Eudialyte-group minerals (EGM) attract global interest as potential resources for high-field-strength elements (e.g. Zr, Nb, Ta, and rare-earth elements), i.e. critical materials for modern technologies. They are particularly valued for their relative enrichment in the most critical lanthanides, i.e. Nd and heavy rare earth elements (Gd–Lu). However, rare earth element (REE) substitution mechanisms into the EGM structure are still poorly understood. Light and heavy REE may occupy different sites and there may be ordering and/or defect clustering in the structure. This study uses X-ray absorption spectroscopy to determine the structural state of REE in EGM from prospective eudialyte-bearing complexes. Yttrium K-edge and Nd L3-edge spectra were collected as proxies for heavy and light REE, respectively, and compared to natural and synthetic REE-bearing standards. Extended X-ray absorption fine structure data yield best fits for Y in six-fold coordination with Y–O distances of 2.24–2.32 Å, and a second coordination sphere comprising Fe, Na, Ca, Si and O at radial distances of 3.6–3.8 Å. These findings are consistent with dominant Y3+ substitution for Ca2+ on the octahedral M1 site in all the samples studied, and exclude preferential substitution of Y3+ onto the smaller octahedral Z site or the large low-symmetry N4 site.

Using lattice strain theory, we constructed relative partitioning models to predict site preferences of lanthanides we have not measured directly. The models predict that all REE are favoured on the Ca-dominant M1 site and that preferential partitioning of heavy over light REE increases in EGM containing significant Mn in the M1-octahedral rings (oneillite subgroup). Thus, the flat REE profiles that make EGM such attractive exploration targets are not due to preferential partitioning of light and heavy REE onto different sites. Instead, local ordering of Mn- and Ca-occupied M1 sites may influence the capacity of EGM to accommodate heavy REE.

In the UK, radioactive wastes currently planned for disposal in a geological disposal facility (GDF) are intermediate-level waste, some low-level waste and high-level waste. Disposal of other materials, including spent fuel, separated uranium and separated plutonium are also included in the planning of a GDF, if such materials are classified as wastes in the future. This paper gives an overview of the radionuclide behaviour research studies of the Nuclear Decommissioning Authority Radioactive Waste Management Directorate (NDA RWMD). The NDA RWMD's current understanding of the processes that control radionuclide behaviour in groundwater and how the engineered and natural barriers in a GDF would contain radionuclides is presented. Areas requiring further work are also identified.

In support of the disposal system safety case for a geological disposal facility (GDF) there is a requirement to consider 'what-if' hypothetical scenarios for post-closure nuclear criticality. Although all such scenarios are considered very unlikely, one 'what-if' scenario is the mobilization of fissile material from a number of waste packages and its slow accumulation within the GDF or the immediate surroundings. Should sufficient fissile material accumulate a quasi-steady-state (QSS) transient criticality event could result. A computer model has been developed to understand the evolution and consequences of such an event.

Since a postulated QSS criticality could persist for many millennia, building confidence in the modelling approach is difficult. However, the Oklo natural reactors in Africa operated for similar durations around two billion years ago, providing a natural analogue for comparison. This paper describes the modelling approach, its application to hypothetical criticality events for a GDF, and how the model can be compared to Oklo. The model results are found to be in agreement with the observational evidence from Oklo, building confidence in the use of the QSS model to simulate postulated post-closure criticality events in GDFs.

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.

In situ atomic force microscopy (AFM) experiments were used to study the effect of trace amounts of a commercial inhibitor on the (001) baryte surface during growth. The additive tested was a copolymer, used as a scale inhibitor in oil recovery (maleic acid/allyl sulfonic acid copolymer with phosphonate groups, partial sodium salt). The morphology of the growth was used to gain a better understanding of the inhibition mechanism. Without an inhibitor, barium sulfate growth occurred by 2D island nucleation and spreading. The addition of a small amount (0.1 ppm and 0.5 ppm) of copolymer inhibitor enhances 2D nucleation but blocks growth. Just 1 ppm of inhibitor blocks nucleation and growth by adsorption of a copolymer layer onto the baryte surface. Similarly in 3D studies, small amounts of inhibitor seem to act on growth and not on nucleation and larger amounts of copolymer act on both by adsorption of the copolymer to all baryte surfaces keeping the particles in their embryo stage.

A cementitious backfill has been proposed in many geological disposal concepts for intermediate-level waste and low-level waste in the UK and elsewhere. In this paper, the main features of the chemical evolution of backfill and the associated changes in the near-field pH are illustratedwith results from recent work. For example, interaction of the groundwater with calcium silicate hydrate (C-S-H) phases in a backfill is expected to play an important role in the long-term pH-buffering behaviour. Existing experimental data for the dissolution of C-S-H gels are compared withrecent experimental results from leach tests on gels of a lower calcium to silicon ratio (C/S) to provide a consistent set of data across the full C/S range. The results confirm that a congruent dissolution point around C/S = 0.8 is approached by leaching from below (i.e. for gels with 0.29< C/S < 0.8), as well as from above, as reported elsewhere. In addition, a spreadsheet model has been developed to calculate the volume of backfill required at the vault scale to meet specified pH performance criteria. This model includes the major reactions of the backfill with thegroundwater, waste encapsulants and waste components. It can also consider the effects of specific waste packages on local pH performance to allow comparison with the vault-scale calculations.

Carbon-14 has been identified as one of the more significant radionuclides in solid radioactive wastes in a repository, due to the potential radiological impact arising if 14C were to be released and enter the biosphere. However, the assessment of radiation doses is complicated by the major role of carbon in biological processes, and this has tended to lead to the adoption of a cautious assessment approach.

An international comparison of five models used to predict uptake of 14C to agricultural crops has been undertaken, within the BIOPROTA framework. Processes investigated include conversion of 14C-labelled CH4 into CO2 in soils, carbon accumulation in and release from soil carbon pools, gaseous emanation to, and dispersion from, the plant canopy atmosphere and, incorporation into plants by photosynthesis.

For a unit rate of entry of 14C to soil, modelled activity concentrations in cereal crops differ by three to five orders of magnitude. This reflects, in part, differing assumptions for mixing and dispersion of air above the soil surface and within the crop canopy layer. For a unit activity concentration of 14C in air, the modelled uptake to cereal crops converges significantly. Following an assumed irrigation of crops with groundwater containing unit activity of 14C, the predicted uptake to crops varied by two to four orders of magnitude, again largely dominated by assumptions regarding the canopy atmosphere. In all cases, there is some convergence in model predictions as field size increases.

A continuing programme of field research is being undertaken in parallel with the assessment work.

The effect of competition from other metal ions on the complexation of Ni with isosaccharinic acid, gluconic acid and picolinic acid, at high pH, is described. The competing metal ions used were divalent Co, trivalent Eu and tetravalent Th. In the majority of cases, competition from these metal ions followed the predicted pattern, with most anomalies seeming to be caused by sorption of Ni to the many different solid phases formed in presence of the competing ions. The Ni solid phase was shown to change with time during the course of the study. No major unexplained competition effects were found.

A range of potential concepts for the geological disposal of high level wastes and spent fuel are being studied and considered in the UK. These include concepts that use bentonite as a buffer material around the waste containers. The bentonite will be required to fulfil certain safety functions, the most important being (1) to protect the waste containers from detrimental thermal, hydraulic, mechanical and chemical processes; and (2) to retard the release of radionuclides from any waste container that fails. The bentonite should have a low permeability and a high sorption capacity.

These safety functions could be challenged by certain features, events and processes (FEPs) that may occur during the evolution of the disposal system. A consideration of how these FEPs may affect the safety functions can be used to identify and to prioritize the important areas for research on bentonite. We identify these important areas (which include hydration of compacted bentonite, illitization and erosion of bentonite), and describe how they are being investigated in current international research on bentonite.

The Äspö EBS Task Force is a collaborative international project designed to carry out research on bentonite. In 2011, the Nuclear Decommissioning Authority Radioactive Waste Management Directorate joined the EBS Task Force partly to benefit from its collective experience. The work of the EBS Task Force is split into two research subareas: (1) the THM subarea, which includes tasks to understand homogenization of bentonite as it resaturates, to investigate the hydraulic interaction between bentonite and fractured rock, and to model in situ experiments; and (2) the THC subarea, which includes tasks to investigate the issue of understanding transport through bentonite, and to model in situ experiments. In particular, the bentonite rock interaction experiment is a large-scale in situ experiment concerned with understanding groundwater exchange across bentonite rock interfaces, with the objective of establishing better understanding of bentonite wetting. In this paper, we describe our work to model the spatial and temporal resaturation of bentonite buffer in a fractured host rock.

The international long-term cement studies (LCS) project aims to increase the understanding of the behaviour of cement within a radioactive waste disposal system and how hyper-alkaline leachates may interact with host rock. Such an understanding enables confident, robust and safety-relevant statements to be made concerning future system behaviour, irrespective of host rock, engineered barrier system, or waste type. The LCS project involves laboratory experiments, in situ tests and numerical modelling to address these issues. The agencies participating are Nagra (Switzerland), JAEA (Japan), the Nuclear Decommissioning Authority, Radioactive Waste Mangement Directorate (UK), Posiva (Finland) and SKB (Sweden).

Project activities have included: the development of conceptual and theoretical models of cement–rock interaction; testing of numerical models against data from laboratory experiments and industrial and natural analogues of cement–rock reaction; and the synthesis and incorporation of performance assessment (PA) relevant data from analogue studies. Key threads running through these studies include an analysis of issues relating to upscaling of processes and data to the greater temporal and spatial scales relevant to PA, and investigations of modelling the changes in physical properties that accompany geochemical reaction. Here we present examples of the results from model test cases, highlighting the important issues that have arisen.

Vitrification is used for the immobilization and conditioning of high-level waste (HLW) arising from the reprocessing of spent nuclear fuel in the UK. Vitrification is also under consideration for the immobilization of certain intermediate-level wastes (ILW), where there may be advantages of volume reduction and removal of uncertainties in long-term waste behaviour, compared to encapsulation in a cement grout. This paper gives an overview of recent work into the chemical durability of UK vitrified wasteforms to inform the technical specification for the disposal facilities for these waste products and the treatment of their long-term behaviour in post-closure performance assessment. This has included: (1) measurements of the initial glass dissolution rates of a simulated HLW Magnox waste glass in a range of groundwater types representative of potential UK host geologies and in simulated high pH near-field porewaters relevant to potential disposal concepts, using Product Consistency Test type-B (PCT-B) at 40°C; and (2) durability testing of three simulant ILW glasses in a saturated calcium hydroxide buffered solution to simulate conditions in cement-based disposal vaults, using PCT-B tests at 50°C.

The experimentally defined initial rate of HLW Magnox waste glass dissolution in a range of simulated groundwater compositions appears to be similar regardless of the ionic strength and major element composition of the solution. The release of caesium from HLW Magnox waste glass appears to be sensitive to solution composition. Caesium is selectively retained in the glass compared to other soluble components in the two low ionic strength solutions, but is released at similar rates to other soluble components in the three groundwaters and Ca(OH)2 solution. Whether this change in caesium retention is an ionic strength effect or is related to changes in the nature of the surface alteration layer formed on the glass, has yet to be established. For HLW Magnox waste glass, dissolution is accelerated at high pH in NaOH solution, however, the presence of calcium acts to mitigate the effects of high pH, at least initially. In Ca(OH)2 solution, calcium is found to react with all the glasses studied leading to the formation of calcium-containing alteration products. The initial dissolution behaviour in Ca(OH)2 solution varies with glass composition and in particular appears to be sensitive to the boron content.

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.

Mining is known to be a major source of metal contamination for fluvial systems worldwide. Monitoring and understanding the effects on downstream water and sediment quality is essential for its management and to mitigate against detrimental environmental impacts. This study aimed to examine the effects of the small-scale, abandoned, hydraulic Black Creek gold mine on the geochemical content of fine (<63 μm) stream bed and floodplain sediment in the Horsefly watershed (British Columbia, Canada), which is a part of the Quesnel River basin. Concentrations of As, Cd, Se and Zn were determined by aqua regia digestion and the modified BCR (European Community Bureau of Reference) sequential extraction procedure followed by inductively coupled plasma-optical emission spectroscopy or inductively coupled plasma-mass spectrometry. Age-depth models for the floodplain cores were constructed using excess Pb-210 and Cs-137 activity concentration profiles. The results show that the mine caused local As contamination of the Horsefly River floodplain sediment during the first years of operation, but at present the contamination is mainly contained in the Black Creek. Present-day and recent concentrations of Cd, Se and Zn in the Horsefly River are elevated and this is probably derived from other upstream mining activities. The spatial and temporal changes in heavymetal concentrations suggest a slight, but not particularly widespread, mining effect on the finesediment geochemistry in the Horsefly River system.

With potential storage of several hundred years underground in a geological disposal facility (GDF) before closure, there is a requirement for radioactive waste packages to perform adequately. Over the past 25 years, cementitous grouts based on blast furnace slag (BFS) and ordinary Portland cement (OPC), have been used in the UK to immobilize intermediate-level waste (ILW), Magnox swarf, and currently such wasteforms are in surface storage awaiting geological disposal. Magnox fuel cladding will slowly corrode when encapsulated in alkaline cementitious grouts to produce hydrogen gas and an expansive corrosion product. Expansive corrosion products may lead to degradation of the wasteform and, if extensive, could affect the container. This study investigated the acute and chronic rates of corrosion of unirradiated Magnox swarf encapsulated in a BFS/OPC grout over a range of temperatures (25, 40, 60, 75 and 90°C) for curing times of up to 2½ years. Structural product degradation starts to develop as the tensile strength of the grout is approached by the expansive forces generated. Deformation of some experimental containers was also noted. An estimate of the time taken for a grouted product to fracture from this study due to the corrosion solely of Magnox is 350 years for storage at 25°C. Although not fully described in this paper, the main cement phases were calcium silicate hydrate, portlandite and gehlenite; the main product of Magnox corrosion was brucite. The curing temperature did not affect the compositions of the Magnox metal, corrosion product or grout.

This paper provides a summary of the research programme undertaken in support of the Low Level Waste Repository's 2011 environmental safety case (ESC). The programme has been developed, based on an understanding of safety issues and the requirements of the ESC. The research requirements to underpin the safety case have been identified by means of an auditable process, and subjected to scrutiny by both the regulators and a peer review group. Key research priorities for the future are identified.

A three-year European Commission project entitled performance assessment methodologies in application to guide the development of the safety case (PAMINA) was conducted in the period 2006–2009 and brought together 27 organizations from ten European countries, including the Nuclear Decommissioning Authority and Galson Sciences Ltd from the UK. The PAMINA project had the aim of improving and developing a common understanding of performance assessment (PA) methodologies for disposal concepts for spent fuel and other long-lived radioactive wastes in a range of geological environments.

Work undertaken within PAMINA focussed on four areas: (1) review of PA methodologies in participating organizations; (2) treatment of uncertainty in PA and the safety case; (3) other methodological advancements in PA; and (4) relevance of advanced PA approaches to practical cases.

The state of development of a radioactive waste disposal programme has a strong influence on the type of safety case and supporting PA that is produced. A range of PA methodologies has been developed by different waste management organizations. This paper presents a selection of conclusions from the PAMINA project, in the context of general understanding developed on what would constitute an acceptable safety case for a geological disposal facility, and outlines areas for further development.

Some of the illustrative concepts for the disposal of intermediate-level waste in a geological disposal facility in the UK employ a cementitious backfill around the waste packages. The concept for higher strength rocks would use a highly alkaline backfill composed of Portland cement (now known as CEM I), hydrated lime and limestone flour, referred to as Nirex reference vault backfill (NRVB).

This paper reports a study of the extensive leaching of cured NRVB in a range of generic leachant compositions (deionized water, 0.1 M and 1 M NaCl solutions) under flow-through conditions using a flexible wall permeameter. The experiments were designed to run for up to two years and to pass at least 1000 volumes of leachant (defined as the cumulative leachate volume produced/NRVB solid specimen volume) through the NRVB samples. Results for the pH evolution profiles of the leachates and the microstructural analysis of the unleached and leached samples are presented.

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.