Measuring the effective diffusion coefficient of dissolved hydrogen in saturated Boom Clay

For the long-term management of high- and intermediate-level radioactive waste and/or spent fuel, many countries prefer disposal in a geological repository. In Belgium, Switzerland and France, argillaceous formations are being explored as potential host formations for this purpose. In this context, knowledge of the diffusion coefficient of He is relevant within two research areas: first, diffusion coefficients are used in safety calculations to evaluate the balance between gas generation (mainly H 2 ) and gas dissipation; and, second, the diffusion coefficients of He and Ar are needed in the diffusion models of natural tracers. Owing to the lack of data on the diffusion coefficients of He and Ar for the different clay host formations, diffusion experiments with dissolved He and Ar were performed on Boom Clay, Opalinus Clay and Callovo-Oxfordian Clay. Samples were confined in a diffusion cell, and diffusion coefficients were measured by using the double through-diffusion technique. The diffusion coefficients ( D p or D pore ) for He in Boom Clay, Callovo-Oxfordian Clay and Opalinus Clay are 12.6×10 −10 , 4.51×10 −10 and 7.13×10 −10 m 2 s −1 , respectively. The diffusion coefficients for Ar in Boom Clay, Callovo-Oxfordian Clay and Opalinus Clay are 18.6×10 −11 , 4.06×10 −11 and 3.65×10 −11 m 2 s −1 , respectively.

Boom Clay is considered to be a potential host formation for a geological repository of radioactive waste in Belgium. In this context, a thorough understanding of its mechanical behaviour is thus important. In this study, the anisotropy of the shear modulus of natural Boom Clay samples was investigated using the bender element method. Seven samples located at different distances from the connecting gallery of the HADES underground research laboratory were tested for shearing in three different directions with respect to the bedding plane. The results obtained confirm the mechanical anisotropy of Boom Clay. Furthermore, the values of small-strain shear modulus in the near field of the gallery were found to be lower than those in the far field, suggesting that the mechanical properties of Boom Clay remain affected by gallery excavation damage that occurred more than 10 years ago.

Gas migration in porous media in the context of nuclear waste disposal is dominated by diffusion. The availability of abundant data on diffusion coefficients of different gases in many clay based materials such as Boom Clay, Opalinus Clay and Callovo-Oxfordian Clays provides a good overview of the rates of gas transport through these different clay rocks [1]. Such formations are being considered as potential host rocks for the disposal of high and intermediate level radioactive wastes across Europe. Diffusion coefficients have been studied with particular interest due to the unavoidable gas generation phase in the nuclear waste repository [2]. However, all the diffusion measurements performed so far have been under fully saturated conditions, with little experimental data on gas diffusivities in partially saturated conditions, which have been hypothesized to exist during the early part of the life cycle of the repository [3]. The objective of the current study is to provide an overview of the effect of desaturation on the rate of diffusive transport of gases in partially saturated clay based materials.
 The experimental methodology is based on the double through diffusion setup used in previous studies to measure gas diffusivities in saturated clays [1]. However, the setup is modified so as to keep a fixed level of unsaturation in the clay sample. This is done by employing the vapour equilibrium technique, using oversaturated salt solutions in both upstream and downstream reservoirs to maintain a constant relative humidity in the entire setup [4].
 Preliminary pore network modelling has already shown that the gas diffusivity increases with a decrease in water saturation of the clay material. However, the predicted trend also suggests that this increase is only marginal and should not normally exceed an order of magnitude. This is shown in the figure below.
 All the diffusion measurements are carried out on synthetic clay samples of dimensions 3.8 cm (diameter) and 2 cm (height). In Table 1, the Sw stands for water saturation. The compositions of the synthetic clay samples (60% clay, 20% sand, 20% silt) have been adjusted so as to most closely mimic the mineralogical compostion of Boom Clay. Synthetic samples are used instead of natural samples like Boom Clay because of their better physical response to drying than Boom Clay.
 The selected control variable to lower the saturation of the material is suction, which is imposed by changing the relative humidity of the environment using oversaturated salt solutions [4]. Suction can be kept constant given constant ambient temperature and pressure. Thus, the clay sample is conditioned at a certain suction, which in turn leads to desaturation in the range of 70-100% of total saturation depending on the level of suction. For this study, the saturation range has been fixed in between 70-100% of total saturation. This range has been chosen because of what is expected in a real life nuclear waste repository. The first measured diffusion coefficients in two partially saturated samples are shown in Table 1. The initial measurements of gas diffusion coefficients in unsaturated synthetic clays show that the increase in diffusivity of helium is higher than that of argon. Helium is the second lightest known element and has an extremely small molecular size and hence the impact of desaturation on helium diffusivity could be more pronounced than the same for heavier gases.

The mechanical behaviour of Boom Clay has been studied for many years in the context of geological disposal of radioactive waste in Belgium. The aim of this study is to investigate the anisotropic behaviour of Boom Clay in terms of compressibility and hydraulic conductivity. Oedometer tests (with effective vertical stress (\( \sigma_{v}^{{\prime }} \)) up to 32 MPa) were carried out on samples of various orientations: parallel, perpendicular and inclined 45° to the bedding plane. The compressibility index (C c ) and swelling index (C s ) were compared. Only a slight difference between these parameters was observed, suggesting that the anisotropic behaviour of Boom Clay cannot be revealed under the test conditions adopted. The hydraulic conductivity (k) was also determined by the Casagrande’s method for different values of vertical effective stress (\( \sigma_{v}^{{\prime }} \)). Unlike compressibility, the hydraulic conductivity, however, showed a clear anisotropic behaviour with k ver < k inc < k hor .

In June 2008 the UK government published a ‘White Paper’ as part of the “Managing Radioactive Waste Safety” (MRWS) programme to provide a framework for managing higher activity radioactive wastes in the long-term through geological disposal. The White Paper identifies that there are benefits to disposing all of the UK’s higher activity wastes (Low and Intermediate Level Waste (LLW and ILW), High Level Waste (HLW), Spent Fuel (SF), Uranium (U) and Plutonium (Pu)) at the same site, and this is currently the preferred option. It also notes that research will be required to support the detailed design and safety assessment in relation to any potentially detrimental interactions between the different modules. Different disposal system designs and associated Engineered Barrier Systems (EBS) will be required for these different waste types, i.e. ILW/LLW and HLW/SF. If declared as waste U would be disposed as ILW and Pu as HLW/SF. The Geological Disposal Facility (GDF) would therefore comprise two co-located modules (respectively for ILW/LLW and HLW/SF). A study has recently been undertaken by NDA RWMD to identify the key Thermo-Hydro-Mechanical-Chemical (THMC) interactions which might occur during both the operational and post-closure phases in order to assess the potential implications of co-location in a range of host rocks. This paper presents supporting modelling work used to help understand the potential interactions between the modules. A multi-phase flow and coupled gas generation model was used to help investigate the potential groundwater and gas fluxes between the modules, in particular considering the operational phase and resaturation behaviour of the different modules. These early phases are important because gas generation rates and hydraulic gradients will be at their maximum, and the pressure gradients associated with GDF operations will, at least initially, dominate over the background hydraulic gradient. The gas generation and multi-phase flow study considered a mudstone host rock in which gas pressurisation might significantly influence resaturation behaviour, or drive water from one module to the other. The results of the multiphase flow modelling show that although gas generation affects pressure recovery in the ILW/LLW module, the smaller size of the HLW/SF excavations compared with the ILW/LLW excavations, and the operational timings, mean that in general the groundwater pressure gradient in the GDF is from the HLW/SF module (higher pressure) to the ILW module (lower pressure). Transport of solutes from the HLW/SF module towards the ILW module is not expected to result in any deleterious interactions, indicating that hydraulic interactions during the resaturation period are unlikely to pose a fundamental barrier to co-location.

The release of radionuclides initially encapsulated in a slowly degrading solid waste form and contained in an eventually corroding canister defines the source term for numerical simulations for the assessment of a geologic repository for high-level radioactive waste. While the details of waste degradation, canister corrosion, and dissolution and mobilization of the radionuclides in pore water include complex chemical reaction and transport processes that are coupled to the thermal, hydrological, microbiological, and mechanical conditions in the repository, the source-term model suitable for use in a numerical performance assessment model should be a defensible abstraction of these mechanisms. We developed a radiological source-term model and implemented it into a non-isothermal flow and transport simulator. While the proposed source-term model is applicable to various waste forms, canister systems, and disposal concepts, we specifically considered radionuclide releases from vitrified high-level waste placed in a cylindrical canister disposed in a deep vertical borehole repository. In this model, waste degradation is a function of temperature, and it can be adjusted to evaluate the influence of and propagate uncertainties in pH, passivation reactions, and chemical conditions as well as geometrical factors. The time-dependent, congruent release of safety-relevant radionuclides present in the decaying inventory is then calculated. Finally, the radionuclides are mobilized by diffusive and advective transport according to the thermo-hydraulic conditions prevailing in the near field of the repository, from where they migrate through the geosphere to the accessible environment. We examine the influence of the source-term model’s parameters on performance assessment calculations through sensitivity and uncertainty propagation analyses, identifying influential factors and confirming the upper bound of their impact. These considerations align with the overarching goal of repository design, which is to demonstrate that engineered and natural barriers can collectively delay radionuclide migration for timescales far exceeding human planning, thereby providing multiple, redundant barriers against environmental contamination.

Deep low-permeability clay layers are considered as possible suitable environments for disposal of high-level radioactive waste. In Belgium, the Boom Clay is the reference host formation and the Ieper Clay an alternative host formation for research and safety and feasibility assessment of deep disposal of nuclear waste. In this study, two hydrogeological models are built to calculate the radionuclide fluxes that would migrate from a potential repository through these two clay formations. Transport parameters’ heterogeneity is incorporated in the models using stochastic sequential simulation of hydraulic conductivity, diffusion coefficient and diffusion accessible porosity, using primary information and several types of secondary information, i.e. resistivity, gamma ray and grain size. The calculated radionuclide fluxes in the two clay formations are compared. Results show that in the Ieper Clay larger differences between the fluxes through the lower and the upper clay boundary occur than in the Boom Clay, larger total output radionuclide amounts are calculated than in the Boom Clay, and a larger effect of parameter heterogeneity on the calculated fluxes is observed, compared to the Boom Clay.

The Nuclear Waste Management Organization (NWMO) is responsible for the design and implementation of Canada’s deep geological repository (DGR), which will be constructed ~ 500 m below ground surface to safely contain and isolate used nuclear fuel. Used fuel containers (UFC), designed by NWMO as part of multi-barrier system for DGR, comprises of an inner steel core with an outer copper layer that serves as a corrosion barrier. Surrounding the UFC, a highly compacted MX-80 bentonite (HCB) is used to suppress the transport of corrosive agents to the UFC and to limit the movement of radionuclides out of the DGR, in the highly unlikely event of a UFC failure. Under anaerobic conditions, sulfate-reducing bacteria at the interface of the host rock and bentonite may produce bisulfide (HS−) that can transport to the UFC surface and corrode the copper barrier. Therefore, it is crucial to understand HS− transport mechanisms through bentonite to assess the long-term DGR performance. Due to bentonite’s low permeability, HS− transport will be diffusion-driven; therefore, the apparent diffusion coefficient and retardation are critical parameters for the DGR performance assessment. This study aims to quantify HS− diffusion through bentonite using diffusion experiments under a range of anticipated DGR conditions (e.g., temperature, ionic concentration). This paper outlines the underpinning theory, experimental methodology developed to conduct experiments, and preliminary results. Altogether, this work bolsters confidence in the experimental design and methodology that will be used to determine necessary key parameters to model reactive transport of HS− through the DGR’s bentonite barrier.

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

This report evaluates the feasibility of high-level radioactive waste disposal in shale within the United States. The U.S. has many possible clay/shale/argillite basins with positive attributes for permanent disposal. Similar geologic formations have been extensively studied by international programs with largely positive results, over significant ranges of the most important material characteristics including permeability, rheology, and sorptive potential. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in shale media. We develop scoping performance analyses, based on the applicable features, events, and processes identified by international investigators, to support a generic conclusion regarding post-closure safety. Requisite assumptions for these analyses include waste characteristics, disposal concepts, and important properties of the geologic formation. We then apply lessons learned from Sandia experience on the Waste Isolation Pilot Project and the Yucca Mountain Project to develop a disposal strategy should a shale repository be considered as an alternative disposal pathway in the U.S. Disposal of high-level radioactive waste in suitable shale formations is attractive because the material is essentially impermeable and self-sealing, conditions are chemically reducing, and sorption tends to prevent radionuclide transport. Vertically and laterally extensive shale and clay formations exist in multiple locations in the contiguous 48 states. Thermal-hydrologic-mechanical calculations indicate that temperatures near emplaced waste packages can be maintained below boiling and will decay to within a few degrees of the ambient temperature within a few decades (or longer depending on the waste form). Construction effects, ventilation, and the thermal pulse will lead to clay dehydration and deformation, confined to an excavation disturbed zone within a few meters of the repository, that can be reasonably characterized. Within a few centuries after waste emplacement, overburden pressures will seal fractures, resaturate the dehydrated zones, and provide a repository setting that strongly limits radionuclide movement to diffusive transport. Coupled hydrogeochemical transport calculations indicate maximum extents of radionuclide transport on the order of tens to hundreds of meters, or less, in a million years. Under the conditions modeled, a shale repository could achieve total containment, with no releases to the environment in undisturbed scenarios. The performance analyses described here are based on the assumption that long-term standards for disposal in clay/shale would be identical in the key aspects, to those prescribed for existing repository programs such as Yucca Mountain. This generic repository evaluation for shale is the first developed in the United States. Previous repository considerations have emphasized salt formations and volcanic rock formations. Much of the experience gained from U.S. repository development, such as seal system design, coupled process simulation, and application of performance assessment methodology, is applied here to scoping analyses for a shale repository. A contemporary understanding of clay mineralogy and attendant chemical environments has allowed identification of the appropriate features, events, and processes to be incorporated into the analysis. Advanced multi-physics modeling provides key support for understanding the effects from coupled processes. The results of the assessment show that shale formations provide a technically advanced, scientifically sound disposal option for the U.S.

In the framework of studies for the disposal of radioactive waste, ONDRAF/NIRAS, the Belgian National Agency for Radioactive Waste and enriched Fissile Materials investigates the option of a geological repository. In a geological repository, the waste is sidelined within a deep and stable geological layer, named host rock, behind a whole series of artificial barriers. Together, natural and artificial barriers ensure the isolation of waste, their confinement and the delay and spread of release of radioactive substances. The interaction of the host rock and the concrete support of disposal galleries is investigated to analyse the long-term sustainability of the structure. The host rock considered in this study is a poorly indurated clay present in the northern part of the Belgium, the Boom Clay. The behaviour of the interface between the Boom Clay and the concrete of the galleries is studied experimentally from both mechanical and gas transfer perspectives. Samples have been prepared consisting of two half cylinders, one made of concrete and the other one of Boom Clay to obtain an interface. A dedicated device has been designed using an HOEK type cell, which allow applying confining pressure to simulate lithostatic loading, injecting gas to measure transfer properties and generating relative displacement of both sides of the interface. Mechanical behaviour of the interface under short loading has first been studied. One originality of this study consists in using gas injection and Poiseuille’s law to monitor and quantify the opening and closure of the interface during mechanical loading. A partial irreversible closure of the interface and effects of long term loading have been observed. By increasing the confinement, the opening of the interface (clay-concrete) gradually decreases (initially 38 μm for 1.2 MPa of confining pressure) until it reaches 5 μm for 4.5 MPa. There is moderate expected irreversibility during unloading (19 μm at 1.7 MPa). The objective is to identify the main parameters allowing predicting the properties of the interface, such as initial roughness of the interface, water content, applied stress and time. The second part of the work consists in studying transfer properties at the interface. The aim is to identify paths available for gas produced during waste storage.KeywordsInterfacesPoorly indurated clayConcreteTransfer propertiesShearing

The optimal diffusion experiment

26 - Corrosion issues of radioactive waste packages in geological disposal systems

The biogeochemistry of gas generation from low-level nuclear waste: Modelling after 18 years study under in situ conditions

The Waste Form Disposal Options Evaluation Report (SNL 2014) evaluated disposal of both Commercial Spent Nuclear Fuel (CSNF) and DOE-managed HLW and Spent Nuclear Fuel (DHLW and DSNF) in the variety of disposal concepts being evaluated within the Used Fuel Disposition Campaign. That work covered a comprehensive inventory and a wide range of disposal concepts. The primary goal of this work is to evaluate the information needs for analyzing disposal solely of a subset of those wastes in a Defense Repository (DRep; i.e., those wastes that are either defense related, or managed by DOE but are not commercial in origin). A potential DRep also appears to be safe in the range of geologic mined repository concepts, but may have different concepts and features because of the very different inventory of waste that would be included. The focus of this status report is to cover the progress made in FY16 toward: (1) developing a preliminary DRep included inventory for engineering/design analyses; (2) assessing the major differences of this included inventory relative to that in other analyzed repository systems and the potential impacts to disposal concepts; (3) designing and developing an on-line waste library (OWL) to manage the information of all those wastes and their waste forms (including CSNF if needed); and (4) constraining post-closure waste form degradation performance for safety assessments of a DRep. In addition, some continuing work is reported on identifying potential candidate waste types/forms to be added to the full list from SNL (2014 – see Table C-1) which also may be added to the OWL in the future. The status for each of these aspects is reported herein.