The radwaste repository consists of a multi-barrier, including natural and engineered barriers. The repository’s long-term safety is ensured by using the isolation and delay functions of the multi-barrier. Among them, natural barriers are difficult to artificially improve and have a long time scale. Therefore, in order to evaluate its performance, site characteristics should be investigated for a sufficient period using various analytical methods. Natural barriers are classified into lithological and structural characteristics and investigated. Structural factors such as fractures, faults, and joints are very important in a natural barrier because they can serve as a flow path for groundwater in performance evaluation. Considering the condition that the radioactive waste repository should be located in the deep part, the drill core is an important subject that can identify deep geological properties that could not be confirmed near the surface. However, in many previous studies, a unified method has not been used to define the boundaries of structural factors. Therefore, it is necessary to derive a method suitable for site characteristics by applying and comparing the boundary definition criteria of various structural factors to boreholes. This study utilized the 1,000 m deep AH-3 and DB-2 boreholes and the 500 m deep AH-1 and YS- 1 boreholes drilled around the KURT (KAERI Underground Research Tunnel) site. Methods applied to define the brittle structure boundary include comparing background levels of fracture and fracture density, excluding sections outside the zone of influence of deformation, and confining the zone to areas of concentrated deformation. All of these methods are analyzed along scanlines from the brittle structure. Deriving a site-specific method will contribute to reducing the uncertainties that may arise when analyzing the long-term evolution of brittle structures within natural barriers.

Advanced countries in the field of nuclear research and technology are currently examining the feasibility of deep geological disposal as the most appropriate method for the permanent management of high-level radioactive waste, with no intention of future retrieval. Deep geological disposal involves the placement of such waste deep underground within a stable geological formation, ensuring its permanent isolation from the human environment. To guarantee the enduring isolation and retardation of radionuclides with half-lives spanning tens of thousands to millions of years from the broader ecosystem, it is imperative to comprehend the long-term evolution of deep disposal systems, especially the role of natural barriers. These natural barriers, typically consisting of bedrock, encase the repository and undergo long-term evolutions due to tectonic movements and climate variations. For the effective disposal of high-level radioactive waste, a thorough assessment of the site’s long-term geological stability is essential. This necessitates a comprehensive understanding of its tectonic evolution and development characteristics, including susceptibility to seismic and magmatic events like earthquakes and intrusions. Furthermore, a detailed analysis of alterations in the hydrogeological and geochemical environment resulting from tectonic movements over extended time frames is required to assess the potential for the migration of radionuclides. In this paper, we have examined international evaluation methodologies employed to elucidate the predictive long-term evolution of natural barriers within disposal systems. We have extracted relevant methods from international case studies and applied a preliminary scenario illustrating the long-term evolution of the geological environment at the KURT (KAERI Underground Research Tunnel) site. Nevertheless, unlike international instances, the scarcity of quantitative data limits the depth of our interpretation. To present a dependable scenario in the future, it is imperative to develop predictive technologies aimed at comprehensively studying the geological evolution processes in the Korean peninsula, particularly within the context of radioactive waste disposal.

Kori-1 and Wolseong-1 nuclear power plants were permanently shut down in June 2017 and December 2019, and are currently in the preparation stage for decommissioning. In this regard, it is necessary to secure nuclear power plant decommissioning capacity in preparation for the domestic decommissioning marketplace. To address this, the Korea Research Institute of Decommissioning (KRID) was established to build a framework for the development of integrated nuclear decommissioning technology to support the nuclear decommissioning industry. The institute is currently under construction in the Busan-Ulsan border area, and a branch is planned to be established in the Gyeongju area. Recently, R&D projects have been launched to develop equipment for the demonstration and support verification of decommissioning technology. As part of the R&D project titled “Development and demonstration of the system for radioactivity measurement at the decommissioning site of a nuclear power plant”, we introduce the plan to develop a radioactivity measurement system at the decommissioning site and establish a demonstration system. The tasks include (1) measurement of soil radioactive contamination and classification system, (2) visualization system for massive dismantling of nuclear facilities, (3) automatic remote measurement equipment for surface contamination, and (4) bulk clearance verification equipment. The final goal is to develop a real-time measurement and classification system for contaminated soil at the decommissioning site, and to establish a demonstration system for nuclear power plant decommissioning. The KRID aims to contribute and support the technological independence and commercialization for domestic decommissioning sites remediation of nuclear power plant decommissioning site by establishing a field applicability evaluation system for the environmental remediation technology and equipment demonstration.

The deep geological disposal system is aimed to permanently isolate the high-level radioactive waste from the biosphere through a multi-barrier system composed of engineered and natural barriers. The buffer material used for the engineered barrier should have the performance to prevent and retard the migration of radionuclides to the outside of the deep disposal facility when radionuclides are released from a disposal canister by infiltration of groundwater after a long period of time. When the hydraulic conductivity of compacted bentonite is sufficiently low, the migration of radionuclides released into the surrounding rock can be significantly reduced since they are sorbed to bentonite during the diffusion process. Therefore, an investigation on diffusion of radionuclides in compacted bentonite is a fundamental task to obtain essential data for the safety assessment of the deep geological disposal system. The migration of radionuclides by diffusion can be evaluated by diffusion coefficient. In order to obtain the apparent diffusion coefficients of Sr, Sm, and Eu in compacted Ca-bentonite (1.6 g/cm3) a through-diffusion experiment have been carrying out. A cylindrical apparatus consists of a source cell with an appropriate concentration of radionuclides and diffusion cell filled with radionuclide free solution where the concentration is gradually increased by diffusion of radionuclides. The compacted bentonite was installed between the both cells. The sample used for the experiment is a Ca-type bentonite named Bentonil-WRK, and the diffusion experiment was performed under an oxidizing condition using a synthetic groundwater simulating KURT groundwater composition. The diffusion experiment will be terminated when an increasing rate of concentration of nuclides in the diffusion cell becomes constant over time. The concentration change with regard to the geochemical characteristic of the nuclide may appear to be apparently slow.In this study, the experimental results of the through-diffusion test of Sr, Sm, and Eu in the initial stage (~4 months) were presented. Through the results of the initial stage, the period of the through-diffusion experiment can be rearranged and also it is expected that the initial results provide the qualitative and quantitative diffusion properties of each nuclide.

The deep geological repository consisting of a multi-barrier system (engineered and natural barriers) is generally designed to isolate the high-level radioactive waste. The natural barrier is outermost portion to secure safety of the disposal. Crystalline rocks are considered for potential geological repository media to retard and inhibit the migration of radionuclides when the radionuclides leak from the canister and break through the engineered barrier. Sorption and diffusion processes play a major role in retardation of the radionuclides in deep underground environment. In order to evaluate the migration of radionuclides in the safety assessment or geochemical modelling, distribution coefficient and diffusivity of radionuclides are required as input data. In this study, we performed mineralogical and geochemical analysis for a crystalline rock (e.g., granite) to use the sorption and diffusion experiment. The fresh rock samples are obtained from a deep core samples (DB-2) drilled up to 1 km from the surface at KURT (KAERI Underground Research Tunnel) site. For the optical and microscopic examination, thin sections of the rock sample were provided. The rock samples were crushed into powder size to analyze major and trace elements of the whole-rock aliquots. The powdered specimens also used for mineral identification and measurement of specific surface area. The major constituent minerals of the granite are plagioclase, quartz, and K-feldspar and the minor minerals are phlogopite, biotite, and chlorite. According to the results of geochemical analysis, the granite specimens generally contain more than 70wt% of SiO2 and 8wt% of total alkali oxides (Na2O + K2O). The trace elements normalized to primitive mantle compositions show positive Cs, Rb, U, K, and Pb anomalies and negative Nb and Ti anomalies. The rock samples have an average density of 2.62 g·cm−3 and an average porosity of 0.222%. The crushed samples represent the specific surface area of 0.2087 m2·g−1 for the 75–150 μm fraction and 0.1616 m2·g−1 for the 150–300 μm fraction by BET method, respectively. The granite specimens will be used for the sorption and diffusion experiments to evaluate the radionuclides’ geochemical behaviors. The mineralogical and geochemical properties provided in this study can be useful in understanding the sorption and diffusion processes of significant radionuclides under the geological disposal environments.

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.