High-level 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.

Research insights of argillaceous-based site for high-level radioactive waste disposal in China

Research insights of argillaceous-based site for high-level radioactive waste disposal in China

Laboratory assessment of real-time water infiltration measurement in SPV200 bentonite using TDR for high-level radioactive waste disposal design

The water content movement within compacted bentonite holds significance in appraising the thermo-hydro-mechanical (THM) performance of the engineered barrier system (EBS) for the disposal of high-level radioactive waste (HLRW), but it demands considerable time to monitor water infiltration in a representative model. Therefore, employing bentonite with varying water content is favored in the laboratory. A comprehensive 3-D phase diagram was established as a benchmark for dielectric constant, temperature, and volumetric water content (VWC) using Time Domain Reflectometry (TDR). SPV200 bentonite at different densities (1400, 1500, and 1600 kg/m3) and temperatures (25, 40, and 60 °C) were prepared with TDR, achieving real-time VWC measurements within a 6 % relative error. Furthermore, numerical outcomes exhibited trends like the TDR tests. Comparation between the simulation and test data is within a relative error ±10 %, which showcased the potential for efficient and effective estimations at a reduced cost for HLRW disposal design.

Research on the implementation of integrated coastal management principles in Taiwan to mitigate disputes related to nuclear waste disposal

Taiwan's nuclear power plants are situated in coastal regions characterized by abundant socio-economic and ecological significance. In the context of managing high-level radioactive waste storage and disposal in the future, alongside technical considerations, there exist challenges related to public acceptance and communication. Radioactive waste management, including the siting of storage or disposal facilities, is a socially sensitive matter. The storage and disposal of radioactive waste represent a complex and multifaceted challenge that requires a comprehensive approach spanning multiple disciplines. One of the functions of Integrated Coastal Management (ICM) is to coordinate conflicts related to coastal use. The concept has been progressively embraced and implemented by nations globally. This study employs grounded theory to examine the perspectives of the interviewed stakeholders. It subsequently introduces the principles of integrated coastal management for conflict resolution post-compilation. The study utilizes the principles and mechanisms of integrated coastal management to address or alleviate the challenges related to radioactive waste management in our country. Issues concerning conflict in relation to storage and disposal.

Highly Efficient Removal of Americium(III) from High-Level Waste Using Novel Phenanthroline Diamide Ligands

Highly Efficient Removal of Americium(III) from High-Level Waste Using Novel Phenanthroline Diamide Ligands

Borehole disposal of spent fuel and other high-level wastes: the case for deep, vertical, fully cased holes in saturated “hard” rock

Driven by major advances in deep drilling technology and the geological understanding of the deep continental crust over the past 70 years, disposal in deep boreholes has moved from being technically unachievable to the point that it now offers a viable solution for the most hazardous nuclear wastes that could effectively be implemented “tomorrow”—i.e., within a few years. Moreover, disposal in deep boreholes is arguably superior in almost every respect to the mined and engineered repositories being pursued for high level waste by most countries. During the first 50 years of their evolution, almost all deep borehole disposal concepts shared five key aspects: (i) the hole was as deep as possible, (ii) it was vertical, (iii) it was fully cased, and (iv) it was in “hard” basement rock (v) saturated with aqueous fluid (groundwater). Technical advances in drilling over the last 20 years have encouraged proposed versions of the concept which depart from one or more of these aspects, but it is our contention that all five fundamental aspects should be retained. This paper summarises the more important arguments supporting this view. In order to meet the necessary post-closure (radiological) safety requirements, engineer out possible operational problems during construction and waste-package deployment, and capitalise on the main benefits of borehole disposal, the hole itself must be over 3 km deep, vertical, fully cased, and in suitably hard (ideally granitic) host rock saturated with aqueous fluid.

Influence of Chemical Weathering and Microcracks on Permeability Variations in Crystalline Rocks

Rock permeability, an important factor in subsurface fluid migration, can be influenced by microcracks and chemical weathering due to water–rock interactions. Understanding the relationship between permeability, chemical weathering, and microcracks is crucial for assessing fluid flow in rocks. This study focuses on the hydrogeological characteristics of granite and gneiss, potential host rocks for high-level radioactive waste disposal in South Korea. Samples were analyzed for permeability, porosity, P-wave velocity, and chemical weathering indices. Regression analysis revealed a weak correlation between permeability and both porosity and rock density, while an inverse correlation was observed between permeability and chemical weathering indices. Interestingly, some samples showed low permeability (10−21 to 10−22 m2) despite high weathering, while others showed high permeability (10−18 to 10−19 m2) despite low weathering. SEM-EDS analysis indicated the presence of microcracks within the rocks or the filling of these cracks with secondary minerals. The findings suggest that chemical weathering generally increases pore size and porosity, but actual permeability can vary depending on the presence and connectivity of microcracks and the extent to which they are filled with secondary minerals. Therefore, both chemical weathering and microcrack connectivity must be considered when evaluating the hydrogeological characteristics of crystalline rocks.

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.

Transdisciplinary research on the safety case for nuclear waste repositories with a special focus on uncertainties and indicators

In the search for a repository site for high-level radioactive waste in Germany, the perception of safety and trust in the actors are central to public acceptance. In communicating safety, methods of safety assessment and the role of uncertainties need to be addressed. Given the complexity of the issue, there is a need for indicators that are suitable both for assessing the long-term safety of repositories and for communicating with the general public. Similarly, there is a requirement to communicate uncertainties in an accessible manner. The TRANSENS project provides basic research in nuclear waste management (NWM) and utilizes a transdisciplinary approach: Non-experts who are not directly affected by the site selection process and who have no stated interest in NWM are involved in the research process, as are practice actors. A series of four transdisciplinary workshops was specifically designed to explore the perspectives of individuals with a high level of disciplinary knowledge but no system knowledge of NWM. Participants were selected from doctoral students in science and technology who had no prior knowledge in this area. Two of these workshops address the questions stated above and are presented here. The article describes the considerations underlying the workshop planning and implementation phases, and the content developed in the workshops on indicator selection and visualisation of uncertainties. The participants compiled a list of desirable indicator properties, which showed a high degree of congruence with the relevant literature. A proposal for a database to collect, administer and assess uncertainties shows similarities with the approach followed by the German implementer and complements it with an interactive visualisation. Transdisciplinary work is resource-intensive and its use in a research context must be carefully considered for each individual application. A transdisciplinary approach was successfully used for the purposes of method validation, method optimisation and the development of disciplinary impulses. An application of transdisciplinary approaches for optimising the Safety Case of nuclear repositories is feasible.

Site differentiation strategy for selective strontium uptake and elution within an all-inorganic polyoxoniobate framework

Selective uptake and elution of trace amounts of hazardous radioactive 90Sr from large-scale high-level liquid waste (HLW) is crucial for sustainable development. Here, we propose a site differentiation strategy, based on the presence of distinct selective metal capture sites (concavity site and tweezer site) within the giant polyoxoniobate (PONb) nanoclusters of an all-inorganic PONb framework (FZU-1). Through this strategy, FZU-1 can not only effectively remove 98.9% of Sr²⁺ from simulated nuclear liquid waste, performing best among the reported Sr adsorbents, but also achieve desorption of adsorbed Sr²⁺ ions by selectively loading Na⁺ ions, thus enabling the recycling of FZU-1. Based on the well-defined single-crystal structures and theoretical studies, it can be revealed that the rapid and selective uptake of Sr²⁺ is attributed to the strong binding energy between the Sr²⁺ ions and the concavity sites. The effective elution of Sr²⁺, on the other hand, stems from the preferential binding of Na⁺ ions at the tweezer sites, elevating the cluster’s electrostatic potential and indirectly facilitating the elution of Sr²⁺ ions. The exceptional stability of FZU-1, along with its rapid and selective Sr²⁺ capture and elution capabilities, positions it as a promising candidate for large-scale nuclear waste treatment and groundwater remediation applications.

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.

Interaction behaviour of alloy 690 upon exposure to P2O5 containing borosilicate glass at simulated vitrification conditions

This study investigates the interaction between alloy 690 and two types of glasses: pristine borosilicate and P2O5-containing borosilicate glass, at typical glass pouring temperatures encountered during the vitrification of high-level radioactive waste in the back-end of Nuclear Fuel Cycle (NFC). Partial crystallization of both glasses was observed at the alloy 690/glass interface, with certain, though not entirely identical, crystalline phases forming at the interface. The density of these crystalline phases is significantly higher than that of the surrounding glass, which raises concerns about these reaction products settling at the bottom of the furnace. Such sedimentation could potentially obstruct the freeze valve, thereby halting the vitrification process. Additionally, intergranular grooves on the alloy surface exposed to P2O5-containing borosilicate glass were found to disappear with prolonged exposure. This phenomenon is attributed to the strong corrosive action of the highly basic and oxidizing P2O5-bearing glass, leading to the peeling away of the entire exposed surface of alloy 690.

The Functionalized N-Rich Covalent Organic Framework for Palladium Removal from Nuclear Wastewater.

Efficiently adsorbing Pd(II) from acidic radioactive waste liquid is crucial for ensuring the safety of the radioactive waste vitrification process and significantly alleviating the scarcity of precious metals. However, the stability and selectivity of most current adsorbents are limited, hindering their practical application under acidic conditions. To address these limitations, a covalent organic framework (DHTP-TPB COF) was prepared with a high nitrogen content, leveraging the high affinity of its soft ligand N with palladium to achieve high selectivity. This work demonstrated that DHTP-TPB COF exhibits rapid adsorption kinetics, with equilibrium achieved within 10 min. The framework also boasts a high adsorption capacity of 142.8 mg/g and impressive reusability in 1.0 M nitric acid. Moreover, the DHTP-TPB COF displays excellent selectivity for Pd(II), even in the presence of 13 interfering ions. By combining FT-IR, XPS spectroscopy, and DFT theoretical calculations, the dense N sites in the framework have a strong affinity for Pd(II), resulting in exceptional adsorption performance that was confirmed. The findings of this study highlight the potential of COFs with robust linkers and customized functional groups to effectively and selectively capture Pd(II) under harsh environmental conditions of high-level liquid waste.

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.

Dissolution behavior of calcium uranate under oxidizing and reducing conditions

Calcium uranate solid phase is a secondary mineral found in geological environments. It may form in the residues of high-level radioactive liquid waste and in the fuel debris of the TEPCO's Fukushima Daiichi Nuclear Power Plant under high-temperature conditions. To assess the chemical stability of CaUO4 in different aqueous environments, static immersion tests were performed under various redox conditions and carbonate ion concentrations. pH and Eh values, as well as concentrations of uranium, calcium, and total carbonate in the solutions, were measured after the supernatants were filtered through a membrane with a 10 kDa molecular weight cutoff. The components of the solid phase were also evaluated using X-ray diffraction and X-ray absorption fine structure analyses. The dissolution mechanism of CaUO4 was examined using data from solid and liquid analyses, along with chemical thermodynamic calculations. Under reducing conditions and without carbonate, U(VI) in CaUO4 was reduced to U(V) and the mineral was converted into non-stoichiometric CaUO4−x. The dissolved uranium was then further reduced to U(IV) in the aqueous media, forming UO2(am), which controlled the U solubility. Under oxidizing conditions and in the absence of carbonate, dissolved uranium formed metaschoepite ((UO3)·2H2O(cr)) at pH ≤ 7 and sodium diuranate (Na2U2O7·H2O(cr)) at pH > 7, which controlled uranium solubility. In oxidizing conditions with carbonate present, the apparent solubility of U was lower than predicted by the solid-phase solubility calculations. The concentration of U was constrained to levels similar to that of calcium when the calcium concentration reached saturation with CaCO3. Additionally, the dissolution of calcium from CaUO4 influenced uranium dissolution.

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.

An acid-resistant covalent organic polymer for detection of palladium ions in high-level liquid waste

An acid-resistant covalent organic polymer for detection of palladium ions in high-level liquid waste

High extraction efficiency of N,N,N′,N′-tetracyclohexyldiglycolamide for Sr(II): An experimental and crystal structure study

To improve the ability of diglycolamide extractants for the extraction of Sr(II) from high-level waste liquid, N,N,N′,N′-tetracyclohexyl-diglycolamide (TCHDGA) was proposed and studied to extract Sr(II) from nitrate media. TCHDGA was prepared and characterized by 1H Nuclear Magnetic Resonance Spectroscopy (NMR), 13C NMR, and Fourier Transform Infrared Spectroscopy (FT-IR). Various factors affecting extraction were studied systematically. In just 20 s, the extraction rate can reach approximately 98.2 %. The extraction capacity of cyclohexyl-substituted extractant TCHDGA is tens of times higher than that with linear or branched chain alkyl. The chemical structure of the complex has been demonstrated to be [Sr 3TCHDGA]·(NO3)2, based on FT-IR, X-ray Photoelectron Spectroscopy (XPS), and crystal structure analysis. The crystal belongs to the monoclinic system, space group P21, and a strontium ion coordinates with nine oxygen atoms, all of which contribute from TCHDGA. The stripping rate can reach over 99 % when using distilled water or 0.50 mol·L−1 oxalic acid as stripping agents.

Evolution of activation energy for boron dissolution in the borosilicate glass during the initial leaching stage

Borosilicate glass is widely used for immobilizing high-level radioactive waste. In this study, two types of borosilicate glasses, Na-borosilicate and Zr-Na-borosilicate, were subjected to leaching within a temperature range of 50-90°C, at 10°C intervals. The rate of boron consumption was measured by Inductively Coupled Plasma Optical Emission Spectrometer (ICP-OES) and Fourier-Transform infrared spectroscopy (FTIR) techniques, and the boron activation energy was determined through the Arrhenius equation. The boron activation energies for Na-borosilicate and Zr-Na-borosilicate glasses are approximately 46 kJ/mol and 42 kJ/mol, respectively. The activation energy, which FTIR calculated, was consistent with that obtained by ICP-OES during the initial leaching time and decreased as the leaching time increased. The decrease in activation energy of boron is because the [BO3] structure was replaced by molecular water. The FTIR method offers an alternative approach for calculating the boron activation energy and contributes to understanding the leaching mechanism in the initial leaching stage.

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.