A detailed understanding of the response of mineral phases to the radiation fields experienced in a geological disposal facility (GDF) is currently poorly constrained. Prolongued ion irradiation has the potential to affect both the physical integrity and oxidation state of materials and therefore may alter a structure's ability to react with radionuclides. Radiohalos (spheres of radiation damage in minerals surrounding radioactive (α-emitting) inclusions) provide useful analogues for studying long term α-particle damage accumulation. In this study, silicate minerals adjacent to Th- and U-rich monazite and zircon were probed for redox changes and long/short range disorder using microfocus X-ray absorption spectroscopy (XAS) and high resolution X-ray diffraction (XRD) at Beamline I18, Diamond Light Source. Fe3+ → Fe2+ reduction has been demonstrated in an amphibole sample containing structural OH– groups – a trend not observed in anhydrous phases such as garnet. Coincident with the findings of Pattrick et al. (2013), the radiolytic breakdown of OH– groups is postulated to liberate Fe3+ reducing electrons. Across all samples, high point defect densities and minor lattice aberrations are apparent adjacent to the radioactive inclusion, demonstrated by micro-XRD.

Radioactive Waste Management Limited (RWM) is tasked with implementing geological disposal of the United Kingdom's (UK) higher activity radioactive wastes. This paper describes how RWM's biosphere modelling capability has been extended from a solely terrestrial model to allow potential contaminant releases to estuarine, coastal and marine systems around the UK to be represented. The new models aim to strike a balance between being as simple as can be justified, erring on the side of conservative estimates of potential doses, while also representing the features and processes required to reflect and distinguish UK coastal systems. Sediment dynamics (including meandering of estuaries and sediment accumulation) are explicitly represented in a simplified form that captures the accumulation and remobilization of radionuclides. Long-term transitions between biosphere systems (such as from a salt marsh to a terrestrial system) are outside the scope of the study. The models and supporting data draw on information about the UK that is representative of present-day conditions and represent potential exposures arising from both occupational and recreational habits.?

Generic calculations demonstrate that potential doses to humans arising from releases to estuarine, coastal and marine systems are typically more than two orders of magnitude lower than those for equivalent releases to terrestrial systems via well water and groundwater discharge to soil. The extended capability (i) ensures that RWM is able to undertake assessments for potential coastal site contexts, if and when required, and (ii) provides RWM with quantitative evidence to support the principal focus on terrestrial releases ( particularly for more generic assessments).

The design of a novel MEMS (Micro-Electro-Mechanical System) sensor-based monitoring system is presented in this article for the in situ monitoring of the conditions (temperature, relative humidity) of an engineered bentonite barrier for the underground disposal of radioactive waste in a geological disposal facility (GDF). A first prototype of the monitoring system incorporating multiple state-of-the-art MEMS sensors has been developed on a PCB-based (Printed Circuit Board) structure, in order to measure the variation of temperature and relative humidity inside a cylindrical bentonite block during the hydration process. The monitoring system comprises separate sensor boards, the microcontroller-equipped interface board and the software user interface in the Labview environment. One of the main design priorities is to reduce the size of the embedded sensors in order to minimize their influence on the hydro-mechanical response of the bentonite block. The sensor boards are encapsulated in different manners to protect them from moisture, chemical corrosion and mechanical shocks. The sensor system has been tested and calibrated in the temperature range between –20°C and 120°C, and at different relative humidity levels implemented by saturated salt solutions in enclosed containers. Test results demonstrate that the sensors have shown good functionality and robustness in harsh test environments such as high temperature and high humidity. Both temperature and relative humidity sensors have shown satisfactory precision level and temporal stability, which are in good accordance with the design specification of these devices.

In the deep geological disposal of nuclear waste in crystalline rock, erosion of the bentonite buffer may occur during periods of glaciation. Previous researchers have examined the mechanism and rates of extrusion and erosion for purified montmorillonite samples in smooth planar fractures. In this paper, we investigate the influence of using MX-80 material (as delivered, i.e. including accessory minerals) and a naturally varying aperture on bentonite erosion. A bespoke fracture flow cell was constructed for this purpose and flow through conducted with deionized water. Throughout the experiment, gravimetric analysis was undertaken on the effluent and the swelling pressure of the bentonite monitored. Quantitative image analysis of the extrusion process was also undertaken. When the swelling pressure data were analysed, alongside both the oscillations in erosion rate and the area of the accessory-mineral ring, a two-stage mechanism governing the erosion process became apparent. Once an accessory-mineral ring had formed at the edge of the extruded material, further increases in swelling pressure resulted in a breach in the accessory-mineral ring, triggering an erosive period during which, the mineral ring was supplemented with additional minerals. The cycle repeated until the ring was sufficiently strong that it remained intact. This observed process results in erosion rates one order of magnitude less than those currently used in long-term safetycase calculations.

Carbon-14 (C-14) is a key radionuclide in the assessment of a geological disposal facility (GDF) for radioactive waste. In the UK a significant proportion of the national C-14 inventory is associated with reactor-core graphite generated by the decommissioning of the UK's Magnox and AGR reactors.

There are a number of uncertainties associated with the fate and transport of C-14 in a post-closure disposal environment that need to be considered when calculating the radiological impacts of C-14-containing wastes. Some of these uncertainties are associated with the distribution of C-14-containing gaseous species such as 14CH4 and 14CO2 between the groundwater and gaseous release pathways. As part of the C14-BIG programme, a modelling framework has been developed to investigate these uncertainties. This framework consists of a biogeochemical near-field evolution model, incorporating a graphite carbon-14 release model, which interfaces with a geosphere/biosphere model. The model highlights the potential impact of the microbial reduction of 14CO2 to 14CH4, through the oxidation of H2, on C-14 transport. The modelling results could be used to inform the possible segregation of reactor graphite from other gasgenerating wastes.

In support of the Radioactive Waste Management (RWM) safety case for a geological disposal facility (GDF) in the UK, there is a regulatory requirement to consider the likelihood and consequences of nuclear criticality. Waste packages are designed to ensure that criticality is not possible during the transport and operational phases of a GDF and for a significant period post-closure. However, over longer post-closure timescales, conditions in the GDF will evolve.

For waste packages containing spent fuel, it can be shown that, under certain conditions, package flooding could result in a type of criticality event referred to as 'quasi-steady-state' (QSS). Although unlikely, this defines a 'what-if' scenario for understanding the potential consequences of post-closure criticality. This paper provides an overview of a methodology to understand QSS criticality and its application to a spent fuel waste package.

The power of such a hypothetical criticality event is typically estimated to be a few kilowatts: comparable with international studies of similar systems and the decay heat for which waste packages are designed. This work has built confidence in the methodology and supports RWM's demonstration that post-closure criticality is not a significant concern.

ENTRAP comprises a pan-European cooperation of leading scientific institutions and regulatory bodies in the field of nuclear-waste characterization and its quality assurance for the safe disposal of radioactive waste. Here, the scope of this cooperation is presented and explained and links or interfaces for a potential collaboration with partners fulfilling tasks of IDG-TP are pursued.

The existing capability and current development needs for implementing electroforming as a viable manufacturing process to produce copper cylinders with dimensions comparable to the Swedish KBS-3 design are discussed. Large freestanding copper cylinders can be produced readily, but there is a need to address challenges associated with controlling the electro-deposition process to conform to compositional and mechanical requirements of the copper layers produced. The methodology to optimize the manufacturing process, based on a study of key parameters, such as the effects of electrolyte additives on grain size and the chemical composition of electroformed layers, is described here. Possible ways to introduce a robust manufacturing route are also presented.

The first comprehensive assessment of the dissolution kinetics of simulant Magnox–ThORP blended UK high-level waste glass, obtained by performing a range of single-pass flow-through experiments, is reported here. Inherent forward rates of glass dissolution were determined over a temperature range of 23 to 70°C and an alkaline pH range of 8.0 to 12.0. Linear regression techniques were applied to the TST kinetic rate law to obtain fundamental parameters necessary to model the dissolution kinetics of UK high-level waste glass (the activation energy (Ea), pH power law coefficient (η) and the intrinsic rate constant (k0)), which is of importance to the post-closure safety case for the geological disposal of vitreous products. The activation energies based on B release ranged from 55 ± 3 to 83 ± 9 kJ mol–1, indicating that Magnox–THORP blend glass dissolution has a surface-controlled mechanism, similar to that of other high-level waste simulant glass compositions such as the French SON68 and LAW in the US. Forward dissolution rates, based on Si, B and Na release, suggested that the dissolution mechanism under dilute conditions, and pH and temperature ranges of this study, was not sensitive to composition as defined by HLW-incorporation rate.

Due to its favourable properties, in particular, low permeability and swelling capacity, bentonite has been favoured as an engineered-barrier and backfill material for the geological storage of radioactive waste. To ensure its safe long-term performance it is important to understand any changes in these properties when the material is subject to heat-emitting waste. As such, this study investigates the hydraulic response of bentonite under multi-step thermal loading subject to a constant-volume boundary condition, to represent a barrier system used in a crystalline or other hard-rock host rock. The experimental set up allows continuous measurement of the hydraulic and mechanical responses during each phase of the thermal cycle. After the initial hydration of the bentonite, the temperature was raised in 20°C increments from 20 to 80°C followed by a final step to reach 120°C. Each temperature was held constant for at least 7–10 days to allow the hydraulic transients to equilibrate. The data suggest that the permeability of bentonite appears to be sensitive to changes in temperature which may extend beyond those explained by simple changes in water viscosity. However, permeability may be boundary-condition dependent and this should be considered when designing experiments or applying these results to other repository host rocks. Either way, the magnitude of the change in permeability observed in this study is minor and its impact on the hydraulic performance of the barrier is negligible.

A geological disposal facility (GDF) will include fissile materials that could, under certain conditions, lead to criticality. Demonstration of criticality safety therefore forms an important part of a GDF's safety case.

Containment provided by the waste package will contribute to criticality safety during package transport and the GDF operational phase. The GDF multiple-barrier system will ensure that criticality is prevented for some time after facility closure. However, on longer post-closure timescales, conditions in the GDF will evolve and it is necessary to demonstrate: an understanding of the conditions under which criticality could occur; the likelihood of such conditions occurring; and the consequences of criticality should it occur.

Work has addressed disposal of all of the UK's higher-activity wastes in three illustrative geologies. This paper, however, focuses on presenting results to support safe disposal of spent fuel, plutonium and highlyenriched uranium in higher-strength rock.

The results support a safety case assertion that post-closure criticality is of low likelihood and, if it was to occur, the consequences would be tolerable.

The European InSOTEC project aims to identify remaining socio-technical challenges for the implementation of geological disposal of radioactive waste in European countries. Geological disposal has been recognized as the reference solution in radioactive waste management by the European Commission, which has set up the IGD-TP as a platform to foster exchanges between authorities on research and development (R&D) on this topic. This paper turns to the national level to analyse national R&D programmes on radioactive waste management (RWM) and particularly, geological disposal (GD) in order to assess to what extent there are national platforms that broaden interaction beyond an expert community. For this, we focus on identifying those practices and structures that reflect a degree of inter- and transdisciplinarity in the design and implementation of national R&D programmes. Our argument, embedded in the Science and Technology Studies (STS) literature, is that the way GD is framed by the research and policy community has important implications for the processes of developing and governing a technology. From our empirical case studies, we will explore the extent to which social-science research is integrated and the way stakeholders may have an impact on the development of R&D programmes. Overall, including social science research as an integral part in R&D programmes on GD and fostering collaboration between different types of knowledge in the R&D agenda-setting phase may help in the consideration of GD as a socio-technical problem, where the social and the technical are inextricably linked.

Many countries encourage national forums for transparency, dialogue and participation with regards to radioactive waste disposal. However, the local actors (authorities, non-government organisations and the public) often note a lack of public participation in the decision making process. Civil society is often frustrated with its limited involvement in the consultative process. Participation is regulated by national laws and rules and the right to participate in environmental decision-making is covered by the Aarhus Convention. Continuous dialogue amongst stakeholders is seen as important in building sustainable solutions in radioactive waste management. In addition, understanding public concerns and needs can increase the trust between the partners and build confidence in the process.

Different national and local contexts have contributed to the development of quite a broad set of approaches and tools for stakeholder engagement. This paper describes the use of such tools in the engagement with the Saligny community in the siting process of a repository for low- and intermediate-level wastes in Romania. Some specific issues are highlighted such as: the low level of interest amongst the public in relation to long-term projects; over-estimation of benefits in comparison to the negative aspects of hosting a repository; lack of a coherent public voice; and a perceived lack of information on the project from the authorities and the implementer. The present study describes the setting up of the participatory approach to engage with the public and the different methods employed (including citizen juries, workshops, open days, etc.). A number of criteria were developed for evaluating the effectiveness of these methods particularly with regards to their adaptability to a local context such as Saligny. The paper then focuses on the results of one of these methods – the use of focus groups covering a cross-section of civil society – including members of the general public, a group of professionals and a group of local councillors. The study has resulted in a number of recommendations to the implementer on how to build a new programme for public participation.

In Lithuania all the spent nuclear fuel (SNF) came from operation of the Ignalina nuclear power plant with two reactors of RBMK type (RBMK is a Russian acronym for 'Channelized Large Power Reactor' which is a water-cooled graphite-moderated reactor: RBMK-1500). Approximately 22,000 SNF assemblies are due for geological disposal in Lithuania. Currently it is envisaged that SNF will be stored in dry interim storage facilities (new and existing) for at least 50 y prior to possible deep geological disposal.

The decision on the final SNF management option (disposal in a national repository, disposal in regional repositories, etc.) has not yet been made but some investigations of the possibilities to dispose of the SNF in Lithuania have been initiated. With the support of Swedish experts, analysis of possible geological formations for SNF disposal was performed and the existence of potentially suitable formations agreed. The geological formations prioritized as prospective include the crystalline rocks in southern Lithuania and two clayey formations: the Lower Triassic clay formation and the Lower Cambrian Baltic Group clay formation, with priority given to the Lower Triassic clay formation.

This paper presents the main aspects of the research and other activities undertaken over the past decade in the field of SNF disposal: international cooperation; current status and plans for the Lithuanian national program; further investigations required; and competence developments.

The Implementing Geological Disposal of Radioactive Waste Technology Platform (IGD-TP) was launched in November 2009 to facilitate international cooperation in common areas of research, development and demonstration (RD&D) with a view to advancing the implementation of geological disposal facilities for spent fuel, high-level and other long-lived waste in Europe.

The IGD-TP's Vision is that “by 2025, the first geological disposal facilities for spent fuel, high-level waste and other long-lived radioactive waste will be operating safely in Europe”. Aside from most European waste management organisations, the IGD-TP currently has 124 members covering most of the RD&D actors in the field of implementing geological disposal of radioactive waste in Europe.

Five years after its inception, the IGD-TP has been shown to play a leading role in coordinating joint actions for RD&D in radioactive waste geological disposal programmes. The work of the platform takes into account differences between the timing and challenges for the respective waste management programmes. Following implementation of Posiva's geological disposal facility in Finland it is expected that within the next 5 years the construction of the Swedish and French geological disposal facilities will commence. Within IGD-TP, the SecIGD2 project whose remit is “Coordination and Support Action under the 7th Framework programme” aims at supporting, at the European level, the networking and structuring of RD&D programmes and competences in countries with less advanced geological disposal programmes, including those in the new European Union Member States. Furthermore, the SecIGD2 supports the development and coordination of the necessary competences to meet the Vision 2025 as a part of the platform's Competence Maintenance, Education and Training (CMET) working group.

Geological disposal (GD) of radioactive waste is close to becoming a reality for Finland, Sweden and France. High-technology development and advanced knowledge has made it possible to defend the feasibility and the safety of such facilities, making the European Union a leader in the field. Other European countries are closely behind, developing high competence through advanced research programmes, research infrastructures and public engagement.

At the other extreme, there are countries whose GD programmes are at an early stage and no systematic research programmes exist. These include several new Member States but not the Czech Republic and Hungary, both of which have already initiated a siting process.

There are several common reasons for this delay in schedule: small and relatively younger nuclear energy programmes, return of the spent fuel (especially from research reactors) to the countries of origin, open fuel cycle concept (requiring at least 50 years of wet and dry storage). In this context, there has been little pressure on setting up an early GD programme. Currently their disposal concepts are only generic and in most of these countries need updating, taking into account the current socio-economic context.

However, some of these new Member States still aim to have a GD in operation within several decades, e.g. 2055 in Romania and 2067 in Slovenia. Strategic planning based on the experience of more advanced programmes shows the GD process should start immediately in order to be able to achieve these deadlines.

In this context, the implementation of the EC Directive 70/2011 gives the opportunity to progress the advancement of the GD process in these countries.

In safety assessments for nuclear waste disposal, the biosphere is a completely open system, whereas the bedrock can be treated with comparably simple boundary conditions. The bedrock has a vital role in providing and maintaining favourable conditions for the waste, but the public interest tends to focus on the biosphere. More importantly, the bedrock groundwater does arrive from the biosphere. Also, the regulations usually set the safety criteria in terms of doses occurring in the biosphere. Thus, it is reasonable to address the biosphere as a part of the disposal system, performance of which should be described based on conceptual models that cover phenomena and processes controlling radionuclide release and transport. From these, the actual assessment models and data are expected to be derived based on high-level research and expertise obtained through empirical study. In this paper, the complexity expected of a biosphere assessment within a mature safety case is discussed through the example of the Finnish regulations and recent safety cases for nuclear waste disposal. Both elaborate and complex presentations and stylised approaches have their benefits, and a balance needs to be sought.

The extended timescales involved in the decay of radioactive wastes to safe levels mean that geological disposal facilities must continue to function effectively long into the future. It is therefore essential to consider long-term climate evolution in post-closure performance assessments in order to evaluate a geological disposal system's response and robustness to a variety of potential environmental changes, driven by both natural and anthropogenic forcings. In this paper, we illustrate the multiple decay components that characterize the primary driver of climate change – atmospheric CO2 – in response to fossil fuel carbon emissions. We perform a multi-exponential analysis on a series of atmospheric CO2 decay curves predicted by an Earth system model and create an empirical response function that encapsulates the long-term (>1 kyr) removal of excess CO2 from the atmosphere. We present this response function as a simple tool for rapidly projecting the future atmospheric CO2 concentration resulting from any plausible cumulative release of CO2. We discuss the implications of the long 'tail' to this atmospheric CO2 decay curve, both in terms of future climate evolution as well as potential impacts on radioactive waste repositories.

Heat-generating waste provides a number of additional technical challenges over and above those associated with the disposal of ILW. A priority area of work for Radioactive Waste Management (RWM) concerns the effect of heat on the engineered barrier system, and how this may be mitigated through the management of heat (thermal dimensioning) in a UK Geological Disposal Facility (GDF). The objective of thermal dimensioning is to provide a strategy to enable acceptable waste package loading and spatial configurations of the packages to be determined in order to enable high-heat generating waste to be successfully disposed in a GDF. An early focus of the work has been to develop a thermal modelling tool to support analyses of different combinations of package assumptions and other GDF factors, such as spacing of those packages, to assess the compliance with thermal limits. The approach has a capability to investigate quickly and efficiently the implications of a wide range of disposal concepts for the storage of spent fuel/HLW and the dimensions of a GDF. This study describes the approach taken to undertaking this work, which has included a robust appraisal of the key data (and the associated uncertainty); recent thermal dimensioning analysis has been performed to identify constraints on those disposal concepts.

The UK has published a generic Disposal System Safety Case for a geological disposal facility (NDA, 2010) and is planning to update this in 2016. However, it is a challenge to present a meaningful safety case when the location and hence the design of a geological disposal facility are not known. Consequently, this paper describes our aim to present a narrative, explaining how we can have confidence in the long-term safety of a geological disposal facility. This narrative is based on an understanding of the environmental safety functions of a geological disposal facility and the features, events and processes (FEPs) that support them. The highest level environmental safety functions required for a geological disposal facility are isolation and containment. By isolation we mean removal of the wastes from people and the surface environment. By containment we mean retaining the radioactivity from the wastes within various parts of the disposal facility for as long as required to achieve safety. Beneath these top-level environmental safety functions we have identified generic environmental safety functions associated with each of the key safety barriers within a geological disposal facility, namely: the wasteform, the container, the local buffer or backfill, the mass backfill (in the access tunnels and service ways), the plugs and seals and the geosphere. This paper discusses the application of environmental safety functions and FEPs to building a safety narrative and explains how it is proposed to use such an approach to develop a generic environmental safety case for the UK to provide confidence in the longterm safety of a geological disposal facility after it has been sealed and closed.