Deep Geological Waste Repository Research Articles

Physico-mechanical properties of halite and gypsum–mudstone crushed rock backfills of the Mercia Mudstone Group

Permanent disposal of radioactive waste in deep geological repositories (DGRs) relies on engineered barriers, such as crushed rock backfills, to ensure that wastes are isolated and contained. The performance of crushed rock backfill, during construction, operation and closure, relies on optimizing its mechanical properties. This paper explores how density and composition influence the strength of (i) crushed halite and (ii) crushed gypsum–mudstone mixtures of the Triassic Mercia Mudstone Group, which is a potential DGR host rock in the UK. Unconfined compressive strength (UCS) testing of compacted halite demonstrated that strength correlated with compaction density, with UCS increasing from 220 kPa at a density of 1.72 g cm −3 up to 6600 kPa at the maximum achieved density of 1.83 g cm −3 . Direct shear testing revealed an angle of internal friction for crushed halite samples of 39°–40° at a density of 1.5 g cm −3 . Direct shear testing of gypsum–mudstone composites from 100% gypsum to 100% mudstone revealed that higher gypsum percentages provide greater frictional shearing resistance, whilst higher mudstone percentages promote more cohesive strength. Data are also presented on bulk materials characterization by X-ray fluorescence (XRF), scanning electron microscopy with energy-dispersive spectrometry (SEM-EDS) and optical microscopy. This research provides insights into how the strength properties of compacted crushed halite and gypsiferous mudstone backfills can be engineered through control of density and composition.

Stress state estimation – new data and variability assessment of model results

Abstract. The upper crust of the Earth is used more and more to transport and extract raw materials and energy. It is also used for the final disposal of radioactive waste in deep geological repositories with a particularly crucial requirement for long-term safety over 1×106 years. An accurate knowledge of the hosting rock and surrounding formation coupled with comprehensive modelling is fundamental to demonstrate the geological site has the required properties for safe and long-term underground storage. Besides other criteria, geomechanics play an important role. Even if the determination of rock properties is already a challenge, the estimation of the contemporary stress state in the upper crust is even more challenging. To predict the stress state 3-D geomechanical–numerical models are used, but high-quality underground structure, rock properties and in situ stress data for model calibration are needed. In particular, data of maximum and minimum horizontal stress magnitudes (SHmax and Shmin) are required to find suitable lateral displacement boundary conditions to produce a best-fit model. During the recent exploration phase for a deep geological repository for radioactive waste in Switzerland, a unique dataset of stress magnitude data has been acquired from eight cored boreholes. Rock mechanical properties were constrained from geophysical logging and laboratory testing. The empirically correlated rock properties were not simply averaged, but a probability distribution was provided. The stress field was explored by conducting more than 120 micro-hydraulic fracturing (MHF) and dry sleeve re-opening (SR) tests in different stratigraphic units, to estimate the magnitudes of Shmin and SHmax. We present the results of a 3-D geomechanical–numerical model that shows the best fit with respect to the measured stress magnitudes. Considering the uncertainties of the MHF/SR tests and the ones resulting from rock property variability, the model can reproduce most of the measurements. However, we do show not only the best-fit result but also a bandwidth of individual stress components within a P05–P95 probability range. We also show that there are still outliers for this bandwidth, but we will show that these can be explained by scaling effects.

Investigating the influence of bentonite colloids on strontium sorption in granite under various hydrogeochemical conditions

The disposal of high-level radioactive waste in deep geological repositories is a critical environmental issue. The presence of bentonite colloids generated in the engineering barrier can significantly impact the transport of radionuclides, but their effect on radionuclide sorption in granite remains poorly understood. This study aimed to investigate the sorption characteristics of strontium (Sr) on granite as well as on the coexistence system of granite and colloids under various hydrogeochemical conditions, through batch experiments. Fourier transform infrared spectroscopy was employed to analyze the sorption forms of Sr on granite before and after sorption. Several hydrogeochemical factors were examined, including contact time, pH, ionic strength, coexisting ions, and bentonite and humic acid colloid concentration. Among these factors, the concentration of bentonite colloids exhibited a significant effect on Sr sorption. Within a specific range of colloid concentration, the sorption of Sr on the solid system increased linearly with the bentonite colloid concentration. pH and ionic strength were also found to play crucial roles in the sorption process. At low pH, Sr sorption primarily occurred through the outer sphere's surface complexation and Na+/H+ ion exchange. However, at high pH, inner sphere surface complexation dominated the process. As the ionic strength increased, electrostatic repulsion gradually increased, resulting in fewer binding sites for particle aggregation and Sr sorption on bentonite colloids. The results also indicate that with increasing pH, the predominant forms of Sr in the solution transitioned from SrHCO3+ and SrCl+ to SrCO3 and SrCl+. This was mainly due to the ion exchange of Ca2+/Mg2+ in plagioclase and biotite, forming SrCO3 precipitation. These findings provide valuable insights into the transport behavior of radionuclides in the subsurface environment of the repository and highlight the importance of considering bentonite colloids and other hydrogeochemical factors when assessing the environmental impact of high-level radioactive waste disposal.

Experimental factors and their impact on understanding corrosion in anoxic environments

AbstractThe long‐term disposal of nuclear waste in deep geological repositories requires safe containment for up to one million years. To understand how the waste will respond to the gradually evolving environment, researchers can perform sensitive experiments in the laboratory to replicate the repository conditions at specific time points. In Belgium, spent nuclear fuel will be housed within a carbon steel overpack, which is encased within a cement buffer, which may include a stainless steel liner (the “envelope”). The range of corrosion behaviors for carbon steel, and to a lesser extent, stainless steel, has been studied through hydrogen evolution, an end‐product of oxidation under repository conditions. Considerable time may be required for metallic surfaces to approach a steady hydrogen evolution and this is dependent upon the starting condition of the specimen surfaces. The use of nominally‐similar materials can result in significant variation in the reported corrosion rates. The presence of cements, which generate their own hydrogen and influence water chemistry as a function of time, add complexity to even simple experimental configurations that requires careful interpretation.

Confronting the Uncertainties Associated with Long-Time Scales: Analysis of the Modes of Preservation of Memory of Radioactive Waste Burial Sites

The burial of radioactive waste in deep geological repositories has gradually imposed itself since the 1980s in various countries. Considered more stable, safe, and responsible than storage above ground, this solution is seen as a way of keeping the waste out of human reach, and of freeing oneself of the obligation to monitor repositories. However, it soon became clear that the idea of relying on the relative stability of deep geological repositories for the safe disposal of nuclear waste to remove it from the ‘all too human’ risks associated with history’s turbulences, has produced new uncertainties, and raised new questions, given the multi-millennial time scales involved. This article presents a study of the strategies adopted by the actors in charge of radioactive waste management in the face of the temporal constraints imposed by the length of their radioactive life. More specifically, it is intended to question the representations of temporality that allow the designers of these projects to assume their responsibilities by transmitting information and warnings towards to an extremely distant future, to recipients who are not easily representable and for whom they already have a responsibility even though they do not yet exist.

Adsorption Properties of Cs(I) and Co(II) on GMZ Bentonite Colloids

The alteration of the morphology of bentonite, a key barrier to isolating high-level radioactive waste in deep geological repositories under long-term interaction with groundwater, was studied. Bentonite colloids were prepared from Gaomiaozi bentonite, and its various properties were analyzed by various characterization methods. The effects of various factors on the adsorption of Cs(I) and Co(II) on bentonite colloids were investigated by batch static adsorption experiments. The results showed that the colloids were mainly composed of montmorillonite and were lamellar in shape, with a zeta potential of −enton mV and an average size of approximately 209.10 nm. The adsorption of Cs(I) and Co(II) by the colloids was a rapid process, and 31.78 and 88.24 mg/g, respectively, were adsorbed at equilibrium. Pseudo-second-order kinetic fitting showed that chemisorption plays a dominant role and acid-base interactions affect adsorption by influencing the stability of colloids and the chemical form of simulated nuclides. This work can be helpful for evaluating the safety of waste repositories.

Exploring the local environment of the engineered nanoclay Mica-4 under hydrothermal conditions using Eu3+ as a luminescent probe

High charge mica Na4Al4Si4Mg6O20F4, Mica-4, is a promising candidate as a filling material to immobilize high-level radioactive waste in deep geological repositories due to its extraordinary adsorption capacity. In contrast to traditional clay materials, the structural composition of this mica, with a high content of aluminum in the tetrahedral sheet, enhances its chemical reactivity, favoring the formation of new crystalline phases under mild hydrothermal conditions, and thus providing a definitive isolation of the radionuclides in the engineered barrier. Moreover, this synthetic clay has some features that allow its use as an optical sensor by doping with luminescent rare earth cations such as Eu3+. In this paper we discuss the local structure of the nanoclay Mica-4 using Eu3+ as a local probe to track the physical and chemical modifications under hydrothermal conditions. For that purpose, a set of hydrothermal experiments has been carried out heating Mica-4 and an aqueous Eu(NO3)3 solution in a stainless steel reactor at different temperatures and times. Optical properties of the as-treated samples were characterized by spectroscopic measurements. The fine peak structure of emission and the relative intensity of different Eu3+ transitions as well as the luminescence lifetime have been correlated with the structure and composition of this nanoclay, and the interaction mechanisms between the lanthanide ions and the clay mineral at different temperatures and times. Special attention has been paid to understanding the role of the aluminum content, which may act as either an aggregating or dispersing agent, in the optical features and reactivity of the system.

Anaerobic corrosion of carbon steel in compacted bentonite exposed to natural Opalinus clay porewater: Bentonite alteration study

Abstract. Carbon steel is a potential canister material for the disposal of high-level radioactive waste in deep geological repositories in clays and clay rocks. Bentonite is considered as a potential backfill material for those multi-barrier systems. To predict the long-term performance and for safety assessment the knowledge of canister corrosion behavior is important. The corrosion products formed and mineralogically altered bentonite at the canister/bentonite interface can potentially provide an additional barrier against radionuclide migration. In-situ corrosion experiments were performed at the Mont Terri underground research laboratory. Coupons of carbon steel were embedded in Volclay MX-80 bentonite with controlled densities, installed in a borehole under simulated repository and anaerobic conditions and exposed to natural Opalinus clay porewater for a period up to 5.5 years (Smart et al., 2017). In the present study, the bentonite layer at the canister/bentonite interface was characterized by complementary microscopic and spectroscopic techniques (XPS, SEM-EDX, µXANES) under anoxic conditions. The interface revealed reddish-brown staining up to 2 mm depth into the bentonite in the zone adjacent to the steel in all three obtained samples. The XPS analysis revealed formation of sulfides at the interface consisting of iron and other trace metals present in the steel. The SEM-EDX analyses of the interface (embedded cross-cut with steel removed) showed different degrees of calcium enrichment in the bentonite adjacent to the metal for various samples. The µXRF analysis performed on the bentonite at the interface showed a scarce or distinct calcium-enriched rim up to 100 µm into the bentonite and iron-enriched rim depending on the sample (one sample in Fig. 1), while µXANES analysis revealed formation of iron silicate compounds in the reacted reddish-brown zone. The iron appears to displace calcium from the interlayer sites in montmorillonite. The calcium then precipitates at the interface as calcite. The extent of this process seems to be strongly related to the bentonite density. The steel coupon was removed prior to embedding, with the location marked as resin in Fig. 1. A line scan from the edge towards the bulk bentonite did not indicate any systematic gradient in the Fe2+/3+ ratio. The formation of mixed Fe2+/3+ silicate compounds appears to be heterogeneous. This work contributes to an increasing understanding of steel corrosion mechanisms in clay, which can improve the robustness of canister lifetime predictions.

Lessons from archaeology and heritage studies for the long-term preservation of records, knowledge and memory concerning deep geological disposal sites for nuclear waste

Abstract. Safe disposal of nuclear waste in deep geological repositories requires secure knowledge transfer or knowledge recovery in time spans of many tens of thousands of years. Never before has any detailed record, knowledge or memory been reliably preserved or recovered over comparable time periods. This challenge has been extensively addressed since the late 1980s, initially during the SANDIA workshops in the USA and more recently in the Nuclear Energy Agency/Organisation for Economic Co-operation and Development (NEA/OECD) project on Preservation of Records, Knowledge and Memory Across Generations (Schröder, 2019). Experts from many disciplines including engineering, the natural sciences, information technology, social studies of science and technology, semiotics, public management, and design as well as artists have contributed to these discussions. Some scholars from the humanities have been involved in working on these issues, especially in recent decades. At the same time, much of the existing work has drawn on assumptions about human history, archaeological monuments and cultural heritage that have been scrutinized and deemed deeply flawed by Joyce (2020). The authors of the present paper are archaeologists and cultural heritage experts. For the past decade, they have been working with the challenge of preserving records, knowledge and memory concerning deep geological disposal sites for nuclear waste (Holtorf and Högberg, 2021). From the perspective of the human sciences, in particular archaeology and heritage studies, the unique task at hand involves not only the previously recognized challenges that require consideration of long-term material durability, linguistic intelligibility, and appropriate sense-making of any communicated information but also two challenges not previously addressed: Human action as informed by cultural and social processes. In designing of various long-term mechanisms, we risk overlooking that what people will do is not going to be governed by mechanics. How human beings learn, reason, value, decide, and act is informed by specific cultural and social processes creating context and meaning. We must avoid ignoring these complexities governing human thinking and agency. This challenge requires more work on understanding how sentient and intelligent beings like humans act in variable contexts across time and space. Our anticipatory assumptions. A proverb states that “nothing ages faster than the future”. In making assumptions about future generations' understandings, meanings, and significances of our nuclear waste we risk “colonizing” the future, fail to embrace variability over time, and miss realizing multiple futures and emerging conditions. We must therefore not foreclose uncertain futures but instead create circumstances favorable for change and transformation of relevant knowledge and memory. This challenge requires more work with processes of translation between generations. The challenges of assessing our anticipatory assumptions and understanding how humans act will also need to be addressed in transmitting records, knowledge and memory for the benefit of future generations.

Dissolution of simplified nuclear waste glass and formation of secondary phases

Abstract. Immobilization of high-level and intermediate-level nuclear wastes by vitrification in borosilicate glass is a well-established process. There is a consensus between the waste management agencies of many countries and many experts that vitrified nuclear waste should be disposed of in a deep geological waste repository and therefore its long-term behavior needs to be taken into account in safety assessments. In contact with water, borosilicate glass is metastable and dissolves. In static dissolution experiments, often a surface alteration layer (SAL) forms on the dissolving glass, and later sometimes secondary phases form. Based on boron or lithium release rates, commonly three stages of glass dissolution are defined as a function of the reaction progress: (I) initial dissolution, described by a congruent glass dissolution at the highest rate, (II) residual dissolution, characterized by a glass dissolution rate several orders of magnitude lower than the initial one, and (III) resumption of glass alteration with initial rates. Microscopically, the formation of a complex SAL has been identified as a prerequisite for the slower dissolution kinetics of stage II. Stage III is typically observed under specific conditions, i.e., high temperature and/or high pH driven by the uptake of Si and Al into secondary phases. Different glass dissolution models explaining the mechanisms of the SAL formation and rate-limiting steps have been proposed and are still under debate. In this article different aspects of glass dissolution from recent studies in the literature and our own work are discussed with a focus on the microscopic aspects of SAL formation, secondary phase formation and the resumption of glass dissolution. Most of the experiments in the literature were performed under near-neutral pH conditions and at 90 ∘C, following standard procedures, to understand the fundamental mechanisms of glass dissolution. The example of interaction of glass and cementitious materials as discussed here is relevant for safety assessments because most international concepts include cement e.g., as lining, for plugs, or as part of the general construction of the repository. The aim of the investigations presented in this paper was to study the combined effect of hyperalkaline conditions and very high surface area/volume ratios (SA/V=264000m-1) on the dissolution of international simplified glass (ISG) and the formation of secondary phases at 70 ∘C in a synthetic young cement water containing Ca (YCWCa). The new results show that the SA/V ratio is a key parameter for the dissolution rate and for the formation of the altered glass surface and secondary phases. A comparison with similar studies in the literature shows that especially on the microscopic and nanoscale, different SA/V ratios lead to different features on the dissolving glass surface, even though the SA-normalized element release rates appear similar. Zeolite and Ca-silicate-hydrate phases (CSH) were identified and play a key role for the evolution of the solution chemistry. A kinetic dissolution model coupled with precipitation of secondary phases can be applied to relate the amount of dissolved glass to the evolution of the solution's pH.

Uranyl binding mechanism in microcrystalline silicas: A potential missing link for uranium mineralization by direct uranyl co-precipitation and environmental implications

Genetic models for the formation of uranium deposits almost invariably invoke the reduction of U(VI) to U(IV) as the deposition mechanism. However, the questions of when and how this reduction occurred are often not clear in most uranium deposits. For example, mineralization in the giant Olympic Dam Cu-U-Au-Ag deposit (Australia) occurs mainly in hematite-rich breccias of the oxidizing hematite-quartz-barite assemblage. Similarly, unconformity-related uranium deposits in the Athabasca Basin (Canada) are accompanied by extensive and intensive hematite- and quartz-rich alteration halos, pointing to a possible linkage between uranium mineralization and silicification. However, this linkage is not understood due to the lack of knowledge about the speciation and mechanism of uranium incorporation in microcrystalline and macrocrystalline silicas. In this contribution, microcrystalline silicas (agate and opal-CT) containing 27–137 ppm U have been investigated by microbeam synchrotron X-ray fluorescence mapping, microbeam synchrotron X-ray diffraction analysis, synchrotron X-ray absorption spectroscopy, Raman spectroscopy, Fourier transform infrared spectroscopy, and electron paramagnetic resonance spectroscopy. These data unravel the binding mechanism of uranyl silicate complexes in microcrystalline silicas and suggest a new mechanism of uranium deposition involving uranyl co-precipitation without the putative reduction process, providing a potential missing link between uranium mineralization and silicification for the formation of uranium deposits. In addition, new data on the uptake and long-term retention of elevated uranyl contents in microcrystalline silicas extend previous findings from their amorphous counterparts on these materials as an effective sink for uranium attenuation to higher temperatures, important to the storage and disposal of heat-generating nuclear waste in deep geological repositories.

Effects of contact time, pH, and temperature on Eu(III) sorption onto MX-80 bentonite

MX-80 bentonite is considered a popular buffer/backfill material for the construction of artificial barriers. In particular, it plays an important role in the disposal of high-level nuclear waste in deep geological repositories. In the present study, the sorption of Eu(III) on MX-80 bentonite was investigated using a batch technique; furthermore, a comprehensive analysis was performed on the influence of contact time, pH, temperature, and concentration on sorption. The results showed that the sorption regime leveled off within 30 min and the Eu(III) sorption capacity of MX-80 bentonite was approximately 46.06 mg/g under certain conditions. A second-order kinetic equation was used to fit the equilibrium sorption data. Additionally, the variation in the sorption capacity of MX-80 bentonite increased as the pH increased from 1 to 9. The optimum pH value was determined to be 7.0 at a temperature of 293.15 K. The negative value of enthalpy change (ΔH°) indicates that the temperature had a negative effect on the sorption. The sorption process was most accurately described by the Langmuir isotherm model. These results will definitely be benefit to the artificial barriers in the disposal of high-level nuclear waste.

Microbial diversity in bentonite, a potential buffer material for deep geological disposal of radioactive waste

Many countries are considering long-term disposal of nuclear waste in deep geological repositories (DGRs), encapsulated in metal containers, surrounded by a barrier, and emplaced in the host rock. Bentonite is found to be a safe barrier material because of its physical and chemical properties. The bentonite montmorillonite interlayer contains adsorbed organic matter and microorganisms, and therefore biogeochemical activity must be considered. This study focused on investigating the microbial diversity in Chinese bentonite by 16S rRNA analysis via Illumina sequencing to determine how these microorganisms can potentially influence DGR performance. The bentonite sample contained high microbial diversity, dominated by Archaea and abundant other bacteria. The community of Archaea included Nitrososphaera, Thermogymnomonas, and Methanobrevibacter, which are thought to play significant roles in nitrogen cycling, H2 and O2 consumption, CH4 production, and organic matter degradation. These processes can generate excessive pressure and may compromise the safety of DGRs, and therefore the use of bentonite as a barrier material should be carefully considered.

Earthly graves for environmental futures: Techno-burial practices

In this paper we explore contemporary investments in the underground. We argue that alongside excavation processes the subsurface the underground is increasingly being presented as site for the burial of toxic and radioactive waste and carbon dioxide. We explore two particular burial practices; the long-term disposal of industrial and toxic waste in deep geological repositories and the sequestration of carbon and greenhouse gases in soil. Across these practices, we argue that subsurface appears as a repository for both the sacred and profane remains of environmental pasts and the hopes for an ecological modernity.

Nuclear Corrosion: Achievements and Challenges

Corrosion science faces new challenges in various nuclear environments. Three main areas may be identified where increases of knowledge and understanding have been done and are still needed to face the technical needs: (i) the extension of the service time of nuclear power plants from 40 years, as initially planned, to 60 years and probably more as expected now, (ii) the prediction of long term behaviour of metallic materials in nuclear waste disposal where the corrosion processes have to be predicted over large periods of time, some thousands years and more, (iii) the choice of materials for use at very high temperatures as expected in Generation IV power plants in environments like gas (helium), supercritical water, liquid metals or salts. Service time extension, deep geological waste repositories and high temperature reactors sustain researches and developments to model corrosion phenomena at various scales, from atoms to components.

A comparative study of engineering options for the disposal of high-level radioactive waste with regard to thermal effects and chemical degradation

ABSTRACTThe disposal of high-level radioactive waste in deep geological repositories requires stable and foreseeable physical conditions over very long time scales. During this period, the chemical stability of both the natural and the engineered barriers is governed by thermally activated processes. These in turn are driven by the heat pulse generated by the nuclear decay of waste products. The technical concept to cap the temperature peak in the repository is thus an important aspect for the proof of safety of disposal facilities. It is shown that densely stocked repositories, as currently foreseen in several countries, do not necessarily represent the optimal choice with regard to temperature effects, long-term reaction kinetics and chemical degradation of components. It is suggested that the optimization of the temperature peak rather than the fulfillment of cut-off conditions for peak temperature should be a cardinal issue in engineering concepts.

Natural analogues: studies of geological processes relevant to radioactive waste disposal in deep geological repositories

The geological disposal of radioactive wastes is generally accepted to be the most practicable approach to handling the waste inventory built up from over 70 years accumulation of power production, research–medical–industrial and military wastes. Here, a brief overview of the approach to geological disposal is presented along with some information on repository design and the assessment of repository post-closure safety. One of the significant challenges for repository safety assessment is how to extrapolate the likely long-term (i.e. ten thousand to a million years) behaviour of the repository from the necessarily short term data from analytical laboratories and underground rock laboratories currently available. One approach, common to all fields of the geosciences, but also in such diverse fields as philosophy, biology, linguistics, law etc., is to utilise the analogue argumentation methodology. For the specific case of radioactive waste management, the term ‘natural analogue’ has taken on a particular meaning associated with providing supporting arguments for a repository safety assessment. This approach is discussed here with a brief overview of how the study of natural (and, in particular, geological) systems can provide supporting information on the likely long-term evolution of a deep geological waste repository. The overall approach is discussed and some relevant examples are presented, including the use of uranium ore bodies to assess waste form stability, the investigation of native metals to define the longevity of waste containers and how natural clays can provide information on the stability of waste tunnel backfill material.

A sensitivity and probability analysis of the safety of deep geological repositories situated in crystalline rock

The aim of this paper is to present the sensitivity and probability analysis of a model of the Czech safety concept of deep geological waste repository. The sensitivity analysis was performed for a set of critical radionuclides, 14C, 36Cl, 79Se, 129I and 226Ra using a variance based method. The main contributors to the variance of model output were identified. It was shown that, if the model output is properly transformed before performing sensitivity analysis, then the calculated sensitivity coefficients are robust indicators of the importance of uncertainties of the individual parameters.

Clays in Natural and Engineered Barriers for Radioactive Waste Confinement: an introduction

The concept of engineered geological disposal has been developed for the safe long-term management of long-lived radioactive waste. This involves emplacement of radioactive waste in deep geological repositories that contain and isolate the waste and, consequently, protect humans and the environment. The ‘multiple barrier concept’ is the cornerstone of all proposed schemes for the geological disposal of radioactive waste. Based on the principle that uncertainties in performance can be minimized by conservatism in design, the concept invokes a series of complementary barriers, both engineered and natural (geological), between the waste and the surface environment. Each successive barrier represents an additional impediment to the movement of radionuclides. Regarding engineered barriers, for lowand intermediate-level waste the waste may be incorporated in a relatively stable and inert matrix such as cement, bitumen, lead-alloy or polymer resin (the choice varying depending on the waste management organization); glass may be used in the case of certain high-level reprocessing wastes. Owing to the very low leach-rate of glass in groundwater, vitrification is widely accepted to be one of the best methods of immobilizing the aqueous products from the reprocessing of spent fuel. Many waste containers will provide some form of physical barrier to groundwater. However, because of the relatively small volumes of waste involved, spent fuel, vitrified waste and other highly active wastes will be totally encapsulated in corrosion-resistant metal canisters that are designed to prevent groundwater entry for extended time periods in excess of 100 000 years. Depending on the disposal concept, engineered barriers may comprise the buffer/backfill medium enclosing the waste containers, the tunnel/borehole liner, and the backfill and high-integrity seals placed in the repository access ways or emplacement boreholes. The buffer/backfill medium enclosing the waste will often also provide both a physical and a chemical barrier to radionuclide migration. The functions of the engineered/chemical barriers are: † to reduce the rate of corrosion of the waste containers and thus extend their life; † to limit the rate of hydraulic transport; † to limit the release of radionuclides from the waste-form to the far-field (geosphere) after container failure; † to limit the migration of radionuclides along the pathway provided by the access tunnels and shafts of a repository or the boreholes in the case of a deep borehole emplacement.

Lanthanide sorption on smectitic clays in presence of cement leachates

Due to their potential retention capacity, clay minerals have been proposed for use in the engineered barriers for the storage of high-level radioactive actinides in deep geological waste repositories. However, there is still a lack of data on the sorption of actinides in clays in conditions simulating those of the repositories. The present article examines the sorption of two lanthanides (actinide analogues) in a set of smectitic clays (FEBEX bentonite, MX80 bentonite, hectorite, saponite, Otay montmorillonite, and Texas montmorillonite). Distribution coefficients ( K d) were determined in two media: water and 0.02 mol L −1 Ca, the latter representing the cement leachates that may modify the chemical composition of the water in contact with the clay. The K d values of the lanthanides used in the experiments (La and Lu) varied greatly (25–50 000 L kg −1) depending on the ionic medium (higher values in water than in the Ca medium), the initial lanthanide concentration (up to three orders of magnitude decrease inversely with lanthanide concentration), and the examined clay (up to one order of magnitude for the same lanthanide and sorption medium). Freundlich and Langmuir isotherms were used to fit sorption data to allow comparison of the sorption parameters among smectites. The model based on the two-site Langmuir isotherms provided the best fit of the sorption data, confirming the existence of sorption sites with different binding energies. The sites with higher sorption affinity were about 6% of the total sorption capacity in the water medium, and up to 17% in the Ca medium, although in this latter site sorption selectivity was lower. The wide range of K d values obtained regarding the factors examined indicated that the retention properties of the clays should also be considered when selecting a suitable clay for engineered barriers.