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.

The Colloid Formation and Migration (CFM) international joint research initiative continues as a part of the GTS’s Radionuclide Retardation Programme, which has been in progress since 1984. This project focuses on examining the formation of colloids from a bentonite-engineered barrier system and exploring how these colloids impact the migration of radionuclides in fractured host rock when subjected to advective flow. Phase 1 of the project was launched in 2004 and concluded in early 2008, focusing on preliminary studies related to in-situ boundary conditions, predicting models, and supplementary lab works. Following that, Phase 2 spanned from 2008 to 2013 and aimed at fortifying the field setup by adding three new monitoring boreholes and suitable instrumentation in both the boreholes and tunnel. This phase also tested the system’s resilience while mapping the flow domain. Phase 3 kicked off in January 2014 and extended until December 2018. During this period, the Long-term In-situ Test (LIT) was introduced in May 2014, featuring a set of compacted bentonite rings laced with radionuclide tracers. These were placed in a borehole to serve as a colloid and radionuclide source. CFM Phase 4 initiative commenced in January 2019, marking the successful deployment of the i-BET (In-situ Bentonite Erosion Test). This project component involves placing approximately 50 kg of compacted bentonite in a natural water-conducting shear zone. Korea Atomic Energy Research Institute (KAERI) joined CFM in 2008 to examine the behavior of colloid generation and migration with radionuclides in the Underground Research Laboratory. The fourth phase of the CFM project was also scheduled to include a post-mortem evaluation of the LIT and additional tracer experiments in the well-mapped MI shear zone. This study aims to provide an interim update on the ongoing i-BET, a key component of Phase 4 of the CFM project. We will also discuss the current status of the post-mortem analysis for the LIT experiment. In addition, we will outline plans for the forthcoming Phase VI of the project. These plans will continue to advance our understanding of radionuclide migration and the influence of bentonite-based disposal systems.

In the high-level waste disposal systems, colloids generated through the chemical erosion of bentonite buffers can serve as critical mediators for the transport of radionuclides from the disposal environment to the biosphere. The stability of these colloids is influenced by the chemical composition of the groundwater. According to DLVO theory, the Critical Coagulation Concentration (CCC) is the ionic strength at which the total repulsive force between colloids is either less than or equal to the total attractive force. At ionic strengths lower than the CCC, electrostatic double-layer repulsion outweighs van der Waals attraction, forming a repulsive barrier between particles. Conversely, at ionic strengths higher than the CCC, attractive forces dominate, leading to particle aggregation. To investigate the CCC of bentonite colloids, this study focused on Ca-type WRK bentonite. Colloids separated from a ten g/L bentonite suspension underwent centrifugation (1,200 g for 30 minutes) and dialysis (3,500 MWCO) to produce colloid samples. After adjusting the ionic strength from 0.1 mM to 10 mM, the particle size distribution was monitored as a function of aggregation time for approximately 20 days. Rate constants, calculated based on variations in ionic strength, were used to interpret the observed results. The experimental outcomes revealed that the CCC value for WRK bentonite colloids was an order of magnitude lower with CaCl2 than with NaCl. This suggests that Ca ions have a more significant impact on colloid stability, which has implications for the longterm safety of high-level waste disposal systems.

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 final disposal of Spent Nuclear Fuel (SNF) will take place in a deep geological repository. The metal canister surrounding the SNF is made of cast iron and copper, designed to provide longterm containment of radionuclides. Canister is intended to be safeguarded by a multiple-barrier disposal system comprising engineered and natural barriers. Colloids and gases are mediators that can accelerate radionuclide migration and influence radionuclide behavior when radionuclides leak from the canister at the end of its service life. It is very important to consider these factors in the assessment of the long-term stability of deep dispoal repository. An experimental setup was designed to observe the acceleration of nuclide behavior due to gas-mediated transport in a simulated environment with specific temperature and pressure conditions, similar to those of a deep disposal repository. In this study, experiments were conducted to simulate gas flow within an engineered barrier under conditions reflecting 1000 years post repository closure. The experiment utilized bentonite WRK with a dry density of 1.61 g/cm³ after compaction. The compacted bentonite was subsequently saturated under a water pressure of 5 MPa, equivalent to the hydrostatic pressure found 500 meters underground. Gas was introduced into the saturated bentonite at different pressures to assess the permeation behavior of the bentonite relative to gas pressure variations. Consequently, it was observed that under specific pressures, gas permeated the saturated bentonite, ascending in the form of bubbles. Furthermore, it was noted that when a continuous flow was initiated within the bentonite, erosion took place, leading to the buoyant transportation of eroded particles upward by the bubbles. The particles transported by the bubbles had a relatively small particle size distribution, and cesium also tended to be transported by the bubbles and moved upward. When high-pressure gas is generated at the interface of the canister and the buffer, flow through the buffer can occur, and cationic nuclides such as cesium and strontium can be attached to the gas bubble and migrate. However, the pressure of the gas to break through the saturated buffer is very high, and the amount of cesium transported by the gas bubbles is very limited.

The spent fuels derived from the nuclear reactor facilities may be finally disposed in a deep underground below 500 m. It majorly has uranium with minor iodine, which is a typical anionic radionuclide. In particular, radioiodine has higher mobility from its spent fuel source. It has been well known that it could freely pass through a compacted bentonite that is one of underground engineering barriers that are used to retard some nuclide’s migration from the spent fuel. We installed a small laboratory apparatus in an anaerobic glove box imitating such an underground repository and evaluated the iodine mobility in compacted bentonites with or without copper. Some copper-bearing bentonites were prepared in two types, a copper ion-exchanged form and a copper nanoparticle-mixed one. In our study, we tried to find an effect of sulfate that has an ability to retard mobile iodine from the compacted bentonite for a long-term period. Conclusively, we found an effective way to limit the iodine release from the compacted bentonite. This condition can be achievable by exchanging the bentonite interlayer cations with copper ions or by simply mixing copper nanoparticles with bentonite powder. In those cases, soluble iodine can be easily immobilized as a solid phase (i.e., marshite (CuI)) by combining with copper via the geochemical role of sulfate and indigenous SRB (sulfate reducing bacteria) of bentonite.

A disposal system for spent nuclear fuel mainly divides into two parts; Engineered barriers include spent nuclear fuel, canister, buffer and backfill and natural barriers mean a host rock surrounding engineered barriers. If radionuclides released from a repository, they can migrate to the ecosystem. Sorption plays an important role in retarding the migration of released radionuclides. Hence, the safety assessment for the disposal of a spent nuclear fuel should consider the migration and retardation of radionuclides in geosphere. Distribution coefficient is one of input parameters for the safety assessment. In this work, distribution coefficients for crystalline rock as a natural barrier were collected and evaluated for the purpose of safety assessment for the deep geological disposal of a spent nuclear fuel. The radionuclides considered in this work are as follows; alkali and alkaline earth metals (Cs, Sr, Ba), lanthanides (Sm), actinides (Ac, Am, Cm, Np, Pa, Pu, Th U), transition elements (Nb, Ni, Pd, Tc, Zr), and others (C, Cl, I, Rn, Se, Sn). The sorption of radionuclides is influenced by various geochemical conditions such as pH/carbonates, redox potential, ionic strength, radionuclide concentration, kinds and amounts of minerals, and microbes. For the evaluation of distribution coefficients, the data from Sweden (SKB), Finland (Posiva), Switzerland (Nagra), and Japan (JAEA) were collected, analyzed, and the recommended distribution coefficients have been suggested.

Raman characteristics of various minerals constituting natural rocks collected from uranium deposits in Okcheon metamorphic zone in Korea are presented. Micro-Raman spectra were measured using a confocal Raman microscope (Renishaw in Via Basis). The focal length of the spectrometer was 250 mm, and a 1800 lines/mm grating was installed. The outlet of the spectrometer was equipped with a CCD (1,024256 pixel) operating at -70°C. Three objective lenses were installed, and each magnification was 10, 50, and 100 times. The diameter of the laser beam passing through the objective lens and incident on the sample surface was approximately 2 m. The laser beam power at 532 nm was 1.6 mW on the sample surface. Raman signal scattered backward from the sample surface was transmitted to the spectrometer through the same objective lens. To accurately determine the Raman peak position of the sample, a Raman peak at 520.5 cm-1 measured on a silicon wafer was used as a reference position. Since quartz, calcite, and muscovite minerals are widely distributed throughout the rock, it is easy to observe with an optical microscope, so there is no difficulty in measuring the Raman spectrum. However, it is difficult to identify the uraninite scattered in micrometer sizes only with a Raman microscope. In this case, the location of uraninite was first confirmed using SEM-EDS, and then the sample was transferred to the Raman microscope to measure the Raman spectrum. In particular, a qualitative analysis of the oxidation and lattice conditions of natural uraninite was attempted by comparing the Raman properties of a micrometer-sized natural uraninite and a laboratory-synthesized UO2 pellet. Significantly different T2g/2LO Raman intensity ratio was observed in the two samples, which indicates that there are defects in the lattice structure of natural uraninite. In addition, no uranyl mineral phases were observed due to the deterioration of natural uraninite. This result suggests that the uranium deposit is maintained in a reduced state. Rutile is also scattered in micrometer-sizes, similar to uraninite. The Raman spectrum of rutile is similar in shape to that of uraninite, making them confused. The Raman spectral differences between these two minerals were compared in detail.

Bentonite is a promising buffer material for high-level radioactive waste (HLW) disposal due to the high nuclides sorption capacity and swelling property. However, bentonite has the potential to generate colloid particles, with small particle sizes less than 1,000 nm when in contact with groundwater. The bentonite colloids easily form pseudo-colloid with the released nuclides and migrate through the water-conducting rock to the biosphere. Therefore, understanding the generation and migration of bentonite colloids is crucial in assessing the safety of the HLW repository. In this study, an artificial fracture system was prepared to investigate colloid release from compacted bentonite. A 250 mm diameter acrylic artificial fracture system was used, with 30 mm of compacted calcium bentonite installed. Artificial groundwater flow was injected into the system at a flow rate of 250 μL/h, and every 6 mL of leachate was collected by a fraction collector. A film-type pressure sensor was equipped to monitor the swelling pressure, and the swelling was observed using a digital microscope. The results indicate that the compacted bentonite formed a mineral ring originating from the swelling of the bentonite, and the end of the ring generated colloid particles due to chemical erosion. Although the release rate of colloids increased with increasing flow rate, the colloid ratio depended on the low ionic strength of the injected artificial groundwater. This work contributes to the understanding of the chemical erosion and colloid release mechanism of compacted bentonite.

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 system for spent nuclear fuel divides into two parts; (1) engineered barriers including spent nuclear fuel, canister, buffer and backfill, (2) natural barriers surrounding engineered barriers. Sorption and diffusion are main retardation mechanisms for the migration of released radionuclides. We analyzed the sorption properties of radionuclides for bentonite as a buffer material and collected/ evaluated the distribution coefficients for the purpose of safety assessment for the deep geological disposal of a spent nuclear fuel. Through this, we presented recommended distribution coefficients for radionuclides required for the safety assessment. This work included the radionuclides as follows; alkali and alkaline earth metals (Cs, Sr, Ba), lanthanides (Sm), actinides (Ac, Am, Cm, Np, Pa, Pu, Th U), transition elements (Nb, Ni, Pd, Tc, Zr), and others (C, Cl, I, Rn, Se, Sn). The sorption of radionuclides affected various geochemical conditions such as pH/carbonates, redox potential, ionic strength, radionuclide concentration, kinds and amounts of minerals, and microbes. Among the evaluated radionuclides, Cs, Ni, Pd, and Ra is sensitive to the ionic strength, while Np, Pu, U, Se, and, Tc is sensitive to the redox condition. For the evaluation of distribution coefficients, the data from Sweden (SKB), Finland (Posiva), Switzerland (Nagra), and Japan (JAEA) were collected, analyzed, and the recommended distribution coefficients were suggested.

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.

The design of a radioactive waste disposal system should include both natural and engineered barriers to prevent radionuclide leakage and groundwater contamination. Colloids and gases can accelerate the movement of radionuclides and affect their behavior. It is important to consider these factors in the long-term stability evaluation of a deep geological repository. An experimental setup was designed to observe the acceleration of nuclide behavior caused by gas-mediated transport in a simulated high temperature and pressure environment, similar to a deep disposal repository. The study used specimens to simulate gas flow in engineered barriers, based on conditions 1000 years after repository closure. In the experiment, bentonite WRK with a dry density of 1.61 g/cm3 was used after compaction. Measurements were taken of the saturation time and gas permeability of compacted bentonite. In this study, gas was injected into saturated buffer materials at various pressures to evaluate the penetration phenomenon of the buffer material according to the gas pressure. It was observed that gas penetrated the buffer material and moved upward in the form of gas bubbles at a specific pressure. Furthermore, when a flow was continuously induced to penetrate the buffer material, erosion occurred, and the eroded particles were found to be able to float upward or be transported by gas bubbles. In future studies, analysis will be conducted on the transport rate of fine particles according to the size of gas bubbles and the characteristics of the nuclides adsorbed on the fine particles.

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.

Safe storage of spent nuclear fuel in deep underground repositories needs an understanding of the long-term alteration (corrosion) of metal canisters and buffer materials. We conducted a small-scale laboratory alteration tests on some metal (Cu and Fe) chips by embedding them into the compacted bentonite blocks, which were placed in anaerobic water for 1 year. Some additives like lactate, sulfate, and bacteria were separately loaded into the water to promote biochemical reactions. The bentonite blocks immersed in the water were finally dismantled after 1 year, and they showed that their alteration was insignificant. However, the Cu chip exhibited some microscopic etch pits on its surface, wherein sulfur component was slightly detected. Overall, the Fe chip was more corroded than the Cu chip under the same condition. The secondary phase of the Fe chip was locally found as carbonate materials, such as siderite (FeCO3) and calcite ((Ca, Fe)CO3). These secondary products could imply that the local carbonate production around the Fe chip may be initiated by an evolution (alteration) of bentonite and a diffusive provision of biogenic CO2 gas. These laboratory scale test results suggest that the long-term alteration (corrosion) of metal canister/bentonite blocks in the engineered barrier could be possible and may be promoted by microbial activities.