Modeling of Thermo-Hydro-Mechanical Processes in a Bentonite Buffer Under High Temperature

The corrosion of waste canisters in the deep geological disposal facilities (GDFs) for high‐level radioactive waste (HLRW) can generate gas, which escapes from the engineered barrier system through the interfaces between the bentonite buffer blocks and the host rock and those between the bentonite blocks. In this study, a series of water infiltration and gas breakthrough experiments were conducted on granite and on granite–bentonite specimens with smooth and grooved interfaces. On this basis, this study presents new insights and a quantitative assessment of the impact of the interface between clay and host rock on gas transport. As the results show, the water permeability values from water infiltration tests on granite and granite–bentonite samples (10−19–10−20 m2) are found to be slightly higher than that of bentonite. The gas permeability of the mock‐up samples with smooth interfaces is one order of magnitude larger than that of the mock‐up with grooved interfaces. The gas results of breakthrough pressures for the granite and the granite–bentonite mock‐up samples are significantly lower than that of bentonite. The results highlight the potential existence of preferential gas migration channels between the rock and bentonite buffer that require further considerations in safety assessment.

Reactive transport model of the long-term geochemical evolution in a HLW repository in granite at the disposal cell scale: Variants, sensitivities, and model simplifications

Salt formations have long been recognized as a highly favorable host rock for the final disposal of high-level radioactive waste (HLW) in deep geological repositories. Their unique properties, including exceptional impermeability, self-healing capabilities, and thermal conductivity, make them a reliable natural barrier for the deep disposal of radioactive waste. This report focuses on the development and application of a methodology for assessing the integrity and per formance of the Engineered Barrier System (EBS) within salt-based repositories, a critical component of the multi-barrier system ensuring safe radioactive waste disposal.

In low/intermediate-level waste (L/ILW) repositories, anaerobic corrosion of metals and degradation of organic materials produce hydrogen, methane, and carbon dioxide. Gas migration in a L/ILW repository is one of the processes evaluated in the safety assessment of deep geological disposal in low-permeability formations, in particular with respect to the development of gas pressures in the repository caverns which could negatively affect the host rock or the engineered barrier system (EBS). In order to restrict build-up of gas overpressures in the emplacement caverns, Nagra (National Cooperative for the Disposal of Radioactive Waste, Switzerland) has proposed design options aimed at increasing the gas transport capacity of the backfilled underground structures, compromising neither the low hydraulic conductivity nor the radionuclide retention capacity of the EBS (Nagra, 2008). They involve specially designed backfill and sealing materials such as high porosity mortars as backfill materials for the emplacement caverns and sand/bentonite (S/B) mixtures with a bentonite content of 20% to 30% for the seals themselves and for backfilling other underground structures. These increased gas permeability materials can supplement the gas flow that is expected to occur through the excavation damaged zone (EDZ) and avoid the creation of overpressures. Preliminary experimental studies have confirmed the gas transport capacity of the S/B mixtures and demonstrated the ability to design mixtures with specific target permeabilities for water and gas flow (Nagra, 2008). Two-phase flow modelling studies have shown that the gas transport capacity of seals is largely dependent on their permeability and length. More detailed models of sealing elements show a rather complex history of seal saturation during the early saturation phase and the later gas escape phase (Gaus et al., 2010). Note, however, that current modelling approaches are based on parameters and conceptual understanding of small-scale laboratory experiments. Two large(r) scale experiments which aim at validating and, if necessary, improving current conceptual models for the resaturation and gas invasion processes into S/B seals and the determination of up-scaled gas / water permeabilities of S/B seals (i.e. two-phase flow parameters for large-scale models) have been initiated and will be highlighted in the paper. The first one, a mock-up experiment, was set up in 2010 as part of the EU 7th FP project FORGE, aiming at demonstrating seal performance on an intermediate (decimetre scale). The second one is a large-scale experiment (metre-scale), the Gas-Permeable Seal Test (GAST), which was also initiated in 2010 at the Grimsel Test Site (GTS). For GAST, a seal will be emplaced at the GTS to demonstrate the effective functioning of gas-permeable seals on a realistic scale and with realistic boundary conditions (‘proof of concept’).

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.

On the impact of temperatures up to 200 °C in clay repositories with bentonite engineer barrier systems: A study with coupled thermal, hydrological, chemical, and mechanical modeling

Geological repositories for disposal of high-level nuclear waste generally rely on a multibarrier system to isolate radioactive waste from the biosphere. The multi-barrier system typically consists of the natural barrier system (NBS), which includes the repository host rock and its surrounding subsurface environment, and the engineered barrier system (EBS). The EBS represents the man-made, engineered materials placed within a repository, including the waste form, waste canisters, buffer materials, backfill, and seals (OECD, 2003). The EBS plays a significant role in the containment and long-term retardation of radionuclide release.

Coupled THMC models for bentonite in an argillite repository for nuclear waste: Illitization and its effect on swelling stress under high temperature

THM coupling sensitivity analysis in geological nuclear waste storage

An engineered barrier system (EBS) for the geological disposal of high-level radioactive waste (HLW) consists of a disposal canister packed with spent fuel, buffer material, backfill material, and gap-filling material. The buffer material fills the space between the canister and the near-field rock, thus serving to restrain the release of radionuclides and protect the canister from groundwater penetration. Furthermore, as significant amounts of heat energy are released from the canister to the surrounding rock, the thermal conductivity of the buffer plays an important role in maintaining the safety of the entire disposal system. Therefore, given the high levels of heat released from disposal canisters, this study measured the thermal conductivities of compacted bentonite buffers from Gyeongju under temperature variations ranging 25 to 80~90℃. There was a 5~20% increase in thermal conductivity as the temperature increased, and the temperature effect increased as the degree of saturation increased.

The Deep Geological Repository (DGR) design, the internationally safest option for the long-term disposal of high-level radioactive waste (HLW), features metal canisters encased in compacted bentonite clay and embedded deep within a host rock. Despite presenting a hostile environment for microorganisms, DGRs scenarios with favorable microbial-activity conditions must be considered for the safety assessment of this disposal. This study investigated the impact of Se(IV), as a natural analogue of 79Se present in the HLW, in anoxic microcosms of bentonite slurry spiked with a bacterial consortium and amended with lactate, acetate, and sulfate as electron donors/acceptor. The addition of the bacterial consortium promoted the rate of Se(IV) reduction to Se(0), while the tyndallization (heat-shock) of bentonite slowed this process. Se(IV) reduced the relative abundance of most genera of sulfate-reducing bacteria (SRB), while stimulating the abundance of Se-tolerant bacteria, which played an important role in Se(IV) reduction. Moreover, it was observed that lactate was the preferred electron donor, linking to the production and subsequent consumption of acetate. X-ray absorption spectroscopy (XAS) and high-resolution transmission electron microscopy (HRTEM) revealed the reduction of Se(IV) forming amorphous Se(0) nanospheres. In addition, HRTEM showed that the biogenic Se(0) undergo a biotransformation to more stable crystalline forms, contributing to the immobilization of Se in the case of HLW release. Additionally, the sulfide generated by the activity of SRB reacted with Cu producing corrosion products (CuxS) on the surface of the copper material.

Uncertainty and sensitivity analysis for the early failure scenario classes in the 2008 performance assessment for the proposed high-level radioactive waste repository at Yucca Mountain, Nevada

Specific heat capacity model for compacted bentonite buffer materials

ABSTRACT: The deep geological repository is one of the most preferred disposal methods in many countries for high-level radioactive waste materials and spent fuel after nuclear energy production. The engineered barrier system (EBS) in the repository holds the nuclear waste-containing canister in place. However, it is known that the EBS can be readily dried near the canister due to the heat emitted from the spent fuel inside the canister, which could induce the generation of cracks in the EBS. Inorganic microfiber can be an alternative to mitigate such cracks, and thus, enhance the performance of the engineered buffer material. This study investigated the coupled effect of glass microfiber reinforcement on the swelling and compressibility characteristics of sodium bentonite at both ambient (22°C) and elevated (50°C) temperatures. By using the one-dimensional consolidation test setup with temperature controls, the swelling pressure test was performed to investigate the swelling pressure, compression, and swelling indices for 0.0% and 3.0% glass fiber contents by weight. A unique coupled relationship between the fiber inclusion and temperature elevation for the bentonite can better explain the engineering performance of the deep geological repository to safely store the high-level radioactive waste materials and spent fuel. 1. INTRODUCTION Nuclear energy is one of the main sources of energy utilized across the globe. According to the U.S. Energy Information Administration (2021), the United States of America generated 809 BkWh of total annual nuclear electricity last 2019. This amount translates to 30.5% of the total world nuclear electricity generation. At present, the United States has 94 nuclear reactors in operation, in which 30 of its states house many commercial nuclear power plants (U.S. Energy Information Administration, 2021). According to the U.S. Environmental Protection Agency (2022), fossil fuel accounts for 65% of global greenhouse gas emissions. On the other hand, nuclear power currently represents more than half of all low-carbon electricity generation in the U.S. and 30% of the world’s low-carbon electricity (Hsu, 2021). With that, nuclear energy is gaining traction due to its non-production of greenhouse gas emissions during operation (World Nuclear Association, 2022). However, given the national and global reliance on this energy, its production entails a huge amount of nuclear waste generated afterward.

The Federal Company for Radioactive Waste Disposal (BGE mbH) is responsible for the execution of the German site selection procedure for high-level radioactive waste. The aim is to identify a repository site that ensures the best possible safety for the disposal of high-level radioactive waste for at least one million years. Three potential host rocks are considered for disposal in Germany: rock salt, claystone and crystalline rock. The German site selection procedure consists of three consecutive phases with a continuously increasing level of detail. The first step of phase I served to determine sub-areas based on different geoscientific criteria. In this process 90 suitable areas were identified. The second step of phase I comprises representative preliminary safety assessments to further narrow down the sub-areas (see Figure 1).Figure 1: German Site Selection ProcedureFor these assessments, a preliminary concept and dimensioning of the deep geological repository is required, among other aspects. The primary input data for the preliminary design of the repository are the respective host rock properties, the inventory data of the high-level radioactive waste as well as a preliminary safety concept for the disposal system. Since there are only limited area-specific data available as no exploration is part of this early stage of the selection procedure, a two-stage method has been developed for the determination of the required area of the potential repository, consisting of a host rock specific and a site specific part. The objective of the first, host rock specific stage is to perform an analysis irrespective of the prevailing geological conditions at the considered locations. The aim is to obtain the areal extend of a potential repository as a function of the depth, the initial temperature in the host rock and mechanical properties of the host rock. Another important topic is the assessment of favorable depth ranges of the repository. Within the second stage, the function is used with available site-specific data to obtain the possible size of the repository at this specific site.This contribution will provide an overview of the first stage within the aforementioned methodology for the development of the preliminary design of the repository as part of the representative preliminary safety assessments.