High-level Waste Research Articles

The radiation resistance and chemical stability optimization of analog actinide nuclides incorporated Gd2Zr2O7 ceramics via grain size control

Gd2Zr2O7 is considered a promising host material for immobilizing high-level radioactive waste. However, its interplay between grain sizes, radiation resistance and chemical stability is not well understood yet. Herein, Gd1.7Nd0.3Zr0.5Ce1.5O7 ceramics with defect-fluorite phase and three different grain sizes (96, 390 and 1959 nm) were successfully fabricated and irradiated by 140 keV He ion beam up to a fluence of 8 × 1017 ions/cm2. Phase evolutions, microstructure changes and chemical stabilities were investigated. The nano-grained ceramic exhibits the lowest degree of amorphization and a delayed irradiation-induced lattice expansion and damage evolution process, demonstrating an enhanced ability in suppressing the coalescence of He bubbles inside the grain. Grain boundaries parallel to the surface were found to be more favorable sinks for irradiation-induced defects. Chemical stability tests showed that the 42 days leaching rates of all elements were between 10−7∼10−5 g/m2·d. All samples have maintained this superior chemical stability even after irradiation. Interestingly, the leaching test results also suggested that grain size may be not the primary factor influencing the long-term leaching rate.

A review on thermal conductivity of unsaturated bentonite

As unsaturated bentonite is used a buffer/backfill material in the construction of engineered barriers for high-level nuclear waste disposal, understanding its thermal properties is crucial for maintaining the operational stability of the repository. This review paper synthesizes research on the thermal conductivity of unsaturated bentonite, encompassing aspects of heat transfer mechanisms, measurement techniques, influencing factors, and prediction models. The results highlight that the transient and steady-state methods have emerged as the predominant techniques for measuring the thermal conductivity of unsaturated bentonite owing to their efficiency and the prevention of water redistribution within the sample throughout the testing phase. Factors such as dry density, degree of saturation, additive content, and temperature were observed to positively influence the thermal conductivity of unsaturated bentonite, whereas an increase in the ion concentration of the pore fluid decreased the thermal conductivity. The prediction models for the thermal conductivity of unsaturated bentonite are categorized into empirical, normalized, and theoretical models. Although empirical and normalized models provide some insight, they suffer from a lack of theoretical foundation, and the parameters they incorporate often lack clear physical significance, complicating their practical application despite their capacity to represent the multifield coupling characteristics of thermal conductivity. The advancements in microscopic testing methods and computational technology herald the potential of mesoscale models and machine learning as formidable tools for predicting the thermal conductivity of unsaturated bentonite, suggesting a promising direction for future research.

Insight into the elastoplastic behavior of Beishan granite influenced by temperature and hydraulic pressure

Granite is an ideal host rock for high-level radioactive waste (HLW) repositories. In the deep environment of HLW repositories, granite is subjected to variable temperature and hydraulic pressure. In order to investigate the influence of temperature and hydraulic pressure on the elastoplastic behavior of the Beishan granite, the present paper implemented a set of triaxial compression experiments at various temperatures and hydraulic pressures corresponding to the environment of HLW repositories. The deformation processes, damage evolution, elastic coefficients, and strength of the Beishan granite were analyzed, resulting in a limited change versus temperature and hydraulic pressure. The residual strengths of granite are similar, but a higher temperature surpasses about 70 °C degrades the friction strength of granite. Evidence from the acoustic emission (AE) hits shows the damage rate is accelerated by improved temperature and hydraulic pressure. According to the experiment-based mechanism, a modified elastoplastic model was proposed which considered thermal pressurization and changeable compressibility. The results of computation based on the proposed model match well the experiment data.

Measurements of Thermodynamic Data of Water in Ca-Bentonite by Relative Humidity Method

Buffer material (compacted bentonite), one of the engineered barrier elements in the geological disposal of a high-level radioactive waste, develops swelling stress due to groundwater penetration from the surrounding rock mass. Montmorillonite is the major clay mineral component of bentonite. Even previous studies provide few mechanical and thermodynamic data on Ca-montmorillonite. In this study, thermodynamic data on Ca-montmorillonite were obtained as a function of water content by measuring relative humidity (RH) and temperature. The activities of water and the relative partial molar Gibbs free energies of water were determined from the experimental results, and the swelling stress of Ca-bentonite was calculated using the thermodynamic model and compared with measured data. The activities of water and the relative partial molar Gibbs free energies obtained in the experiments decreased with decreasing water content in water contents lower than about 25%. This trend was similar to that of Na-montmorillonite. The swelling stress calculated based on the thermodynamic model was approximately 200 MPa at a montmorillonite partial density of 2.0 Mg/m3 and approximately 10 MPa at a montmorillonite partial density of 1.4 Mg/m3. The swelling stresses in the high-density region (around 2.0 Mg/m3) were higher than that of Na-montmorillonite and were similar levels in the low-density region (around 1.5 Mg/m3). Comparison with measured data showed the practicality of the thermodynamic model.

New approaches to the separation and concentration of americium in high oxidation forms for the fractionation of high-level waste

New approaches to the separation and concentration of americium in high oxidation forms for the fractionation of high-level waste

Development of thermal-hydraulic coupling model for deep-geological disposal of high-level radioactive wastes

The buffer plays an important role in a engineered barrier system for the deep geological repository of high-level radioactive waste (HLW). High temperature may result in degradation of buffer performance. It is crucial to investigate the buffer temperature behavior so that its peak temperature is not higher than the criterion of 100 °C. In the present work, the thermal-hydraulic (T-H) coupling model is developed to simulate the histories of temperature and saturation in the disposal repository. This model is validated with the previous simulation works in the DECOVALEX 2015 Task B2 in order to simulate the geological repository of HLW with confidence. The whole repository of HLW in the present works can be reasonably simplified as a single hole with the symmetric boundary conditions on its peripheral sides. Effects of tunnel spacing and hole spacing of repository on the histories of buffer peak temperature are considered since the layout with various spacing is strongly related to the peak temperature of buffer. Based on the comparison of simulation results from the T-H coupling model and T-only model for the different layout of whole HLW repository, the maximum value of peak temperature in the buffer predicted by the T-H coupling model is lower than that from the T-only model. In other words, less area is needed for the selection of repository, rendering that the present T-H coupling model can assist in the design of HLW repository.

A four-fold interpenetrated MOF for efficient perrhenate/pertechnetate removal from alkaline nuclear effluents

The sequestration of 99Tc represents one of the most challenging tasks in nuclear waste decontamination. In the event of a radioactive waste leak, 99TcO4– (a main form of 99Tc) would spread into the groundwater, a scenario difficult to address with conventional anion exchange materials like resin and inorganic cationic sorbents. Herein, we present a nickel(II) metal-organic framework (MOF), TNU-143, featuring 3D four-fold interpenetrated networks. TNU-143 exhibits efficient ReO4– (a nonradioactive analogue of 99TcO4–) removal with fast anion exchange kinetics (<1 min), high sorption capacity (844 mg/g for ReO4–), and outstanding selectivity over common anions. More importantly, TNU-143 shows superior stability in alkaline solution and can remove 91.6% ReO4– from simulated alkaline high-level waste (HLW) streams with solid-liquid ratio of 40 g/L. The uptake mechanism is elucidated by the single-crystal structure of TNU-143(Re), showing that ReO4– anions are firmly coordinated to nickel cation to result in a 2D layered structures. Density functional theory (DFT) calculations confirm the transformation from TNU-143 to TNU-143(Re) is a thermodynamically favorable process. This work presents a new approach to the removal of ReO4–/99TcO4– from alkaline nulcear fuel using MOF sorbents.

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.

Curing temperature’s effect on argillite-metakaolin based geopolymer grouts

This study concerns the project of high and intermediate level radioactive waste disposal (Cigéo project for Centre Industriel de stockage GEOlogique). Millions of tons of argillite will be excavated during the construction of the site. Therefore, the aim of this work is to reuse this argillite in the formulation of an injectable geopolymer grout and study its evolution in different storage conditions (T = 20 and 80 °C). To achieve this challenge, a metakaolin and calcined Callovo-Oxfordian (COx) argillite were used with potassium silicate solutions. The grouts were successfully obtained from both metakaolin and COx argillite in both temperatures 20 and 80 °C. The range of viscosity of several mixtures respected the required specifications (viscosity ≤5 Pa s and setting time superior to 6 h). The metakaolin-based geopolymers didn’t display any major changes no matter the storage temperature, except for the pore size, a coalescence phenomenon was observed with the increase of temperature. In contrast, for argillite or argillite-metakaolin based samples, a decrease of pore size as well as in pH values were observed with the increase of temperature. This behaviour was attributed to the dissolution of argillite’s impurities such as illite, providing more oligomers to allow the formation of different geopolymer networks.

Water Content Evolution in the EDZ of Opalinus Clay: A Methodic Approach for a Comparative Interpretation of Measurements and Modelling

In the Mont Terri Rock Laboratory (Switzerland), an interdisciplinary examination program is carried out to increase knowledge about coupled hydro-mechanical effects in Opalinus Clay, which are of significant interest regarding the stability and integrity of a potential storage facility for high-level radioactive waste. This article focuses on the characterization of the claystone in the near field of excavations and related hydraulic effects due to excavation and ventilation. Electrical resistivity tomography (ERT) is applied to characterize the OPA: Several open fractures correlate with regions of high resistivity values, indicating potential preferential flow paths that are relevant for transport processes. Due to the combined interpretation of ERT long-term monitoring and seasonally repeated nuclear magnetic resonance (NMR) measurements, a relationship between electrical resistivity and water content can be established, resulting also in a time-dependent map of the water content around excavations with different climatic conditions. The statistical interpretation of these measurements indicates the existence of small-scale singularities in contrast to dominating, more homogeneous zones. The presented approach leads to a better process understanding of these heterogeneous near field effects and provides a valuable basis for a pragmatic approach to safety assessment.

Research on calcination thermal decomposition process of simulated high-level liquid waste based on two-step vitrification

The two-step vitrification process involves converting high-level waste from a liquid to a solid state through calcination, followed by melting the calcines with glass matrix material to form glassy wasteforms. The calcination process is a critical stage in the vitrification process. This study investigated the thermal decomposition performance and phase evolution of power reactor simulated high-level waste during the calcination process. As the calcination temperature increased from 100 to 900 °C, the resultant waste calcines showed increasingly darker color and smaller general particle sizes. Below 600 °C, the waste experienced thermal decomposition of nitrate and formate compounds, leading to an approximate 40 wt% weight reduction. At 600 °C, the waste's weight remained constant, and crystalline phases such as ZrxM1-xO2 (M = Ce, Pr, Nd), BaMoO4, CeO2, and GdxCe1-xO2 were observed.

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.

High valency of charge compensator (Mo6+) to substitute Ti site in REE doped zirconolite (REE=Nd, Sm, Gd, Ho and Yb): Solid solubility, phase evolution and structural analysis

Zirconolite ceramics have been proven to be a promising matrix for high-level waste immobilization, especially for minor actinides. In this study, a series of CaZr1-2xREE2xTi2-xMoxO7 (REE = Nd, Sm, Gd, Ho, Yb) samples were prepared to investigate the formation of zirconolite REE as surrogates of minor actinides and Mo6+ as charge compensator. Synchrotron and in-house powder X-ray diffraction (XRD), X-ray absorption near edge spectroscopy (XANES), and scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDX) were used to study their solid solubility, phase evolution, and structural details. XRD results show that the formation of zirconolite-2M can be up to x = 0.15 in the CaZr1-2xNd2xTi2-xMoxO7 samples when co-existence of 2 M and 4 M was observed at x = 0.20 and 0.25. The “yellow phase” appears when x exceeds 0.4. The phase evolutions of CaZr1-2xSm2xTi2-xMoxO7, CaZr1-2xGd2xTi2-xMoxO7, CaZr1-2xHo2xTi2-xMoxO7 and CaZr1-2xYb2xTi2-xMoxO7 solid solutions are similar to each other: (1) zirconolite-2M in x = 0.05 and 0.10; (2) co-existence of 2 M and 4 M and then formation of 4 M phase from x = 0.15 to 0.25; (3) formation of “yellow phase” and pyrochlore when x ≥ 0.30. Interestingly, perovskite exits in all the ceramic samples, and its lattice parameters slightly change when increasing the dopants. Rietveld refinement results reveal that Sm/Ho mainly incorporates into 8f (Wyckoff position) Zr-sites while Mo replaces the 8f (Wyckoff position) Ti-sites. XANES results further confirm the occupation of Mo in TiO5 coordination environments. Furthermore, there is a considerable amount of Sm/Ho incorporated into 8f Ca-sites, demonstrating a different substitution mechanism from the preliminary design.

Quantifying the impact of upscaled parameters on radionuclide transport in three-dimensional fracture-matrix systems

Assessing the long-term safety of geological repositories for high-level radioactive waste is critically dependent on understanding radionuclide transport in multi-scale fractured rocks. This study explores the influence of upscaled parameters on radionuclide movement within a three-dimensional fracture-matrix system using a discrete fracture-matrix (DFM) model. The developed numerical simulation workflow includes creating a random discrete fracture network, meshing of the fractures and matrix, assigning upscaled parameters, and conducting finite element simulations. We simulated the spatiotemporal evolution of radionuclide concentrations in the fractures and matrix over a century, revealing significant spatial heterogeneity driven by a heterogeneous seepage field. Employing geostatistics-based upscaling methods, we predicted the effective ranges of crucial solute transport parameters at the field scale. The matrix diffusion coefficient, matrix distribution coefficient, and longitudinal dispersivity were upscaled by factors of 2.0–3.0, 2.5–4.0, and 10–104, respectively, based on laboratory-scale measurements. Incorporating these upscaled parameters into the DFM model, we analyzed their impact on radionuclide transport. Our findings demonstrate that an upscaled matrix diffusion coefficient and matrix distribution coefficient result in a delayed transport of radionuclides in fractures by enhancing mass transfer between the fractures and rock matrix, while an upscaled longitudinal dispersivity accelerates transport by advancing the positions of concentration peaks. Sensitivity analysis revealed that the matrix distribution coefficient is the most impactful, followed by dispersivity and matrix diffusion coefficient. These insights are important for minimizing parameter uncertainties and enhancing the accuracy of predictions concerning radionuclide transport in multi-scale fractured rocks.

Swelling Stress of Bentonite: Thermodynamics of Interlayer Water in K-Montmorillonite in Consideration of Alteration

The buffer material that makes up the geological disposal system of high-level waste swells by contact with groundwater and seals space with rock mass and fractures in rock mass. The buffer material has a function of mechanical buffer with rock pressure, and swelling stress is important in this case. The alteration of bentonite may occur due to the initial replacement of cations (Na+ ions) in the interlayer with K+ ions upon contact with groundwater, but there are no studies on the swelling stress of K-bentonite. In this study, the author prepared K-montmorillonite samples and obtained thermodynamic data on interlayer water as a function of water content using a relative humidity method. The swelling stress was analyzed based on a thermodynamic model developed in earlier studies and compared with measured data. The activity and the relative partial molar Gibbs free energy of porewater decreased with decreasing water content in the region, below approximately 15%. This behavior significantly differs from that of other ions, such as Na. The swelling stress calculated based on the thermodynamic model and date occurred in the region of high density of 1.9 Mg/m3 with montmorillonite partial density. It was indicated for the first time that K-bentonite scarcely swells under realistic design conditions.

Evaluating Material Attractiveness of Minor Actinide Nuclear Fuel Intended for a Waste Transmutation Lead-Cooled Fast Reactor

Long-lived high-level waste from commercial nuclear power reactors is a problem that concerns stakeholders and scientists working in the back end of the nuclear fuel cycle. Nuclear waste transmutation is under investigation to tackle this problem, transforming nuclides that represent a long-term source of radioactivity, radiotoxicity, and heat into short-lived or stable nuclides. However, the transmutation process will require that several long-lived isotopes be separated from the spent nuclear fuel, which raises proliferation concerns. In this paper, we perform an investigation of the attractiveness characteristics related to the material used in a lead-cooled fast reactor system concept designed to burn minor actinides before and after irradiation. The materials evaluated are separated uranium, neptunium, plutonium, americium, and curium. We also evaluated grouped product materials, neptunium + americium and neptunium + plutonium. Additionally, we present potential safeguards and physical protection implications for the proposed materials. The main conclusion of this paper is that the separated neptunium and plutonium generated by the fast reactor are materials that deserve attention mainly related to physical protection measures.

Phosphate-Based Dechlorination of Electrorefiner Salt Waste using a Phosphoric Acid Precursor.

Electrochemical processing of spent nuclear fuel in molten chloride salts results in radioactive salt waste. Chlorine removal from the salt has been identified as an effective and efficient first step in the management of high-level waste. In this work, a simple salt was dechlorinated with a phosphoric acid phosphate precursor, resulting in a glassy dechlorinated product. The dechlorination efficacy was evaluated in air and argon environments. This work serves as an initial step to advance the Technological Readiness Level of H3PO4-based dechlorination step toward implementation of iron phosphate waste forms to immobilize electrochemical fuel reprocessing salt waste streams.

A thermo-hydromechanical damage model and its application to a deep geological radioactive repository

This study conducts a numerical analysis of long-term thermo-hydromechanical (THM) processes, with a particular focus on deep geological disposal of radioactive waste. It emphasizes the modeling of damage zones resulting from excavation activities, as well as changes in pore pressure and temperature. The numerical model presented in this paper delineates the fundamental relations of thermo-poro-elasticity. It also introduces a double phase field model specifically formulated for rock materials under compressive stress, taking into account THM coupling and time-dependent behavior. The proposed model has been implemented in a finite element code and applied to numerical analysis of short and long terms responses around a horizontal borehole in the context of the French High-Level Waste (HLW) disposal concept. In particular, displacements, temperature changes, pore fluid pressure variations as well as evolution of induced damage zones are presented and analyzed for different periods including excavation, operation and post-closure. It is found that the occurrence of damage zones is clearly related to the spacing distance between two neighboring disposal boreholes. The time-dependent deformation of host rock plays an significant role in the long-term responses.

Changes of Temperature and Moisture Distribution over Time by Thermo-Hydro-Chemical (T-H-C)-Coupled Analysis in Buffer Material Focusing on Montmorillonite Content

Bentonite is used as a buffer material in engineered barriers for the geological disposal of high-level radioactive waste. The buffer material will be made of bentonite, a natural clay, mixed with silica sand. The buffer material is affected by decay heat from high-level radioactive waste, infiltration of groundwater, and swelling of the buffer material. The analysis of these factors requires coupled analysis of heat transfer, moisture transfer, and groundwater chemistry. The purpose of this study is to develop a model to evaluate bentonite types and silica sand content in a unified manner for thermo-hydro-chemical (T-H-C)-coupled analysis in buffer materials. We focused on the content of the clay mineral montmorillonite, which is the main component of bentonite, and developed a model to derive the moisture diffusion coefficient of liquid water and water vapor based on Philip and de Vries, and Kozeny–Carman. The evolutions of the temperature and moisture distribution in the buffer material were analyzed, and the validity of each distribution was confirmed by comparison with the measured data obtained from an in situ experiment at 350 m in depth at the Horonobe Underground Research Center, Hokkaido, Japan.

The mechanical properties and chemical durability of granite wastes based glass-ceramics for immobilization of high-level nuclear wastes

In this study, granite wastes based glass-ceramics containing Eu2O3 as simulated radionuclides is successfully synthesized via solid-state sintering method. The structure, mechanical properties and chemical durability are deeply investigated. The main phase in this glass-ceramics is oxyapatite (Ca2Eu8(SiO4)6O2), which is dispersed homogenously in glass matrix. The density dominates the properties of Vickers hardness and bending strength. The glass-ceramics containing 40 wt % of Eu2O3 has the optimal mechanical properties, with the Vickers hardness, bending strength and density of 5.12 GPa, 79.44 MPa and 98.77 %, respectively. The 28 days leaching experiments show that the glass-ceramics containing 40 wt % of Eu2O3 possess excellent chemical durability. The leaching rate of Al3+, Si4+, Fe3+, Ca2+, Eu3+ are 2.3 × 10−4 g m−2 d−1, 6.0 × 10−4 g m−2 d−1, 1.2 × 10−7 g m−2 d−1, 1.4 × 10−4 g m−2 d−1, 3.6 × 10−8 g m−2 d−1, respectively. This study indicates that granite wastes based glass-ceramics have excellent chemical stability and mechanical properties, which provides a novel idea for the immobilization of radioactive wastes.