Thermodynamic sorption modeling can enhance confidence in assessing and demonstrating the radionuclide sorption phenomena onto various mineral adsorbents. In this work, Ca-montmorillonite was successfully purified from Bentonil-WRK bentonite by performing the sequential physical and chemical treatments, and its geochemical properties were characterized using X-ray diffraction, Brunauer-Emmett-Teller analysis, cesium-saturation method, and controlled continuous acidbase titration. Further, batch experiments were conducted to evaluate the adsorption properties of Cs(I) and Sr(II) onto the homoionic Ca-montmorillonite under ambient conditions, and the diffuse double layer model-based inverse analysis of sorption data was performed to establish the relevant surface reaction models and obtain corresponding thermodynamic constants. Two types of surface reactions were identified as responsible for the sorption of Cs(I) and Sr(II) onto Ca-montmorillonite: cation exchange at interlayer site and complexation with edge silanol functionality. The thermodynamic sorption modeling provides acceptable representations of the experimental data, and the species distributions calculated using the resulting reaction constants accounts for the predominance of cation exchange mechanism of Cs(I) and Sr(II) under the ambient aqueous conditions. The surface complexation of cationic fission products with silanol group slightly facilitates their sorption at pH > 8.

Radionuclides in low- and intermediate-level radioactive wastes from the decommissioning process of nuclear power plants were generally immobilized by cementation methods. Ethylenediaminetetraacetic acid (EDTA), which is extensively used as a decontamination agent, can affect the behaviors of radionuclides immobilized in cement waste forms. In this study, the effects of EDTA contained in simulated radioactive decommissioning wastes on the leaching characteristics of immobilized Co and Cs and the microstructure evolution of cement waste form. Co leaching was accelerated by the formation of Co–EDTA complexes with high mobility and solubility. Cs leaching was hindered by the ion competition with other metal–EDTA complexes for releasing from the cement waste form. Cs leaching was also retarded by carbonated layer at edge of the cement waste form, which process is facilitated by the presence of EDTA. Finally, the effects of EDTA on the leaching characteristics of immobilized Cs and Co and the microstructure evolution of the cement waste form should be considered to ensure the safety of disposal for lowand intermediate-level radioactive wastes.

Montmorillonite, a versatile clay mineral with a wide range of industrial applications, is often found in natural deposits with impurities that limit its effectiveness. This study investigates the use of column froth flotation as an innovative technique to improve the purity of montmorillonite by selectively removing impurities without affecting its essential properties. Column froth flotation, a well-established mineral separation method, is adapted to address the specific challenges associated with enhancing montmorillonite purity. The process involves conditioning raw montmorillonite with carefully chosen reagents to selectively separate impurities, including quartz, feldspar, and other minerals commonly found alongside montmorillonite in natural deposits. Experimental results confirm the effectiveness of column froth flotation in significantly enhancing the purity of montmorillonite. This method allows for efficient impurity removal while preserving the essential properties of montmorillonite, making it suitable for various industrial applications. The study also explores the optimal conditions and reagent choices to maximize the purification process. In conclusion, column froth flotation offers a promising avenue for enhancing montmorillonite purity without compromising its fundamental properties. This study provides valuable insights into optimizing the process for large-scale industrial applications, facilitating the development of highquality montmorillonite products tailored to specific industrial needs.

The permanent disposal of discharged spent nuclear fuel (SNF) and contaminated radioactive waste generated from the subsequent chemical treatments of SNF has become a serious pending issue in many countries that operate the nuclear power plants. Among the diverse engineering solutions proposed for the disposal of high-level radioactive waste (HLW), deep geological disposal (DGD) has been considered as the most proven and safe option to prevent any significant release of radionuclides into the biosphere and to predictably ensure the long-term performance of disposal system. The DGD system consists of multiple structural components; the bentonite clay-based buffer and tunnel backfills are designed to perform the primary hydrogeochemical functions of 1) inhibiting the ingress of groundwater and reactive substances that could compromise the integrity of canister and 2) retarding the migration of released radionuclides into biosphere by providing the sufficient chemisorption sites. Montmorillonite, which is a 2:1 phyllosilicate mineral belonging to smectite group, constitutes the majority of bentonite, and it mainly predominate the swelling and chemisorption capacities of the clay material. Thus, it is essentially required to thoroughly understand the chemical interactions of major radionuclides and other important substances with montmorillonite in advance to accurately evaluate the long-term retention performance of engineered barriers and to reduce the uncertainties in the safety assessment of a deep geological repository (DGR) ultimately. Thus far, sorption of dissolved species onto mineral adsorbents has been generally described and quantified using the simple sorption-desorption distribution coefficient (Kd) concept; since any specific reaction mechanisms are not considered and reflected in the Kd concept, an empirical Kd value is intrinsically dependent on the aqueous conditions under which it was measured. In this framework, substantial scientific efforts have been made to develop a robust basis for geochemically parametrizing the sorption phenomena more reliably, and the application of thermodynamic sorption modeling (TSM), which is based on the chemical principle of mass action laws, has been studied with the aim of improving overall confidence in the description of radionuclide migration under a wide range of aquatic conditions. The disposal performance demonstration R&D division of KAERI introduced a new reference Ca-bentonite clay called Bentonil-WRK (Clariant Korea) for HLW disposal research in 2021 as the domestic Ca-bentonite sources have being depleted. We successfully separated and purified Ca-montmorillonite from the Bentonil-WRK clay, and its geochemical characteristics were meticulously studied by means of XRD, BET, CEC, FT-IR analyses and controlled acid-base titration. In this work, chemical sorption behaviors of aqueous iodide and benzoate, which are a major fission product in HLW and a model ligand of complex natural organic matters present in the deep geological environment, onto the purified Camontmorillonite were assessed under ambient conditions of S/L = 5 g/L, I = 0.01 M CaCl2, pH = 4- 9, pCO2 = 10-3.4 atm, and T = 25°C. Further, their unique adsorption envelopes and corresponding thermodynamic reaction constants refined from the diffuse double layer model (DDLM)-based inverse modeling of experimental sorption data were discussed.

In the design of HLW repositories, it is important to confirm the performance and safety of buffer materials at high temperatures. Most existing models for predicting hydraulic conductivity of bentonite buffer materials have been derived using the results of tests conducted below 100°C. However, they cannot be applied to temperatures above 100°C. This study suggests a prediction model for the hydraulic conductivity of bentonite buffer materials, valid at temperatures between 100°C and 125°C, based on different test results and values reported in literature. Among several factors, dry density and temperature were the most relevant to hydraulic conductivity and were used as important independent variables for the prediction model. The effect of temperature, which positively correlates with hydraulic conductivity, was greater than that of dry density, which negatively correlates with hydraulic conductivity. Finally, to enhance the prediction accuracy, a new parameter reflecting the effect of dry density and temperature was proposed and included in the final prediction model. Compared to the existing model, the predicted result of the final suggested model was closer to the measured values.

A deep geological repository for disposal of high-level radioactive waste (HLW) consists of the canister, buffer material, and natural rock. If radionuclides leak from a disposal container, it can pass through buffer materials and rock, and move into the biosphere. Transport and migration of radionuclides in the rock differently were affected by the fracture type, filling minerals in the fracture, and the chemical and hydraulic properties of the groundwater. In this study, aperture distribution in fractured granite block was investigated by hydraulic test and CFD analysis. The fractured rock block (1 m × 0.6 m × 0.6 m), which is simulated as natural barrier, was prepared from Iksan, Jeollabuk-do. 9 test holes were drilled and packer system was installed to perform hydraulic test at the surface of fracture. 3D model simulated for aperture distribution of rock block was made using results of hydraulic test. And then, CFD analysis was performed to evaluate the co-relation between experiment results and analysis results using FLUENT code.

The safe disposal of high-level radioactive waste (HLW), including the discharged spent nuclear fuel (SNF) and contaminated by-products produced from relevant chemical treatments, has become a serious pending problem for numerous countries that operate the nuclear power plants. The deep geological disposal (DGD) has thus far been considered the most proven and viable solution for isolation of the HLW and preventing any significant release of radionuclides into the biosphere. The DGD system consists of the multiple engineered and natural barrier components. Among them, the montmorillonite-based buffer and tunnel backfills are designed to perform the two major geochemical functions: 1) preventing the ingress of groundwater and any chemicals that compromise the safety of waste canister and 2) retarding the migration of released radionuclides by providing sufficient chemisorption sites. Therefore, it is essential to investigate the sorption mechanism of radionuclides onto montmorillonite and develop a thermodynamic reaction model in advance in order to accurately predict the long-term performance of engineered barriers and to reduce the uncertainties in the safety assessment of a deep geological repository (DGR) ultimately; thus far, sorption of chemical species onto mineral adsorbents has been widely described based on the concept of sorption-desorption distribution coefficient (Kd), the value of which is intrinsically conditional, and active scientific efforts have been made to develop robust thermodynamic sorption models which offer the potential to improve confidence in demonstration of radionuclide migration under a wide range of geochemical conditions. The natural montmorillonites are generally classified into Na-type or Ca-type according to its exchangeable cation, and the Ca-montmorillonite containing clays are being considered as candidate materials for the engineered barriers of DGR in several countries; they generally have advantages of higher thermal conductivity and lower price than the Na-montmorillonite based clays, but their sorption capacities are still comparable. In this framework, we aimed to investigate the chemical interactions of Ca-montmorillonite with selenite [Se(IV)], which is a major oxyanionic species in terms of HLW disposal, and develop a reliable thermodynamic sorption model (TSM). The present work summarizes the characterization of Ca-montmorillonite separated from the newly adopted reference bentonite (Bentonil-WRK) by means of XRD, BET, FTIR, CEC measurement, and acid-base titration. Further, its sorption behaviors with aqueous selenite species under aqueous conditions of S/L = 5 g/L, I = 0.01-0.1 m CaCl2, pH = 4.5-8.5, pCO2 = 10-3.5 atm, and T = 25°C were examined, and the resulting thermodynamic data are discussed as well.

Bentonite is a potential buffer material of multi-barrier systems in high-level radioactive wastes repository. Montmorillonite, the main constituent of the bentonite, is 2:1 type aluminosilicate clay mineral with high swelling capacity and low permeability. Montmorillonite alteration under alkaline and saline conditions may affect the physico-chemical properties of the bentonite buffer. In this study, montmorillonite alteration by interaction with synthetic alkaline and saline solution and its retention capacity for cesium and iodide were investigated. The experiments were performed in three different batches (Milli-Q water, alkaline water, and saline water) doped with cesium and iodide for 7 days. Alteration characteristics and nuclide retention capacity of original- and reacted bentonite was analyzed by X-ray diffraction (XRD), Fourier Transform Infrared (FTIR) spectroscopy, Scanning Electron Microscope (SEM), Nuclear Magnetic Resonance (NMR) and Cation Exchange Capacity (CEC) analysis. From the results, cesium retention occurred differently depending on the presence of competing ions such as K, Na, and Mg ions in synthetic solutions, while iodide was negligibly removed by bentonite. Montmorillonite alteration mainly occurred as cation exchange and zeolite minerals such as merlinoite and mordenite were new-formed during alkaline alteration of the montmorillonite. CEC value of reacted bentonite increased by formation of the zeolite minerals under alkaline conditions.

A disposal research program for HLW has been carried out since 1997 with the aim of establishing the preliminary concept of geological disposal in Korea. The preliminary studies were conducted by conducting manufacture and installation of an in-situ nuclide migration system in KAERI Underground Research Tunnel (KURT). Nuclides could be released from a deep underground disposal facility due to thermal and physicochemical changes into the surrounding environments. Understanding on the migration and retardation processes of nuclides in a fractured rock is very important in the safety assessment for the radioactive waste disposal. In this study, we evaluated fracture filling minerals and aperture distribution (3D map) along the fracture surfaces under the controlled conditions. The fractured granite block which has a single natural fracture of 1 m scale was sampled in a domestic quarry (Iksan), which groundwater had been flowed through. This rock has an interconnected porosity of 0.36 with the specific gravity of 2.57. The experimental set-up with the granite block with dimensions of 100×60×60 (cm). A flow of de-ionized water through the fracture between pairs of boreholes was initiated and the pressure required to maintain a steady flow was measured. In additions, fracture filling minerals were sampled and examined by mineralogical and chemical analyses. There are phyllosilicate minerals such as illite, kaolinite, and chlorite including calcite, which are fracture filling minerals. The illite and kaolinite usually coexist in the fracture, where their content ratio is different according to which mineral is predominant. For the evaluation of fracture, surface was divided into an imaginary matrix of 20×20 sub-squares as schematically. The calculated results are expressed as a two dimensional contour and a three dimensional surface plot for the aperture distribution in the fracture. The aperture value is distributed between 0.075 and 0.114 mm and the mean aperture value is 0.095 mm. The fracture volume is about 55 ml. Also the 137Cs sorption (batch test) distribution coefficients increased to Kd = 800~860 mL/g in the fractured rock because of the presence of secondary minerals formed by weathering processes, compared to the bedrock (Kd = 750~830 mL/g). These results will be very useful for the evaluation of environmental factor affecting the nuclides migration and retardation.

Colloid Formation and Migration (CFM) project is being carried out within the Grimsel Test Site (GTS) Phase Ⅵ. Since 2008, the Korea Atomic Energy Research Institute (KAERI) has joined CFM to investigate the behavior of colloid-facilitated radionuclide transport in a generic Underground Research Laboratory (URL). The CFM project includes a long-term in-situ test (LIT) and an in-rock bentonite erosion test (i-BET) to assess the in-situ colloid-facilitated radionuclide transport through the bentonite erosion in the natural flow field. In the LIT experiment, radionuclide-containing compacted bentonite was equipped with a triple-packer system and then positioned at the borehole in the shear zone. It was observed that colloid transport was limited owing to the low swelling pressure and low hydraulic conductivity. Therefore, a postmortem analysis is being conducted to estimate the partial migration and diffusion of radionuclides. The i-BET experiment, that focuses more on bentonite erosion, was newly designed to assess colloid formation in another flow field. The i-BET experiment started with the placement of compacted bentonite rings in the double-packer system, and the hydraulic parameters and bentonite erosion have been monitored since December 2018.

When the radioactive nuclides are leaked from a deep geological repository by groundwater, the migration path of the nuclides is mostly consisted of rock fractures to the surface biosphere. Thus, assessing the safety of the disposed radioactive wastes depends upon understanding of nuclide migration in the fractured rocks. Fractures in rocks tend to dominate the hydrological characteristics of the dissolved nuclides. To study migration of nuclides in the rock fracture, a granite block of 1 m scale was quarried from the Hwangdeung site. The block has a single natural fracture. The six faces of the rock including fracture gaps were sealed with silicone adhesives to prevent leaking or diffusion of the water. Usually flow in fractured rock is unevenly distributed and most of the water flow occures over a small portion of the fracture zone, that is so called channeling flow. It is caused by uneven distribution of apertures in a fracture field. Flow rate is proportional to the cubic of the aperture. Thus, figuring out aperture distribution in a fracture field is the most important step on the study of the migration of nuclides in the fractured region. The ideal way to figure out the aperture distribution in a fractured rock is to use a non-destructive tool such as X-ray tomagraphe. However, it has a limitation of scale, that is, less than about 30 cm. It is not easy to give a good resolution for this quarried rock of 100×60×60 cm scale. It gives complex and vague images of the fracture. The optimum way to get an aperture distribution in a fractured rock is to drill some boreholes to the fracture and to carry out hydraulic tests. The more number of boreholes gives the more accurate information, but the more disturbance to the fracture field. Thus, it is necessary to optimize between aperture information and disturbing fracture field by selecting a suitable number of boreholes. We drilled nine boreholes from the upper surface of the rock mass just to the fracture without penetrating the fracture. And we carried out dipole tests for the matrix set of 9 boreholes. From each dipole test, an effective average aperture was calculated with the data of flow rate and hydraulic head. Then aperture distribution in the fracture field is calculated with a modified Krigging method. As a result, the aperture is distributed in the range of about 0.03~0.16 mm.

Bentonite, which mostly consists of montmorillonite, is considered as a suitable buffer material for disposal of high-level radioactive wastes in deep geological repository due to its high swelling capacity, low permeability, and strong retention capacity of radionuclide migration. Alkaline and saline solutions originated from degradation of cementitious material and seawater intrusion, respectively, may cause the changes in mineralogical and chemical properties of montmorillonite with various processes such as cation exchange within the interlayer, dissolution of montmorillonite, and precipitation of second minerals. In this study, montmorillonite alteration under alkaline and saline environments and its influences on retention of cesium and iodide by bentonite buffer were investigated. The reactions of bentonite (Bentonil-WRK) with alkaline solutions (0.1 M KOH and NaOH) and simulated saline solution were performed for 7 days in batch experiments at 25°C. After the experiments, reacted bentonite samples were characterized by X-ray diffraction (XRD), Fourier Transform Infrared (FTIR) spectroscopy, Short Wavelength Infrared (SWIR) spectrometry. The concentrations of cesium and iodide dissolved in the solutions were analyzed using an inductively coupled plasma mass spectrometer (ICP–MS). The XRD patterns showed significant decrease in the interlayer space of montmorillonite after the reaction with alkaline solution due to cation exchange and change in hydration status at the interlayer. The retention of cesium and iodide in alkaline and saline solutions were affected by montmorillonite alteration and ion competition. Therefore, the montmorillonite alteration affecting the nuclide retention capacity and long-term stability of bentonite buffer should be considered in the safety assessment of long-term geological disposal performance.

In this study, we evaluated fracture filling minerals and aperture distribution along the fracture surfaces under the controlled conditions. The fractured granite block which has a single natural fracture of 1 m scale was sampled in a domestic quarry (Iksan), which groundwater had been flowed through. This rock has an interconnected porosity of 0.36 with the specific gravity of 2.57. The experimental setup with the granite block with dimensions of 100×60×60 (cm). The fracture is sealed with rock silicone rubbers when it intersects the outer surfaces of the block and the outer surfaces are coated with the silicone to prevent loss of water by evaporation. Nine boreholes were drilled of orthogonal direction at the fracture surface. A flow of de-ionized water through the fracture between pairs of boreholes was initiated and the pressure required to maintain a steady flow was measured. In additions, fracture filling minerals were sampled and examined by mineralogical and chemical analyses. There are phyllosilicate minerals such as illite, kaolinite, and chlorite including calcite, which are fracture filling minerals. The illite and kaolinite usually coexist in the fracture, where their content ratio is different according to which mineral is predominant. For the evaluation of fracture, surface was divided into an imaginary matrix of 20×20 sub-squares as schematically. The calculated results are expressed as a two dimensional contour and a three dimensional surface plot for the aperture distribution in the fracture. The aperture value is distributed between 0.075 and 0.114 mm and the mean aperture value is 0.082 mm. The fracture volume is about 49 ml. These results will be very useful for the evaluation of environmental factor affecting the nuclides migration and retardation.

Bentonite is considered as buffer of engineered barrier for retardation of radionuclide migration. Bentonite has low permeability, high swelling and high sorption capacity for radioactive nuclides. Properties have been widely investigated under various geochemical conditions simulating deep geological environments. The chemical stability of bentonite is an important factor in evaluating the long-term stability of the bentonite buffer. However, the presence of impurities in bentonite clays can reduce the retention capacity for retardation of radionuclide migration value of bentonite. Therefore, the bentonite purification is necessary. In the present study, grade improvement of montmorillonite was conducted using ultrasonic and froth flotation methods. As a result of confirming the grade of montmorillonite according to the optimal ultrasonic intensity for ultrasonic irradiation is 1.0 kHz of bentonite in Gyeongju (KJ-II) increased from 60% to 78%. In case of froth flotation method using PSS (0.1 mM) as a reagent, the grade of montmorillonite increased up to 90%.

Numerous low-and intermediate level radioactive wastes were generated from the decommissioning processes of nuclear power plants. Radionuclides such as Co and Cs contained in decommissioning wastes should be immobilized to prevent the release of radionuclides from the wastes due to its harmful impacts on ecosystem by high radioactivity and long half-life. Ethylenediaminetetraacetic acid (EDTA) used as decontamination agent can be contained in cement waste during decommissioning process of nuclear power plants. In addition, EDTA can be stably and strongly bound with radionuclides, resulting in the acceleration of the nuclide release from solidified cement matrix. Here, we investigated the effects of EDTA on leaching behaviors of Co and Cs immobilized in the cement specimen. The leaching tests were performed according to the ANS 16.1 “Measurement of the leachability of solidified low-level radioactive wastes by a short-term test procedure”. From the results, an increase in the EDTA content in the cement specimen led to an increase in Co leaching, whereas a decrease in Cs leaching. Leaching of Cs was dominantly controlled by diffusion from the pore space of the cement specimen to the solution. The effective diffusion coefficient and leachability index of nuclide were determined using the diffusion-release models of ANS 16.1. The results of present study can be used in the safety assessment for disposal of the radioactive waste generated by decommissioning of nuclear power plants.

A deep geological disposal system, which consists of the engineered and natural barrier components, is the most proven and widely adopted concept for a permanent disposal of the high level radioactive waste (HLW) thus far. The clay-based engineered barrier is designed to not only absorb mechanical stress caused by the geological activities, but also prevent inflow of groundwater to canister and outflow of radionuclides by providing abundant sorption sites. The principal mineralogical constituent of the clay material is montmorillonite, which is a 2:1 phyllosilicate having two tetrahedral sheets of SiO2 sandwiching an octahedral sheet of Al2O3. The stacking of SiO2 and Al2O3 sheets form the layered structures, and ion-exchange and water uptake reactions occur in the interlayer space. In order to reliably assess the radionuclide retention capacity of engineered barrier under wide geochemical conditions relevant to the geological disposal environments, sorption mechanisms between montmorillonite and radionuclides should be explicitly investigated in advance. Thus far, sorption behavior of mineral adsorbents with radionuclides has been quantified by the sorption-desorption distribution coefficient (Kd), which is simply defined as the ratio of radionuclide concentration in the solid phase to that in the equilibrium solution; the Kd value is conditional, and there have been scientific efforts to develop geochemically robust bases for parameterizing the sorption phenomena more reliably. In this framework, application of thermodynamic sorption model (TSM), which is theoretically based on the concept of widely accepted equilibrium models for aquatic chemistry, offers the potential to improve confidence in demonstration of radionuclide sorption reactions on the mineral adsorbents. Specifically, it is generally regarded in the TSM that coordination of radionuclides on montmorillonite takes place at the surficial aluminol and silanol groups while their ion-exchange reactions occur in the interlayer space also. The effects of electrical charge on the surface reactions are additionally corrected in accordance with the numerous theories of electrochemical interface. The present work provides an overview of the current status of application of TSM for quantifying sorption behaviors of radionuclides on montmorillonite and experimental results for physical separation and characterization of Ca-montmorillonite from the newly adopted reference bentonite (Bentonil- WRK) by means of XRD, BET, FTIR, CEC measurement, and acid-base titration. The determined mineralogical and chemical properties of the montmorillonite obtained will be used as input parameters for further sorption studies of radionuclides with the Bentonil-WRK montmorillonite.