Disposal Of High-level Nuclear Waste Research Articles

Evaluation of Soil Water Characteristic Curves of Boron added Sand-bentonite Mixtures using the Evaporation Technique

Compacted bentonite or sand-bentonite mixtures are considered buffer/backfill materials in the engineering barriers of deep geological repositories for high-level nuclear waste (HLW) disposal in many countries. The design and long-term functionality of nuclear repositories have critical importance for environmental safety and public health. The initially unsaturated buffer material could become re-saturated long after following the sealing of the repository. Although the saturation degree of the buffer might decrease due to high temperatures and evaporation, it tends to increase with groundwater intrusion. Therefore, the soil water characteristic curves (SWCCs) for these unsaturated soils are a key factor in geotechnical engineering. Yet, the determination of SWCCs can be time-consuming and prone to inaccuracies. The HYPROP (Hydraulic Property Analyzer) evaporation technique is a preferred method for accurately determining water retention curves of soils. This reliable method was applied to estimate the water retention curves for sand-bentonite mixtures in the presence of boron minerals. Known for their minimal thermal expansion and commonly used in various industries, boron minerals may improve the thermal stability of sand-bentonite mixtures. The findings revealed that the boron addition increased the water retention capacity of the 10% bentonite mixtures but had a negligible impact on the 20% bentonite mixtures.

Harnessing Geothermal Energy Potential from High-Level Nuclear Waste Repositories

The disposal of high-level nuclear waste (HLW) has been one of the most challenging issues for nuclear energy utilization. In this study, we have explored the potential of extracting decay heat from HLW, taking advantage of recent advances in the technologies to utilize low-temperature geothermal resources for the co-generation of electricity and heat. Given that geothermal energy entails extracting heat from natural radioactivity within the Earth, we may consider that our approach is to augment it with an anthropogenic geothermal source. Our study—for the first time—introduces a conceptual model of a binary-cycle geothermal system powered by the heat produced by HLW. TOUGHREACT V3.32 software was used to model the heat transfer resulting from radioactive decay to the surrounding geological media. Our results demonstrate the feasibility of employing the organic Rankine cycle (ORC) to generate approximately 108 kWe per HLW canister 30 years after emplacement and a heat pump system to produce 81 kWth of high-potential heat per canister for HVAC purposes within the same timeframe. The proposed facility has the potential to produce carbon-free power while ensuring the safe disposal of radioactive waste and removing the bottleneck in the sustainable use of nuclear energy.

Operando scanning electron microscopy platform for in situ imaging of fluid evolution in nanoporous shale.

Fluid-solid interactions in nanoporous materials underlie processes fundamental to natural and engineered processes, including the thermochemical transformation of argillaceous materials during high-level nuclear waste disposal. Operando fluid-solid resolution at the nanoscale, however, is still not possible with existing optical and electron microscopy approaches that are constrained by the diffraction limit of light and by vacuum-fluid incompatibility, respectively. In this work, we develop an operando scanning electron microscopy (SEM) platform that enables the first direct in situ imaging of dynamic fluid-solid interactions in nanoporous materials with spatio-temporal-chemical resolutions of ∼2.5 nm per pixel and 10 fps, along with elemental distributions. Using this platform, we reveal necessary conditions for thermochemical pore and fracture generation in shales and measure their surface wetting characteristics that constrain the feasibility of high-level nuclear waste containment. Notably, we show that low heating-rate conditions typical of radioactive decay produce hydrocarbon liquids that wet fracture and pore surfaces in a self-sealing manner to impede aqueous radionuclide advection.

The failure of Euratom to harmonize the EU nuclear safety and environmental regulation. Can a recent ECJ judgment become a breakthrough, beneficial for EU’s quest for a sustainable, renewable-driver future?

The Treaty Establishing the European Atomic Energy Community (known as the Euratom Treaty), which was signed on 25 March 1957, contained commitments to promote nuclear energy technology and to integrate financial and technical efforts to develop that technology in the six signatory countries, in order to enhance wealth creation and the wellbeing of their citizens. The 66 year history of the Euratom Treaty undoubtedly shows that it didn’t fulfil its promises and worse, it became completely obsolete in the light of the European integration efforts contained in the Treaty on the Functioning of the European Union (TFEU). This article examines its fundamental shortcomings, whereby national interests in nuclear energy were at the forefront and whereby the harmonization of Nuclear Safety Standards and the protection of the environment and the health of the EU citizens became secondary, when confronted with industrial interests and the promotion of nuclear technologies (promotion which still continues for an industrial nuclear sector which is fully mature). This provoked a painful discord between nuclear counties and non-nuclear EU Member States, still the Achilles’ heel of Euratom as can be concluded from the discussion regarding nuclear energy as part of the EU Green Taxonomy. Euratom can be seen as the flagship of the pro-nuclear countries, albeit that the content of the Treaty is obsolete, as will be explained below from a legal angle. Because the EU Commission is the executive body of the Euratom Community, it is to blame for the failure to harmonize the safety standards as foreseen in the Euratom Treaty and for the unwillingness to integrate the EU legal environmental principles into nuclear law. It will be argued that the Nuclear Basic Safety Standards, as foreseen by the Euratom Treaty and the secondary Euratom Directives, are ‘empty shells’ and should no longer stand in isolation from the EU environmental acquis. The Justification and the ALARA principles should be informed by the Precautionary Principle and the Polluter Pays Principle. This entails that with the risks of global warming (such as heat waves, floodings, cooling water, wildfires, terrorism etc.) and the many caveats on nuclear decommissioning and disposal of high-level nuclear waste, the equation between the advantages of nuclear power (e.g., minor CO2 emissions) and its major risks should be reconsidered, not leaving this justification process to the nuclear industry, as in Belgium. The EU national governments should undertake this justification process at every stage of their political decisions regarding nuclear energy (e.g., life time extension), in full transparency and giving access to justice, as foreseen in the AARHUS Convention. A specialized EU Nuclear Safety Agency should receive a mandate by the EU Commission to harmonize such a justification process in the EU Member States and to harmonize the Nuclear Basic Safety Standards. The elementary decisions on the future of (nuclear) energy should be taken in a democratic way and not based on the promotional Euratom Treaty, in the light of the EU Green Deal and the energy transition towards renewable energy sources.

Swelling Behavior of a Clay-Pellets Mixture Intended for the Isolation of Spent Nuclear Fuel

Deep geological repository is the most favorable option for the safe disposal of high-level nuclear waste (HLW) and spent nuclear fuel (SNF). The design of such a facility relies on the multi-barrier system, encompassing a cylindrical metallic container (encapsulating the HLW/SNF), an engineered barrier system (EBS, around the metallic container); and the host-rock (or natural barrier). The most preferred buffer material to construct the EBS is bentonite clays. They are being considered as potential buffer materials because of, amongst other reasons, their high swelling capacity provides mechanical stability to the metallic canister containing the HLW/SNF; their good thermal conductivity assists to dissipate the heat released by the HLW/SNF; their low permeability delays the flow of water and gas through the system [1].
 Bentonites are highly swelling smectites clays that exhibit significant volume increase when wetted under free-swelling conditions, and develop high swelling pressure upon soaking when the volume change is restricted. It is envisaged that the EBS will be built using blocks of compacted bentonites (Figure 1(a)), or a combination of high-density clay-pellets mixtures and blocks of compacted bentonites (i.e., where the metallic container sits, Figure 1(b)). Clay-pellets are also considered as possible seal materials to fill gaps that will be present in this type of system, e.g., the gap between the EBS and the surrounding host-rock [2]. Therefore, a better understanding of the swelling capacity and behavior of the bentonite in these two forms (i.e., compacted-clay and clay-pellets mixtures) is critical for a proper design and evaluation of the long-term performance of geological repositories for HLW/SNF.
 Most of the research conducted in this area has been mainly focused on compacted bentonites looking, e.g., at the influence of several factors on their swelling pressure capacity, namely, type of clay minerals, initial dry density, water content (or degree of liquid saturation), and type of water [3]. More recently, investigations have been conducted to study the behavior of high-density pelletized clay mixtures [4, 5], however, relatively less attention has been placed in this type of material. Clay-pellets presents a number of advantage, e.g., they are very suitable for filling (small) technological voids (Figure 1(a)), there is no need for additional in-situ compaction when they are used as a buffer/backfill material (Figure 1(b)), and it is relatively easy to manufacture them.
 The granular clay-pellets samples were manufactured, and the constant volume cells were developed at Texas A&M University to investigate the swelling behavior of expansive clays. It has been shown in figure 2 that both materials (compacted clay and pelletized clay) developed similar maximum values of swelling pressures. However, the patterns associated with the swelling pressures evolutions are different, owing to the different pore structures associated with these two materials, which impact on the kinetic of clay hydration and swelling pressure evolutions. The aim of this research is to gain a better understanding of the swelling pressure behavior of clayed materials intended as potential barriers materials for the safe isolation of the HLW/SNF.

Numerical simulations of thermal fracturing in nuclear waste disposal

Regarding high-level nuclear waste (HLW) and spent fuel (SF) disposal, underground geological disposal is likely to be the most feasible and effective solution [1]. The coupled Thermo-Hydro-Mechanical (THM) behaviour of host rocks has been widely studied by laboratory tests, in-situ experiments and numerical analyses. In field experiments performed in underground laboratories, pore pressure generation and damage evolution have been observed. In this context, coupled THM numerical analyses can be an essential tool to better understand the coupled multiphysics behaviour of host rocks, buffer materials and canisters. In addition, large-scale field experiments provide a crucial opportunity to demonstrate understanding and to assess the predictive/modelling capabilities under realistic conditions.
 A number of hydromechanical (HM) numerical simulations of underground excavations have been carried out and the mechanisms of pore pressure and damage evolution have been studied [2]. Also, numerical analyses have been used to predict the observations of THM behaviour of in-situ heating tests, and there is some research on the damage evolution due to thermal loading [3], but few studies focus on the analysis of strain localization problems, in part due to specific numerical difficulties. However, thermal fracturing behaviour has indeed been observed in experiments and a better understanding of the phenomenon and of shear band evolution is crucial for a proper assessment of the stability of HLW/SF disposal.
 In this study, the coupled THM response of a generic HLW disposal facility under high-temperature has been analysed; in particular, the mechanism of thermal pressurization and the corresponding thermal fracturing behaviour have been examined. For this purpose, an advanced elastoplastic constitutive model with damage has been used to represent the mechanical behaviour of the host rock; it incorporates the following features: (1) a hyperbolic approximation to the Mohr-Coulomb yield surface to represent shear and tensile failure modes, (2) stiffness and strength anisotropy, (3) hardening and softening behaviour, (4) viscous effects and (5) a non-local integral type approach for the plasticity component of the model. The adopted overstress theory and the nonlocal formulation are capable of regularising the spatial distribution of plastic strains, eliminating mesh dependency and improving numerical efficiency. The coupling between accumulated plastic strain and hydraulic permeability has been considered in the analyses. Note that the term damage used in this study, refers to a state of the material, instead of the concept associated with damage mechanics theory; in addition, fracture represents localization phenomena in the damaged area.
 Numerical simulations have been performed using the finite element method software CODE_BRIGHT. Two-dimensional generic disposal configurations have been carried out for the two cases of unsupported and supported tunnels. The host rock is assumed to be the Callovo-Oxfordian claystone (COx). Three modelling stages have been considered: the excavation stage, the waiting stage and the heating stage, as shown in Figure 1 (a). In the heating stage, the thermal flow at the borehole wall is set as a constant power (200 W/m) and a no hydraulic flow boundary condition is prescribed. A maximum temperature of around 80 °C has been obtained on the drift wall at heating time t = 10 years.
 In addition, Figure 1 (b) shows the contours of plastic shear strains and liquid pressures at heating time t = 10 years. Near-field damage is caused by both the excavation of galleries and the effect of thermal pressurization. In the simulations presented, shear bands develop significantly during the heating stage. By comparing two cases in Figure 1 (b), it can be concluded that the installation of liners can efficiently limit the damage evolution and reduce the likelihood of thermal fracturing, significantly improving the stability of HLW disposal. Note that the scales of the two cases in Figure 1 (b) are different. Figure 1(c) shows that a lower pore pressure close to the gallery wall for the unsupported tunnel due to the higher level of damage in this case that results in a larger permeability of the rock in the damaged area.
 Apart from the analysis of the mechanism of THM behaviour in a high-temperature environment, numerical simulations have also been carried out of the large-scale PRACLAY Heater test (performed in the HADEs underground laboratory) and of the CRQ heating experiment (performed in the Meuse-Haute Marne underground laboratory). A good agreement has been observed between numerical predictions and experimental data, in terms of both temperature and pore pressure. These numerical validation exercises are, however, outside the scope of this abstract.

Stable microbial community in compacted bentonite after 5 years of exposure to natural granitic groundwater.

The Materials Corrosion Test (MaCoTe) at the Underground Research Laboratory in Grimsel, Switzerland, assesses the microbiology and corrosion behavior of engineered barrier components of a deep geological repository (DGR) for long-term disposal of high-level nuclear waste. Diversity and temporal changes of bentonite-associated microbial community profiles were assessed under DGR-like conditions for compacted Wyoming MX-80 bentonite (1.25 g/cm3 and 1.50 g/cm3 targeted dry densities) exposed to natural groundwater. Using culture-dependent and molecular techniques, samples taken from the outside layer of 5-year borehole modules revealed up to 66% and 23% of 16S rRNA gene sequences affiliated with Desulfosporosinus and Desulfovibrio, respectively. Putatively involved in sulfate reduction, these taxa were almost undetectable within the bentonite core. Instead, microbial profiles of the inner bentonite core were similar to uncompacted bentonite used to pack modules years earlier, and were consistent with a previously published 1-year time point, revealing no detectable microbial growth. Abundances of culturable aerobic and anaerobic heterotrophic bacteria in the uncompacted bentonite were relatively low, with less than 1,000 and 100 colony-forming units (CFUs) per gram dry weight, respectively. Nearly 5 years after emplacement, culturable heterotrophic bacterial CFUs and sulfate-reducing bacteria did not change significantly inside the bentonite core. Phospholipid fatty acid data indicated similar lipid abundance, and corresponding cell abundance estimates, for inner 5-year MaCoTe bentonite samples compared to those previously obtained for 1-year incubations. Collectively, our results provide complementary evidence for microbial stability inside highly compacted bentonite exposed to conditions that mimic engineered barrier components of a deep geological repository. IMPORTANCE The long-term safety of a deep geological repository for used nuclear fuel is dependent on the performance of the engineered and natural barriers. Microbial activity can produce chemical species that can influence the corrosion of the disposal containers for used nuclear fuel. Although previous studies have evaluated the microbiology of compacted bentonite clay within subsurface environments, these have been limited to relatively short incubations (i.e., 1 year). The current study provides a unique 5-year perspective that reinforces previous findings of growth inhibition for bentonite clay exposed to in situ subsurface conditions.

Laboratory Study of Effective Stress Coefficient for Saturated Claystone

Claystone is potentially the main rock formation for the deep geological disposal of high-level radioactive nuclear waste. A major factor that affects the deformation of the host medium is effective stress. Therefore, studying the effective stress principle of claystone is essential for a stability analysis of waste disposal facilities. Consolidated drained (CD) tests were carried out on claystone samples to study their effective stress principle in this paper. Firstly, two samples were saturated under a specified confining pressure and pore pressure for about one month. Secondly, the confining pressure and pore pressure were increased to a specified value simultaneously and then reverted to the previous stress state (the deformations of the samples were recorded during the whole process). Different incremental combinations of the confining pressure and pore pressure were tried at this step. Finally, the effective stress coefficients of the samples were obtained through a back analysis. Furthermore, some potential influencing factors (the neutral stress and loading rate) of the effective stress coefficient were also studied through additional tests. Some interesting results are worth mentioning: (1) the effective stress coefficient of claystone is close to one; (2) the neutral stress and loading rate may have little effect on the effective stress coefficient of claystone.

Stainless steel corrosion under anoxic, highly saline and elevated temperature conditions

Abstract. Several countries consider hosting a deep geological repository for high-level nuclear waste (HLW) in salt rock. In the unexpected case of solution access to the emplacement caverns during the long-term evolution of such a repository, metallic containers will be exposed to highly saline brines. In this study, corrosion experiments under conditions expected to be representative of HLW disposal in salt rock have been performed with stainless steel, a material typically used to construct containers containing vitrified HLW. Experiments were performed in closed vessels under anoxic and highly saline conditions at 90 ∘C. Polished stainless steel coupons were suspended in 5 mol L−1 NaCl or 3.4 mol L−1 MgCl2 at their natural pH. At the end of the experiments (up to 294 d), vessels were cooled down to room temperature, pH and Eh were measured in situ, the ultra-centrifuged contacting brines were analyzed by (HR)ICP-MS (high-resolution inductively coupled plasma mass spectrometry) and the formed corrosion products were identified. In all corrosion experiments, pHm and Eh hardly changed with exposure time, and the nature of the salt had no significant effect. The overall corrosion rates were low, and the quantities of dissolved metal ions revealed the formation of only sparingly soluble corrosion products. The formation of a passivation layer mostly made of Cr(III) (hydr)oxide was evidenced, but no localized surface attack could be detected. The coupon corroded in 5 mol L−1 NaCl was embedded in resin and cross-cut for further analyses, using synchrotron-based techniques with a small beam footprint (X-Ray fluorescence (μXRF), X-Ray diffraction (μXRD), and X-ray absorption near-edge structure (μXANES)). Element distribution maps revealed the presence of a very thin (<1 µm) layer of corrosion products, with regions enriched either in Cr or in Fe and Ni. X-ray diffractograms identified the presence of spinel-type compounds (Fe3O4 and NiFe2O4) and Cr2O3, which was corroborated by recording μXANES at the Cr, Fe and Ni K-absorption edges. The additional presence of layered double hydroxides in addition to NiO and Ni(OH)2 was also evidenced. Overall, results point to a corrosion layer having a duplex structure, with an inner layer mostly of chromium (hydr)oxides and an outer layer made of Fe- and Ni-based spinel compounds admixed with nickel (hydr)oxides.

Does the Fukushima nuclear accident still matter? Analysis of its mediated effects on five dimensions of nuclear power acceptance by using the parallel multiple mediator model

Although 10 years have passed since the Fukushima nuclear accident, it continues to greatly impact global energy policy and the acceptance of nuclear power. This study aims to empirically analyze the Fukushima accident's effects on five dimensions of nuclear power acceptance and the interactions of these effects with various determinants of nuclear power acceptance. We analyze survey data from 1020 Korean respondents and find that they remain highly worried about the Fukushima accident 10 years later. The Fukushima accident has negative impacts on the general acceptance of nuclear power, acceptance of nuclear power's contributions to addressing climate change, acceptance of high-level radioactive waste disposal sites, and acceptance of extensions of old nuclear power plants. However, the accident has a positive impact on the acceptance of denuclearization policies. Our parallel multiple mediator model shows that the Fukushima accident has only an indirect effect on general acceptance; this effect is negative and significant. The accident has a positive direct and indirect effect on acceptance of denuclearization policies but negative direct and indirect effects on the acceptability of high-level nuclear waste disposal and extensions of old nuclear power plants. Finally, we find a positive direct effect and a negative indirect effect of the accident on acceptance of nuclear power's contributions to addressing climate change. The latter is greater than the former, meaning that acceptance of nuclear power in the context of addressing climate change fell in response to the accident. In short, the Fukushima accident has different direct and indirect effects across different dimensions of nuclear power acceptance.

High-capacity isomorphic immobilization and leaching mechanism of Gd2Zr2O7-based ceramic waste form

Gd2Zr2O7(GZO)-based ceramic waste form has attracted broad interest due to its potential applications in the long-term disposal of high-level nuclear waste. However, its high-capacity isomorphic immobilization and leaching mechanism are not well understood yet. Herein, GZO-based waste form with a single pyrochlore phase and homogeneous element distribution, achieving an isomorphic waste load as high as 50 wt%, was fabricated successfully. Isomorphism failure of the waste form is due to the Ln6MoO12 (Ln is lanthanides) formation and direct precipitation of PdO from the oversaturated solid solution. Chemical durability tests showed that the 42-day leaching rates of nearly all elements in the waste form with the maximum waste load were only between 10−6 and 10−5 gm−2d−1. The leaching of 0.5SW·0.5GZO over a period of 42 days is primarily controlled by dissolution, leading to a gradual restructuring of the surface in two distinct stages: the formation of an amorphous passivation film and the agglomeration of amorphous precipitates. Notably, the formation of the amorphous passivation film stands as the primary factor influencing the leaching rate of the waste form.

Disposal of high-level radioactive waste in crystalline rock: On coupled processes and site development

Safe disposal of high-level radioactive nuclear waste (HLW) is crucial for human health and the environment, as well as for sustainable development. Deep geological disposal in sparsely fractured crystalline rock is considered one of the most favorable methods for final disposal of HLW. Extensive research has been conducted worldwide and many countries have initiated their own national development programs for deep geological disposal. Significant advancements of national programs for deep geological disposal of HLW in crystalline rock have been achieved in Sweden and Finland, which are currently under site development stage, focusing on detailed site characterization, repository construction, and post-closure safety analysis. Continued research and development remain important in the site development stage to ensure long-term safety of the HLW disposal repository. This work presents an overview and discussion of the progress as well as remaining open scientific issues and possibilities related to site development for safe disposal of HLW in crystalline rock. We emphasize that developing a comprehensive and convergent understanding of the coupled thermal, hydraulic, mechanical, chemical and biological (THMCB) processes in fractured crystalline rock remains the most important yet challenging topic for future studies towards safe disposal of HLW in crystalline rock. Advancements in laboratory facilities/techniques and computational models, as well as available comprehensive field data from site developments, provide new opportunities to enhance our understanding of the coupled processes and thereby repository design for safe geological disposal of HLW in crystalline rock.

Tuning the Alkyl Chain of Nitrilotriacetamide for Selectively Extracting Trivalent Am over Eu Ions.

The successful management and safe disposal of high-level nuclear waste necessitate the efficient separation of actinides (An) from lanthanides (Ln), which has emerged as a crucial prerequisite. Mixed donor ligands incorporating both soft and hard donor atoms have garnered interest in the field of An/Ln separation and purification. One such example is nitrilotriacetamide (NTAamide) derivatives, which have demonstrated selectivity in extracting minor actinide Am(III) ions over Eu(III) ions. Nevertheless, the Am/Eu complexation behavior and selectivity remain underexplored. In the work, a comprehensive and systematic investigation has been conducted for [M(RL)(NO3)3] complexes (M = Am and Eu) utilizing relativistic density functional theory. The NTAamide ligand (RL) is substituted with various alkyl groups, namely, methyl, ethyl, propyl, n-butyl, n-pentyl, n-hexyl, n-heptyl, and n-octyl. Thermodynamic calculations show that the alkyl chain length in NTAamide is capable of tuning the separation selectivity of Am and Eu. Moreover, the differences in calculated free energies between Am and Eu complexes are more negative for R = Bu-Oct than Me-Pr. This indicates that elongation of the alkyl chain can increase the efficiency of selective separation of Am(III) from Eu(III). Based on the quantum theory of atoms in molecules and charge decomposition analyses, it has been observed that the strength of Am-RL bonds is higher than that of Eu-RL bonds. This disparity is attributed to a greater degree of covalency in Am-RL bonds and a higher level of charge transfer from ligands to Am within complexes containing these bonds. Energies of occupied orbitals with the central N character are recognized overall lower for [Am(OctL)(NO3)3] than for [Eu(OctL)(NO3)3], indicative of stronger complexation stability of the former. These results offer valuable insights into the separation mechanism of NTAamide ligands, which can help guide the development of more powerful agents for An/Ln separation in future applications.

Diffusion of radioisotopes in Fe-montmorillonite relevant to geological disposal of HLW

Bentonite clay is proposed as backfill or buffer materials in the engineered barrier system for high-level nuclear waste (HLW) disposal in a deep geological repository (DGR). The major clay mineral of bentonite, montmorillonite (Mt), is expected to undergo alteration owing to its interaction with corrosion products from overpack/canister over a long period of time. As a typical and simple altered clay mineral, we used Fe(II)−Mt and Fe(III)−Mt for preliminary investigation on the diffusion behavior of various radioisotopes relevant to HLW. Instantaneous planar source method was used to determine the apparent diffusion coefficient (Da) using compacted (ρd = 1.6 g/cm3) and water saturated Fe−Mt. In Fe(III)−Mt, the Da values in m2/s for 137Cs+ = (3.9 ± 0.2) × 10−12, 133Ba2+ = (1.1 ± 0.6) × 10−12, 152+154Eu3+ = (2.2 ± 0.5) × 10−12 and 131I− = (5.5 ± 0.1) × 10−12. In Fe(II)−Mt, the Da values in m2/s for 22Na+ = (2.0 ± 1.2) × 10−11, 137Cs+ = (3.0 ± 0.4) × 10−12, 133Ba2+ = (1.4 ± 0.4) × 10−11, 152+154Eu3+ = (4.0 ± 0.7) × 10−12 and 131I− = (4.8 ± 1.0) × 10−12. In general, similar Da values are obtained for Fe(II)−Mt and Fe(III)−Mt. This is the first extensive study to report the diffusion behavior of radioisotopes in Fe−Mt.

Saponite as a New Generation Engineered Buffer Material for High Level Nuclear Waste Disposal: Its Chemical Stability in High-pH and High-Temperature Environments

ABSTRACT Saponite, a tri-octahedral smectite with an ideal formula of Mg3(Si,Al)4O10(OH)2·4H2O for an Mg-endmember (saponite-15Å), has similar swelling properties as montmorillonite and therefore can be considered as a potential buffer material for nuclear waste disposal. In this experimental study, we investigated the chemical stability of this material under alkaline conditions (pH ∼ 12) at high temperatures (up to 150 °C) and saturated vapor pressures over an experimental duration of about three weeks. Alkaline conditions in a geologic repository can be induced by the interaction of groundwater with borosilicate waste forms or by the leaching of alkaline porewaters from cementitious materials introduced as structural components, such as seals, liners, and plugs. The interactions of saponite with portlandite [Ca(OH)2]-saturated solutions at 60, 125, and 150 °C were experimentally simulated. The results indicated that saponite is chemically more stable in these environments than montmorillonite and remained unchanged after the interactions. The stability is attributed to its low solubility, with dissolved-magnesium concentrations in the portlandite-saturated solutions estimated to be less than 4×10−7 mol/kg or 2×10−7 mol/kg at 60 or 150 °C, respectively. For comparison, we also investigated the interactions of montmorillonite (SWy-2) with a portlandite-saturated solution at 150 °C. Our results indicate that montmorillonite is not stable, leading to the formation of zeolite (phillipsite) after 9 days of reaction. This study thus demonstrates the feasibility of using saponite as a potential buffer material (as an alternative to montmorillonite currently being considered in many disposal concepts) in an alkaline and elevated-temperature disposal environment.

Machine learning-based models for predicting gas breakthrough pressure of porous media with low/ultra-low permeability.

Gas breakthrough pressure is a significant parameter for the gas exploration and safety evaluation of engineering barrier systems in the carbon dioxide storage, remediation of contaminated sites, and deep geological repository for disposal of high-level nuclear waste, etc. Test for determining gas breakthrough pressure is very difficult and time-consuming, due to the low/ultra-low conductivity of the specimen. It is also difficult to get a comprehensive and high-precision model based on limited results obtained through individual experiments, as the measurements of gas breakthrough pressure were influenced by many factors. In this study, a collected database was built that covered a lot of former test data, and then, two models were developed by the random forest (RF) algorithm and multiexpression programming (MEP) method. The MEP model constructed with explicit expressions for the gas breakthrough pressure overcame the drawbacks of common "black box" models. Meanwhile, five significant indicators were selected from ten common features using the permutation importance algorithm. The RF model was interpreted by the Shapley value and the PDP/ICE plots, while the MEP model was analyzed through the proposed explicit expression, showing strong consistence with that in former studies. Finally, robustness analysis was conducted, and stability of the proposed two models was verified.

Controlled disorder for the Yb2Ti2-xO7-2x (x = 0 to 1) series and corresponding radiation tolerance

The use of ceramics as radionuclide containment matrices for the safe long-term storage and disposal of high-level nuclear waste is studied with emphasis on chemistry flexibility, structural stability, and radiation tolerance. Here we investigate the flexibility of a particular series of compounds Yb2Ti2-xO7-2x (x = 0, 0.2, 0.52, 0.8, and 1) to form a solid solution between the end members Yb2Ti2O7 of pyrochlore structure through to Yb2TiO5 of fluorite structure. Whilst it is shown that cation disorder and corresponding transition from the ordered pyrochlore to disordered fluorite is facilitated via increasing the ytterbium to titanium ratio, it is also shown that sintering conditions may be controlled in such a way as to encourage the growth of the pyrochlore structure even in Yb2TiO5. The radiation response of these materials was tested in situ via 1 MeV krypton irradiation coupled with transmission electron microscopy characterisation. The critical fluence of ions required to transition the crystalline phase to amorphous was found to increase with increasing disorder from Yb2Ti2O7 to Yb2TiO5. However, of the two forms (pyrochlore and fluorite) of Yb2TiO5 studied via ion irradiation that with the greater pyrochlore structure and so more ordered also had the lower critical temperature, 434 K, for maintaining crystallinity during irradiation.

Enhanced Crevice Corrosion of Stainless Steel 316 By Degradation of Cr-Containing Hollandite Crevice Former

The disposal of high-level nuclear waste (HLW) is an important societal problem. It is also an extremely challenging corrosion research topic due to the highly complicated waste chemistry, long lasting radiation effect, and inevitable exposure to aqueous environments for hundreds of thousands of years. To safely accommodate some critical radionuclides such as volatile 129I and strongly heat generating 137Cs, numerous crystalline ceramic waste forms have been created and studied. Among the promising nuclear waste forms, hollandite was developed to isolate Cs waste and the corresponding decay product Ba due to the open tunnel structure capable of accommodating mono- or divalent cations. In this study, the synergistic corrosion interactions between a Cr-containing hollandite (Ba1.15Cr2.3Ti5.7O16) and stainless steel (SS) 316 is explored. This is relevant to the permanent disposal of HLW involving the encasement of glass or crystalline ceramic waste forms containing immobilized radionuclides in a metallic canister made from corrosion resistant alloys such as SS. The experiments performed in this study simulate the potential corrosion interactions occurring at the interface of the SS and ceramic. After corroding the SS316 and Cr-containing hollandite in proximity in 0.6 M NaCl at 90 oC for 28 days, severe crevice corrosion was identified on the surface of the SS, as evidenced by the presence of large pits and crevice damage with diameters of hundreds of microns. Similarly, localized damage was also found at matching sites on the Cr-hollandite surface, indicating interactions between the two materials. The synergistic corrosion interaction was likely due to the continuous release of Cr3+ cations from both materials, including the passive dissolution of the SS and the heterogenous degradation of the Cr-hollandite. The extra Cr3+ cations originated from the hollandite reduces the incubation time required to reach the critical crevice solution condition, thereby accelerating the breakdown of the passive film of SS and the subsequent onset of active dissolution. An adapted crevice corrosion model is applied to quantitively explain the accelerated corrosion of SS observed in this study.

Effect of thermal ageing on physical properties of MX80 bentonite under high-temperature conditions

In a deep geological repository (DGR) system, the buffer layer is indispensable to ensuring the safe disposal of high-level radioactive nuclear waste (HLW). Because the heat generated by the decay of the spent nuclear fuel in a canister is released to the surrounding buffer layers, the bentonite buffer material experiences long-term high-temperature conditions. Therefore, the variations in physical properties of bentonite buffer material under high-temperature conditions are one of the important parameters in the DGRs for HLW. A series of tests on the specific gravity, specific surface area (SSA), Atterberg's limits, and free swelling ratio of MX80 bentonite after heating for different times at a high temperature of 200 °C were conducted, to investigate the influence of thermal ageing time on its physical properties. Then, the microscopic investigations, including X-ray diffraction (XRD), thermogravimetric analysis (TGA), and scanning electron microscopy (SEM), were conducted to explain the aforementioned variations from the microscopic point of view. Based on the crystal layer structure characteristics of montmorillonite, the quantitative relationships between the physical properties of bentonite and its mineral composition or bound water content were established to further explain the mechanism by which thermal ageing affects the physical properties of MX80 bentonite. The results indicate that with increasing heating time, the specific gravity, SSA, liquid limit, plastic limit, plastic index and free swelling ratio decrease sharply by 2.5%, 4.5%, 3.1%, 5.5%, 2.8% and 30.2%, respectively, within 15–30 days of heating. After 30 days of heating, their variations are negligible. Under high-temperature conditions, the transformation of mineral composition, desorption of bound water, and changes in the micro-morphology are the fundamental reasons for the variations in the physical properties of bentonite, and they influence and interact with each other.

Evaluation of hydro-mechano-chemical behaviour of bentonite-sand mixtures

Bentonite-sand mixtures are widely used in engineering barrier of deep geological disposal of high-level radioactive nuclear waste and anti-seepage barrier of civil geotechnical engineering. Under the action of groundwater solution infiltration and external stress, the hydro-mechanical (HM) behaviour of bentonite-sand mixtures, i.e. the swelling characteristics and permeability, will change. Once the anti-seepage and filtration effect is weakened or lost, the pollutants will spread to the biosphere. Therefore, it is necessary to study the swelling characteristics and permeability of bentonite-sand mixtures under coupled mechano-chemical (MC) effect and to establish corresponding prediction model. For this reason, swelling tests under salt solution with different concentrations are conducted on pure bentonite and its mixtures with 30%, 70% and 90% sand contents, the compression tests are carried out on saturated samples, and the saturated permeability coefficient k of the sample under each load is calculated by Terzaghi's one-dimensional consolidation theory. The concepts of true effective stress p e , montmorillonite void ratio e m and critical sand content α s are introduced to determine the e m - p e relationship and finally the k - e m relationship of bentonite-sand mixtures. It is found that when the sand content α ≤ α s , the e m - p e relationship of the mixture is linear and independent of the salt solution concentration, and when α > α s , the e m - p e relationship of bentonite-sand mixture is bi-linear with the true effective deviatoric stress p esα as the intersection. In addition, the e m - k relationship also shows the linear trend when α ≤ α s , and the slope of the line increases with the increase of the salt solution concentration. When α > α s , the k - e m relationship will deviate from the linear relationship. Moreover, the larger the sand content is, the farther the deviation is. On the basis of summing the regularity, a model for predicting the HM behaviour of bentonite-sand mixture under the coupled MC effect is proposed. By comparing the swelling and permeability test results with model prediction results of different types of bentonite and its sand mixtures, the predictive model is verified. The study on the HM behaviour of bentonite-sand mixtures under salt solution infiltration and the model establishment can provide experimental and theoretical basis for the design and construction of anti-seepage engineering by bentonite-sand mixtures.