High-level Nuclear Waste Research Articles

Surface charge characteristics of Gaomiaozi bentonite in high-level nuclear waste repositories

Gaomiaozi (GMZ) bentonite is an excellent backfill material for high-level nuclear waste repositories. The bentonite minerals possess a laminated structure and carry a negative charge. These surface charge characteristics play a crucial role in the development of the electric double layer at the liquid-solid interface and the adsorption of nuclear elements during the hydration process. Following the purification of GMZ bentonite, the surface electric potential of bentonite mineral colloids in suspension and compacted bentonite blocks was measured using atomic force microscopy (AFM). Additionally, the influence of various factors on its surface charge characteristics was examined, including the types of interlayer cations, ambient temperature, number of layers, and dry density. The findings reveal that at 20 °C, the surface electric potential of Na-bentonite quasicrystals averages around −225.2mV, while that of Ca-bentonite quasicrystals averages approximately −155.3mV. Additionally, the surface electric potential increases logarithmically with the number of layers. In the compacted state, its surface potential would decrease. An increase in the dry density of compacted bentonite results in a decrease in the surface electric potential. High ambient temperatures can increase the surface electric potential. The research findings offer valuable data and a theoretical foundation for predicting colloid formation, nuclear element adsorption, and swelling pressure development in bentonites used in high-level nuclear waste repositories.

Assessment of thermal-hydraulic-mechanical-chemical (THMC) performances of Ca-type bentonite-graphite mixtures as buffer materials for a high-level radioactive waste repository

The bentonite buffer material is the most important element of the engineered barrier system (EBS) used to dispose of high-level radioactive waste. Therefore, there are functional criteria or performance targets for the bentonite buffer material to maintain the long-term performance of the EBS for the entire disposal system. Thermal conductivity is a key parameter in the design of an EBS as it has the greatest influence on the buffer temperature. As the peak temperature of the buffer should not exceed the thermal design criterion, it is necessary to increase the thermal conductivity of the bentonite buffer material. Therefore, in this study, a small portion of graphite was added to Ca-type bentonite and thermal, hydraulic, and mechanical properties were measured separately, along with overall THMC (thermal-hydraulic-mechanical-chemical) properties. Additionally, while thermal conductivity was examined for varying graphite contents, other properties such as hydraulic conductivity, swelling pressure, and unconfined compressive strength were tested with a 3 % graphite addition. In particular, the thermal conductivity was measured for 3 % graphite addition to the pure bentonite under the different dry densities and saturation conditions and a regression analysis was performed to predict thermal conductivity based on the dry density and degree of saturation values. Furthermore, the saturated hydraulic conductivity, swelling pressure, unconfined compressive strength, and sorption efficiency of cesium (Cs) ions in bentonite and bentonite-graphite mixtures were measured. Moreover, prototype buffer blocks for bentonite-graphite mixtures were manufactured to evaluate viability of production. The addition of 3 % graphite to pure bentonite satisfied the functional criteria of saturated hydraulic conductivity and swelling pressure when the dry density was 1.63 g/cm3 or higher, and the thermal conductivity of the bentonite-3 % graphite mixtures was 10–30 % higher than that of pure bentonite.

Development of integrated FEPs for safety case scenario development

ABSTRACT To establish a deep disposal repository for high-level radioactive waste, it is crucial to develop a safety case, which is comprehensive evidence and arguments supporting the safety of a nuclear facility or activity. The development of safety cases is achieved by creating scenarios based on a comprehensive understanding of the disposal system. Scenarios are developed using Features, Events, and Processes (FEP) and must follow the rules of comprehensiveness, systematic approach, transparency, and traceability (rules for the scenario). Also, scenarios have to be developed in a way that the stakeholders understand the whole development process clearly. In this study, an intermediate step was introduced for the high-level radioactive waste management program to strengthen the rules for the scenario. An intermediate step was established through the integration of FEP. For integrating FEPs, we classified 268 NEA International FEPs into six classifications concerning the relationships between each FEP to scenario and safety objectives. Finally, we selected Repository system evolution FEPs as the target for integration because they can result in significant changes in system performance.

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

The assessment of the long-term performance of the engineered barrier systems of high-level radioactive waste (HLW) repositories requires the use of reactive transport models. Montenegro et al. (2023) presented a non-isothermal reactive transport model of the long-term geochemical evolution of a HLW disposal cell in a granitic host rock corresponding to a generic reference concept. The model accounted for the vitrified waste, the carbon-steel canister, the bentonite buffer and the reference granitic rock. Here we extend their model by considering model variants (V), sensitivity cases (SC) and model abstractions (MA). Variants V1, V2 and V3 consist of considering MX-80 bentonite instead of FEBEX bentonite (V1), a larger groundwater flux through the granite (V2) and the Czech reference crystalline rock as a host rock (V3). Cases SC1 and SC2 consider a decrease of the silica concentration threshold value in the glass dissolution rate (SC1) and an earlier canister failure time (SC2), respectively. Runs MA1 to MA4 consider smectite as an unreactive mineral phase (MA1), the porosity feedback effect on chemical and transport parameters (MA2), a time-varying corrosion rate (MA3), and a coarser finite element grid (MA4), respectively. Model results of V1 show a larger pH, a smaller precipitation of magnetite, siderite and greenalite and a slightly smaller dissolution of ISG and smectite than the base run of Montenegro et al. (2023). Model predictions are very sensitive to the increase in the groundwater flow through the granitic host rock (V2). However, predictions are not sensitive to the chemical composition of the granite porewater (V3). The decrease in the silica saturation threshold from 1·10−3 to 5·10−4 mol/L in SC1 leads to a significant decrease in glass dissolution. Glass dissolution after 50,000 years in SC2 (earlier canister failure) is much larger than that of the base run. Model results are not sensitive to considering smectite as an unreactive mineral phase (MA1). However, model results are very sensitive to the porosity feedback effect (MA2). A 60% volume fraction of Fe(s) remains uncorroded after 50,000 years when a variable corrosion rate is considered in MA3. In this case the precipitation of corrosion products is much smaller than that of the base run. The general patterns of the numerical results in MA4 (coarser grid) are similar to those of the base case.

Gas breakthrough in compacted Gaomiaozi bentonite under rigid boundary conditions

Predicting the gas breakthrough pressure of saturated compacted bentonite is crucial for ensuring the long-term safe operation of deep geological repositories for the disposal of high-level radioactive nuclear wastes. In this work, the swelling pressure, water injection, gas injection and mercury intrusion porosimetry (MIP) tests on saturated compacted Gaomiaozi (GMZ) bentonite specimens with a dry density of 1.3 Mg/m³, 1.4 Mg/m³, 1.5 Mg/m³, 1.6 Mg/m³ and 1.7 Mg/m³ were conducted. Subsequently, the relationships between the swelling pressure and average inter-particle distance, as well as between the gas entry pressure and the maximum effective pore size were analyzed and established. Considering that gas migration and breakthrough are all closely related to the pore structures of the tested geomaterials, a novel gas breakthrough pressure prediction model based on the pore size distribution (PSD) curve was constructed using an existing prediction model based on gas entry pressure and swelling pressure. Finally, based on the test results of the specimens 1.5 Mg/m³, 1.6 Mg/m³ and 1.7 Mg/m³, gas breakthrough pressures of the specimens with dry densities of 1.3 Mg/m³ and 1.4 Mg/m³ were predicted. The results show that the calculated gas breakthrough pressures of 0.76 MPa and 1.28 MPa are very close to the measured values of 0.80 MPa and 1.30 MPa, validating the accuracy of the proposed model.

Site selection for a high-level nuclear waste repository in Germany: a short guide

The government agency BGE started a site selection process for a high-level nuclear waste repository in 2017. The first phase of data screening was completed in 2020 and resulted in the identification of search areas covering 54% of the national territory. The actual second phase will be completed by 2027 or later, and the areas for aboveground investigations will be identified. The obligatory nongovernment input is provided by follow-up regional conferences, which are best supported by external scientific expertise. The large amount of text documents and maps and the interactive internet tools such a 3D viewer provide a suitable basis for external evaluation. The rationale is to apply stricter criteria than those used in the first phase to drastically reduce the size of the search area. With this updated set of criteria, the size of the search area can be reduced to 2,662 km2 or 0.7% of the national territory. This size is convenient for detailed aboveground investigations such as high-density 3D seismic surveys.

Undrained triaxial compression tests on normally consolidated bentonite considering temperature/confining pressure dependency

In the geological repository of high-level radioactive wastes (HLRW), one of the most important issues is the stability of artificial barrier basically using bentonite. However, due to the exist gaps between bentonite block and surround rock mass, the bentonite can swell relatively freely, whose constraining condition is quite different from the normal swelling pressure test. As an extreme case, the bentonite may engage completely into the surrounding ground and become a complete normal consolidated state. Therefore, in order to investigate the thermo-mechanical behavior of saturated bentonite at normally consolidated state, a newly proposed static compaction method for preparing specimen of saturated normally consolidated bentonite was proposed first. Then, a series of undrained triaxial compression tests were conducted under different confining pressure and temperature, by which the thermo-mechanical behavior of bentonite, used as an artificial barrier of geological repository of HLRW, is investigated systematically. The tests results reveal that the specimen prepared by the proposed method is confirmed to have sufficient saturation (> 0.92), meanwhile the prepared specimen is at slightly overconsolidated state that roughly equal to a K0-line consolidation. The deformation of the bentonite in undrained triaxial compression tests changes from plastic to brittle as the temperature increases. Simulation based on a sophisticated constitutive model considering water compressibility at high confining pressure was also discussed. The results are useful for assessing the mechanical behavior of bentonite after hundreds years of geological repository of HLRW.

Foreword: Selected papers from the 2022 International High-Level Radioactive Waste Management Conference

Foreword: Selected papers from the 2022 International High-Level Radioactive Waste Management Conference

Experimental Study on Heat Conduction and Water Migration of Composite Bentonite Samples.

The joints of buffer material composite blocks as potential weak parts in the engineering barrier system of a high-level radioactive waste (HLW) repository must be studied in depth. Therefore, a laboratory experiment device suitable for unsaturated composite bentonite samples was developed. The evolution of temperature and volumetric water content at different locations of Gaomiaozi (GMZ) composite bentonite samples with time before and after simulated water inflow was measured by the experiment device. According to the experimental results, the thermal conductivity and hydraulic conductivity of the joint location after healing of the composite bentonite samples were obtained. The experimental results show that the change in the internal temperature of the composite bentonite samples is mainly affected by the temperature boundary and that the change in the internal water has little effect on it. In a short period of time, the loading of hydraulic boundary conditions only makes the volumetric water content of the soil near the hydraulic boundary increase significantly but has little effect on other locations. And, affected by the temperature boundary, the volumetric water content of the soil near the temperature boundary gradually decreases with time. The process of hydration swelling of the composite bentonite sample is accompanied by the adjustment of stress. The composite bentonite samples are continuously squeezed to the joint area after hydration swelling, the whole composite samples are generally homogenized, and the joints between the composite bentonite samples tend to heal. The thermal conductivity and permeability of the joint location after healing can meet the requirements of the engineering barrier of the HLW repository.

The value of simplified models of radionuclide transport for the safety assessment of nuclear waste repositories: A benchmark study

In order to assess sites for a deep geological repository for storing high-level nuclear waste safely in Germany, various numerical models and tools will be in use. For their interaction within one workflow, their reproducibility, and reliability version-controlled open-source solutions and careful documentation of model setups, results and verifications are of special value. However, spatially fully resolved models including all relevant physical and chemical processes are neither computationally feasible for large domains nor is the data typically available to parameterize such models. Thus, simplified models are crucial for the pre-assessment of possible sites to narrow down the list of suitable candidates for which detailed site investigations and fully resolved models will be done at a later stage. Still, the accuracy of these simplified models is of importance as the pre-assessment of suitable sites will be based on them. In this study, we compare the modelling capabilities of TransPyREnd, a one-dimensional transport code based on finite differences, specifically developed for the fast estimation of radionuclide transport by the German federal company for radioactive waste disposal (BGE), with OpenGeoSys, which is a modelling platform based on finite elements in up to three spatial dimensions. Both codes are used in the site selection procedure for the German nuclear waste repository. The comparison of the model results of TransPyREnd and OpenGeoSys is augmented by comparisons with an analytical solution for a homogeneous material. For the purpose of numerical benchmarking, we consider a geological profile located in southern Germany as an example where the hypothetical repository is located in a clay-stone formation. TransPyREnd and OpenGeoSys yield overall similar results. However, both codes use different discretizations which impact is the highest for strongly sorbing compounds, while the difference gets negligible for less sorbing and more diffusive compounds as higher diffusion tends to blur the initial conditions. Overall, the OpenGeoSys model is more exact whereas the TransPyREnd model has considerable faster run times. We found in our example, that significant substance amounts are only leaving the host rock formation, if apparent diffusion is high, for which case both codes give similar results, while relative differences are considerable for strongly sorbing compounds. However, in the latter case no significant substance amount of radionuclides leaves the host-rock formation, thus deeming the differences in the model results minor for the overall safety assessments of sites.

Elastic modulus prediction for high-temperature treated rock using multi-step hybrid ensemble model combined with coronavirus herd immunity optimizer

Elastic modulus is a crucial mechanical parameter that measures the stiffness property of rock materials. In underground rock engineering, such as deep energy development and geological disposal of high-level nuclear waste, accurately determining the elastic modulus of engineering rock masses in high-temperature environments plays a vital role in understanding the instability-triggering conditions of hard and brittle rock masses in deep underground engineering and maintaining the safety and stability of deep underground engineering. Traditionally, conducting indoor experimental tests is time-consuming, labor-intensive, and costly. Therefore, developing a convenient and accurate rock elastic modulus prediction model based on machine learning technology holds significant importance. To this end, this study proposed a multi-step hybrid ensemble model (MHEM) for predicting high-temperature treated rock elastic modulus. The input parameters of the model include diameter, height, density, temperature, confining pressure, crack damage stress, and strength, while the output parameter is the elastic modulus. Subsequently, a coronavirus herd immunity optimizer (CHIO) intelligent optimization algorithm was employed to optimize the MHEM, and the CHIO-MHEM was established. Then, the performance of the developed models was compared and evaluated with eight other different prediction models. Finally, SHAP method and tree model feature analysis were used to quantify the feature importance of the prediction model. The research results indicate that, compared with other prediction models, the CHIO-MHEM exhibits superior performance, achieving accurate prediction of the elastic modulus of rocks treated at different high temperatures. Additionally, among all input parameters, rock density and temperature have the most significant influence on the rock elastic modulus.

KJ-2 bentonite-sand mixture conditions satisfying THM criteria to develop an enhanced buffer for a high-level radioactive waste repository

Enhanced buffer materials with improved thermal conductivity are needed to reduce the area of high-level radioactive waste repositories. Enhanced buffer materials can be developed by adding substances to bentonite buffer materials. Sand is one such additive; it is a safe and chemically stable natural material with higher thermal conductivity compared to bentonite. In this study, we compare the properties of Jumunjin and silica sands and evaluate their suitability as additives for the KJ-2 bentonite–sand mixture in terms of thermal, hydraulic, and mechanical (THM) performance through the mixture’s thermal conductivity, hydraulic conductivity, and swelling pressure. The results indicate that silica sand, which has smaller particles than Jumunjin sand, is more suitable as an additive. Furthermore, at higher addition ratios, Jumunjin sand tends to be less effective. Therefore, a suitable KJ-2 bentonite–sand mixture, which satisfies THM criteria simultaneously, requires a maximum of 10 % silica sand with a D50 of up to approximately 200 μm, and the dry density of this mixture should be at least 1.7 g/cm3.

Thermally induced release and generation of CO2 and C1–C4 hydrocarbons in Opalinus Clay from Mont Terri (Switzerland)

Claystone formations are candidate host rocks for high-level heat-emitting nuclear waste (HLW). Temperatures from 90 to 150 °C at the canister surface are discussed internationally as potential emplacement and storage conditions. The thermal energy emitted from waste containers will be transported into the host rock formation, accelerating chemical reactions including the release of sorbed and dissolved gases and the generation of new gases. This study investigated gas release and generation in Opalinus Clay from Mont Terri (Switzerland) at elevated temperature and pressure conditions relevant for HLW storage and beyond. Hydrous pyrolysis experiments were conducted in Dickson-type flexible gold-titanium reaction cells and gold capsules in the temperature range of 80–345 °C and at 20 MPa. CO2(g) was the predominant product, followed by C1–C4 hydrocarbons, which decrease in abundance with increasing carbon atom number. Neither CO nor H2S was detected. H2 was generated only in high temperature experiments at 315 °C and 345 °C, respectively. A combination of CO2(g) quantification, stable carbon isotopic composition data, thermodynamic calculations and aqueous fluid composition (dissolved ions, pH) demonstrated that ≥80 % of the measured CO2(g) originated from carbonate mineral dissolution. The model calculations also suggest that the fraction of CO2(aq) in DIC increases from ∼50 % at 80 °C to nearly 100 % at higher temperatures. Thermal transformation of organic matter represented an additional source for CO2(g) and was the predominant process yielding the C1–C4 hydrocarbons. Our findings stress the importance of quantitative geochemical data for the safety assessment of potential host rocks for HLW storage. We demonstrated that two sources are involved in gas release and generation at temperatures relevant for HLW storage, e.g., in the Opalinus Clay – organic matter and carbonate minerals. Our data will contribute to numerical modelling studies and the refinement of feature, events, and processes (FEP) catalogues.

French nuclear glass synthesis: Focus on liquid waste dissolution kinetics

In France, high-level nuclear wastes are confined in glass using a calcination-vitrification process. The waste can also be vitrified by liquid feeding, which imply to add directly the liquid waste. In this study, we focus on nuclear waste vitrification process by direct liquid feeding in order to decipher the dissolution kinetics of the waste in the molten glass. We propose an experimental and analytical protocol to study the dissolution in the solid glass frit of the liquid waste dried beforehand during the glass melting process. This methodology allows to: i) identify the intermediate reaction phases which formed and dissolved during the process; ii) characterizes the evolution of waste element concentrations in the dried waste as a function of time and temperature on glass obtained after cooling; iii) define kinetic parameters relative to the process of intermediate reaction phases dissolution in the molten glass. The three main intermediate reaction phases formed and dissolved in function of temperature are Ca-REE (Rare Earth Elements)- silicate, Ce- oxide and Zr- oxide. At each temperature step (800 to 1200 °C) the evolution of REE (La, Ce, Nd and Pr) and Zr concentrations in the glass samples obtained after cooling shows a fast increase followed by a stabilization in time, which reflects the fast dissolution of the dried waste in the molten glass, i.e. the fast dissolution of intermediate reaction phases. The molten glass is completely homogeneous at classical synthesis temperature (1200 °C). A parameterization methodology of element concentration evolution as a function of time and temperature is proposed here in order to define the kinetic parameters corresponding to the dried waste dissolution (i.e. concentration at a given time t, maximal concentration at a given temperature, characteristic time and activation energy). The results show that the activation related to the formation and dissolution processes of the intermediate reaction phases, i.e. whether they constitute intermediate compounds formed in temperature by the interaction between the dried waste and the glass frit, and the phases formed without reacting with the glass frit, directly in the dried waste during heat treatment.

Comparative study of hydro-mechanical behaviors of compacted bentonite powder and granular bentonite

In the deep geological disposal repository of high-level radioactive waste, buffer/backfill materials typically consist of compacted bentonite block and granular bentonite. As these materials undergo a long-term hydration, it is anticipated that the two forms of bentonite materials (i.e. compacted bentonite powder (CBP) and granular bentonite (GB)) are expected to exhibit differing hydro-mechanical behaviors due to the differences in their structures. This work aims to investigate the differences in swelling pressure and compressibility through a series of swelling pressure tests, compression tests and mercury intrusion porosimetry (MIP) tests. The experimental results demonstrated that swelling pressure curves of the CBP specimens showed higher first peak values and more pronounced collapse than those of the GB specimens at a given dry density, regardless of vapor-water hydration or liquid-water hydration. The final swelling pressures of the two materials were similar at the same dry density, suggesting an independent correlation between swelling pressure and dry density. At the high suction range, the compression curves exhibited an obvious bi-linear pattern for the CBP specimens and a significant nonlinearity for the GB specimens. Meanwhile, the CBP specimens presented higher pre-consolidation pressures and larger compression indices than the GB specimens at a given suction. As suction decreased, the compression curves of the two materials gradually approached each other and their differences were reduced accordingly. After reaching saturation, a good consistency between them was observed whether for final swelling pressure or compressibility. Pore structure analysis revealed that the two materials both presented an initially double structure, and their differences were primarily manifested at the macrostructural level. Eventually, the differences in swelling pressure or compression curves of the two materials were well interpreted by combining microstructural evolutions.

Rapid determination of the migration parameters of nuclides in intact granite rock under the action of electric field

Deep geologic disposal has been widely accepted as a strategy for long-term disposal of the high-level radioactive waste. It is principal to obtain the migration parameters of radionuclides in natural barrier, such as granite, of a high-level radioactive waste repository for safety assessment of the repository. To quickly determine the diffusion and sorption properties of nuclides in intact granite, two tracers, I− and ReO4−, were tested with a modified electromigration device, by imposing a constant voltage over an intact Beishan granitic rock sample. The breakthrough curves of I− and ReO4− were obtained under condition of five different voltages. To interpret the electromigration experimental results with more confidence, an advection-dispersion model based on first-order adsorption kinetics was developed in this study. Data analysis of the breakthrough curves by this model suggest that the effective diffusion coefficients of I− and ReO4− in intact Beishan granodiorite rock are (6.81 ± 0.53) × 10−13 m2/s and (6.45 ± 0.07) × 10−13 m2/s, respectively. While the distribution coefficient of the two ions are (9.06 ± 1.13) × 10−7 m3/kg and (9.81 ± 0.13) × 10−7 m3/kg, respectively. This indicates that I− and ReO4− hardly adsorb in Beishan granodiorite rock.

A CSS surface based THM bounding constitutive model for volumetric behavior of bentonite

Development of the constitutive model for bentonite under coupled thermo-hydro-mechanical (THM) condition is of great significance for the construction and safety assessment of deep geological disposal repositories for high-level radioactive waste (HLW). In this work, a new temperature-suction-mean net stress (T-s-p) space with the conception of critical saturated state (CSS) surface was defined to represent the actual stress state of bentonite under coupled THM condition. Then, based on the CSS surface, a THM constitutive model was proposed for describing the volumetric behavior of compacted bentonite. Under the THM model framework, two bounding surfaces were proposed to describe the elastoplastic volume changes induced by mechanical response of skeleton and hydration of montmorillonite, respectively. The model responses upon some typical THM paths were simulated and discussed to reveal the performance of the proposed constitutive model for bentonite. Finally, the proposed model was validated by simulating by several volume change tests carried out on different bentonites. The results confirmed that the proposed model shows more advantages in describing the THM volumetric behavior.

Kinetics study on the temperature-dependent reduction of aqueous U(VI) by natural pyrite

Previous studies have indicated that acid-washed pyrite is inert toward aqueous U(VI) under most pH conditions at ambient temperature, even though insoluble UO2 is the thermodynamically predicted product for the reduction of U(VI) by pyrite. Considering the exothermic nature of nuclear waste, the interaction between uranyl nitrate/acetate and natural pyrite was studied at temperatures ranging from 25 °C to 85 °C and at pH values between ∼4.0 and ∼9.5, to simulate the scenarios encountered in a high-level radioactive waste repository. The results revealed that the reactivity of pyrite toward aqueous U(VI) significantly increased with rising temperature, and the reaction could be described by a pseudo-first-order kinetic equation. Activation energies were calculated to be 79.4 ± 10.6 and 45.7 ± 3.8 kJ⋅mol−1 for the reduction of uranyl nitrate, and 78.2 ± 5.2 and 42.2 ± 5.8 kJ⋅mol−1 for the reduction of uranyl acetate, at pH values of ∼4.5 and ∼5.0, respectively. These values indicate that the reaction is controlled by the surface chemical processes. In addition, the complete reduction of aqueous U(VI) to UO2 product was first observed for the reactions at pH ∼4.0 to ∼5.5 when the temperature was ≥75 °C. Conversely, non-stoichiometric UO2+x(s) (0 < x ≤ 0.67) was found at lower temperatures within the same pH range, and also at 85 °C with a reaction pH of ∼9.5. Moreover, the ratio of reduced U(IV/V) on pyrite surfaces increased gradually over time, indicating that reaction time played a significant role in the reduction products. The findings are essential not only for the safety assessment of the high-level radioactive waste repository, but also for understanding the metallogenic mechanism of U(IV)-bearing ore deposits under the relevant anaerobic and hydrothermal conditions.

Density spatial distribution and anisotropy of full-scale bentonite-sand blocks

Constructing buffer barriers using high-quality bentonite-based blocks is crucial for the secure disposal of high-level radioactive waste (HLW). However, the density distribution and anisotropy of full-size bentonite-sand blocks are not yet fully understood, which may affect their quality and performance as buffer barriers. This study aimed to address this critical issue by conducting computed tomography (CT) scanning tests on full-size bentonite-sand blocks in the radial, angular, and vertical directions to obtain CT numbers and density. The analysis of the density distribution and anisotropy in these three directions followed. Subsequently, thermal conductivity and hydraulic conductivity tests were conducted in these directions to explore their relationship with density anisotropy. The findings revealed that the block density gradually decreased with increasing radius in the radial direction and first decreased slightly and then increased slightly in the angular and vertical directions. However, the density of the bentonite-sand mixture inside the blocks ranged from 1.96 g/cm³ to 2.09 g/cm³, with minimal differences. Additionally, there was no significant difference in density between the radial and angular directions, with an average density of 2.06 g/cm³. The density in the vertical direction was slightly lower, with an average density of 2.04 g/cm³. Scanning electron microscope images validated the presence of anisotropy in block density caused by horizontally oriented bentonite aggregates resulting from axial compression. However, the presence of quartz sand and incompressible impurities in the blocks created preferential flow paths, mitigating the impact of block density anisotropy on thermal and hydraulic conductivity. Consequently, thermal conductivity, averaging 1.60 W/m·K, and hydraulic conductivity, averaging 3.12 × 10−10 cm/s, demonstrated isotropy in all three directions. These block parameters meet the design requirements for an effective buffer barrier. This study systematically reveals the density distribution of full-size bentonite-sand blocks and its impact on the performance, providing important references for the safe design of HLW repositories.

Phase composition and stability of Gd2−x Th x Zr2O7 under extreme conditions

Abstract Introduction of Th into synthetic disordered fluorite-type Gd2Zr2O7 induces a transition to an ordered pyrochlore-type phase at a Th concentration of 10 % at the Gd site (Gd1.8Th0.2Zr2O7 composition). The degree of order of the fluorite-type phase reaches 50 % for a Th concentration of 25 % (Gd1.5Th0.5Zr2O7 composition). Upon application of high pressure, the Gd2Zr2O7 phase retains the fluorite-type structure until 33 GPa (K 0 = 167(1) GPa), where it undergoes reversible amorphization. The Gd1.7Th0.3Zr2O7 phase was found to be stable up to at least a pressure of 25 GPa (K 0 = 169(3) GPa). Upon heating to T max of 1135 K, the Gd2Zr2O7 phase retains its disordered fluorite-type structural arrangement (α = 3.03 × 10−5 K−1). The excellent stability of the Gd2−x Th x Zr2O7 phases under extreme conditions of temperature and pressure makes Gd2Zr2O7 a promising candidate as a host matrix for radioactive elements for safe long-term underground storage of nuclear waste.