidence for the formation of transient, dilatant gas pathways within a candidate clay-based host rock. This technique provides a novel diagnostic tool for the identification of processes governing multi-phase flow, supporting robust long-term assessments of repository performance.