Disposal methods of radioactive waste can be mainly divided into near-surface disposal and deep geological disposal. If the radioactive waste is exposed to groundwater for a long time in the disposal environment, no matter how the decommissioning waste generated from the nuclear power plant is disposed of, the radionuclides may be released from the disposal site. Decommissioning waste from nuclear power plant contains radionuclides that are harmful to ecosystem including humans. Radionuclides released from disposal site behave in a complex and sensitive manner affected by geochemical conditions such as soil, geological media and groundwater. Sorption is one of the important mechanisms to retard the migration of radionuclides in a subsurface environment. In this study, geochemical properties of groundwater were collected and analyzed in the disposal environment considering disposal method in order to evaluate the geochemical behavior of radionuclides. The formation of aqueous and precipitated species of cesium and cobalt in a disposal condition were calculated and estimated. The sorption properties were also evaluated and predicted by considering the changes in the geochemical conditions such as pH, redox potential and geological media for the safety assessment.

Kori-1, the nuclear power plants in South Korea, first started operation in April 1978 and was suspended permanently in 2017. The saturation rate time of spent nuclear fuel generated by major nuclear power plants operating in Korea are getting closer. If we fail to dispose spent nuclear fuel, which is equivalent to high-level radioactive waste, the nuclear power plants will have to be shutdown. High-level radioactive waste is permanently disposed through a deep geological disposal system because it contains long-term half-life nuclides and emits high energy. To select the deep geological disposal site and construct the disposal facilities, it is necessary to establish appropriate regulatory policies accordingly. The status of database construction in OECD-NEA, NRC, SITEX, and IAEA, which provides safety regulations for deep geological disposal system, stipulates each requirement for dismantling nuclear power plants. However, details such as specific figures are not specified, and guidelines for the disposal of high-level radioactive wastes are not clearly distinguished. In Korea, the CYPRUS program, an integrated database system, has been developed to support comprehensive performance evaluation for high-level waste disposal. However, due to several difficult situations, maintenance and upgrades have not been performed, so the research results exist only in the form of raw data and the new research results have not been reflected. Other than that, there is no preemptive basis for regulating the deep geological disposal system. With real-time database, we can develop a regulatory system for the domestic deep disposal system by systematically analyzing the regulatory condition and regulatory case data of international organizations and foreign leading countries. The database system processed and stored primary data collected from nuclear safety reports and other related data. In addition, we used relational database and designed table to maximize time and space efficiency. It is provided in the form of a web service so that multiple users can easily find the data they want at the same time. Based on these technologies, this study established a database system by analyzing the legal systems, regulatory standards, and cases of major foreign leading countries such as Sweden, Finland, the United States, and Japan. This database aims to organize data for each safety case component and further prepare a safety regulatory framework for each stage of development of disposal facilities suitable for the domestic environment.

The backfill refills the deep geological disposal system after the installation of buffer in the disposal hole. SKB and Posiva have established the safety function for the backfill such as hydraulic conductivity of 10-10 m/s and swelling pressure of 0.2 MPa. The study on the thermal properties is required for the evaluation of performance design and long-term stability of backfill, since the thermal condition affects the hydraulic and mechanical behavior of backfill. Thermal conductivity is a key characteristic of thermal properties due to heat dissipation from spent fuel. In this study, thermal conductivities of bentonite-sand mixed blocks were measured. The silica sands were used instead of the crushed rock with bentonil-WRK, one of the candidate bentonite of the Korean repository system. The effects of size distribution and mass ratio of sand were evaluated. Four different size of silica sand (i.e., 0.18-0.25, 0.7-1.12, 1.6-2.5, 2.5-5.0 mm) and five mixing ratio (i.e., 1:9, 2:8, 3:7, 4:6, 5:5 of bentonite and sand) were used for characterization of thermal conductivity. As a result, the thermal conductivities were measured ranging from 1.6 to 3.1 W/m∙K depending on the size and mass ratio of the sand. The smaller the size or higher the mixing ratio of sand or the higher the water contents, the higher the thermal conductivity on the surface of backfill block. The higher compressing pressure induce higher thermal conductivity. Meanwhile, the feasibility study of backfill block productivity was reviewed according to the variables of this study. The excessive sand ratio and water contents lead to poor quality that results in the failure of the block. In Korea, the research of backfill is only now in fundamental steps, thus the results of this study are expected to use for setup the experimental conditions of hydraulic and mechanical performance, and can be used for the design of safety function and evaluation of long-term stability for deep geological disposal system.

When a rapid groundwater inflow is introduced from the adjacent rock mass in the early stage of disposal, hydraulic pressure build-up occurs, which may cause piping erosion at the buffer material itself and the interface of the gap-filling material. Such piping erosion in compacted bentonite buffer via interaction between the buffer and the adjacent rock mass may deteriorate the performance of the buffer material. Therefore, it is necessary to understand the conditions and scenarios in which the piping phenomenon around the buffer material occurs for the long-term health of the repository. In this study, laboratory-scale experimental tests of piping erosion in buffer and interfacial rock was introduced. ø 100 mm × 200 mm height compacted bentonite specimens were placed in a cylindrical acetal cell, and the distilled water was continuously injected at a flow rate of 0.068 L/min using a dual syringe pump. The inflow of water was generated from the bottom and side cell of buffer material. During water injection, injected water pressure and amount were measured with visual observation. The results showed that the external saturation of buffer firstly occurs followed by piping crack generation along the wetting front. The additional piping channels were generated and merged with others. As the injection stopped, the swelling and self-sealing behavior of buffer material were observed. Moreover, X-ray CT scanning of the cell was conducted after the piping simulation to analyze the piping channels and saturation depth. The results highlight the piping erosion phenomenon mainly occurs due to the presence of a gap outside the buffer material. Further experimental cases is need to comprehensively understand piping phenomena in buffer material for assessing the long-term stability of underground radioactive waste disposal systems.

According to the continued generation of spent nuclear fuel, a reliable safety assessment is highly required with the precise modeling of the migration and retardation behavior of radionuclides to enhance public acceptance and hinder excessive conservativeness during the construction of the repository. In particular, the colloids formed in the repository-relevant condition are known to accelerate the migration of radionuclides. Thus, geochemical behavior and relevant characteristics of colloids are needed to be unambiguously clarified. The objective of the present work is to investigate the fundamental characteristics of colloids contained in the natural groundwater system by using various analytical methods and the tangential flow ultra-filtration (TFUF) system. The granitic groundwater sample from the DB-3 borehole at the KURT (KAERI Underground Research Tunnel) was taken by an airtight stainless steel cylinder coated on the inside with PTFE to prevent the infiltration of ambient air into the geologic groundwater sample. And then, the groundwater sample was transferred to the inert glovebox filled with Ar gas to monitor the pH and Eh equilibrium of the aqueous sample. For further investigation, the colloid contained in the groundwater sample was concentrated by using the TFUF system equipped with a membrane filter (pore size: 3 kDa). The concentrated groundwater sample was analyzed with various methods such as ICP-MS/OES, IC, DLS/ELS, FE-TEM/SEM-EDS, ATR-FTIR, TOC, LC-OCD, etc. In this study, the size of groundwater colloids was determined to be 182.3 ± 52.7 nm with the major constituents of C, S, O, Fe, Al, Si, etc. The amount of organic carbon and the concentrations of organic substances determined by means of the molecular weight fraction with the TOC and LC-OCD provide further detailed information for the colloids in the KURT groundwater sample. The results obtained in this study are expected to be used as preliminary experimental data for modeling the colloid-facilitated migration of radionuclides to improve the reliability of the safety assessment of the geologic repository.

A rock joint exerts significant influences on the rock mass behavior in terms of thermal, hydraulic, and mechanical (THM) aspects. Therefore, its features should be thoroughly investigated in various rock mechanical projects, such as high-level radioactive waste (HLW) disposal repository, tunnel, and rock slope. Meanwhile, it is essential to guarantee the safety of the disposal repository for a very long period of time and it should prepare measures for various risks, which may possibly encounter during that period. In general, direct shear tests for a rock joint are conducted to investigate the possibility of frictional sliding of the joint under specific loading conditions or to predict the shear strength of the joint. However, it is necessary to consider whether regional sliding of a rock joint or reactivation of a fault might occur due to an earthquake or redistribution of the in-situ stresses because the expected operation period of the repository is quite long, and various situations can happen. A slide-hold-slide test for a rock joint is a practical test that can investigate the time-dependent behavior or frictionalhealing of a joint. The test enables an estimation of the stress build-up phenomenon after strain energy release in a quantitative manner. In this study, a series of slide-hold-slide tests were carried out in order to investigate the characteristics. Joint specimens were made from mortar, which is a rock-like and brittle material, so as to consider the effect of joint roughness and to secure the reproducibility of the tests. At the same time, mechanical conditions as well as thermal and hydraulic were applied in order to take the environment of the repository into account. As a result, the behavior of shear stress recovery was observed, and the effects of THM coupled condition on the recovery were investigated. This study presents fundamental results of the experiments, and further research outcomes, including time dependent behavior of a joint, will be presented sequentially.

For the geological risk assessment of deep-depth underground condition by excavation work or tunneling, since rocks and geologic structure of each country is different, it is necessary to objectify or classify quantitatively deep-depth underground risk evaluation in accordance with Korean geologic characteristics. It could be summarized major factors of rock failure and underground space deformation by geological and geotechnical features as geologic structures, overburden, rock mass characteristics, groundwater, high stress and additional categories. Induced main factors that could be identified and predicted intermediate to deep-depth underground risk through literature investigation and analysis study on research trend related to the underground geological engineering. In order to assessment the risk of rock mass excavated from 100 m or more to several kilometers deep below the ground are classified into about 19 factors, and can be divided into 6 categories. Using these risk factors as basic data, weights for each factor for each category can be set, and further, the risk of excavated rock mass can be calculated.

The buffer block, which is one of the main components of the engineering barrier system, plays an essential role in mitigating groundwater infiltration and radionuclide transport in a high-level nuclear waste repository. To achieve those purposes, the compacted buffer block must satisfy the functional safety criteria for dry density, water content, and many other components. In this study, the compation curves of the compacted bentonite-sand mixtures were evaluated to identify the relationship between the dry density and the water content of the buffer material. The floating die press at 10 MPa and the cold isostatic press at 40 MPa were applied to compaction of a buffer block with a diameter of 100 mm and a thickness of 10 mm. The condition of a bentonite-sand mixing ratio was 6:4, 7:3, 8:2, and 9:1 with 9 to 21% water content. As a result, the maximum dry density increases, the optimum moisture content decreases as the sand content of buffer material increases. This study can provide the conditions for manufacturing the compacted bentonite-sand buffer block.

The change of surface environments (e.g., climate change, uplift/subsidence, and erosion) can undermine the long-term safety of a high-level radioactive waste repository. Therefore, understanding the water cycle between atmosphere, surface, and subsurface is essential to ensure the long-term safety of deep geological disposal and consequently to gain public acceptance for the repository. Among hydrologic components (e.g., precipitation, interception, runoff, infiltration, evapotranspiration (ET), and recharge) which constitute the water cycle, ET is more than half of the total precipitation and plays a crucial role in the water and energy transfer among the three systems. Although various methods for ET evaluation (e.g., Bowen Ratio, Eddy Covariance, Optical Scintillation, and Weighing Lysimeter methods) have been developed, many influential factors such as vegetation, climate, and moisture content make its accurate evaluation still tricky. In this work, we chose weighing lysimeter and Penman-Monteith methods for direct/indirect estimation of ET, and installed a smart field lysimeter and a micro-meteorological station around KAERI Underground Research Tunnel. Water balance in the unsaturated zone and five climatic variables (air temperature, humidity, precipitation, radiation, and wind speed/direction) were measured more than once per 10 minutes for six months from April to September, 2022. From the measurements, daily actual and potential ET values at the study site were calculated and compared. We also discussed the applicability and limitation of current methods and ET assessments at different spatial scales regarding verifying and validating the developing numerical models.

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.

The analysis of uranium migration is crucial for the accurate safety assessment of high-level radioactive waste (HLW) repository. Previous studies showed that the migration of the uranium can be affected by various physical and chemical processes, such as groundwater flow, heat transfer, sorption/ desorption and, precipitation/dissolution. Therefore, a coupled Thermal-Hydrological-Chemical (THC) model is required to accurately simulate the uranium migration near the HLW repository. In this study, COMSOL-PHREEQC coupled model was used to simulate the uranium migration. In the model, groundwater flow, heat transfer, and non-reactive solute transport were calculated by COMSOL, and geo-chemical reaction was calculated by PHREEQC. Sorption was primarily considered as geo-chemical reaction in the model, using the concept of two-site protolysis nonelctrostatic surface complexation and cation exchange (2 SP NE SC/CE). A modified operator splitting method was used to couple the results of COMSOL and PHREEQC. Three benchmarks were done to assess the accuracy of the model: 1) 1D transport and cation exchange model, 2) cesium transport in the column experiment done by Steefel et al. (2002), and 3) the batch sorption experiment done by Fernandes et al. (2012), and Bradbury and Baeyens (2009). Three benchmark results showed reliable matching with results from the previous studies. After the validation, uranium 1D transport simulation on arbitrary porewater condition was conducted. From the results, the evolution of the uranium front with sequentially saturating sites was observed. Due to the limitation of operator splitting method, time step effect was observed, which caused the uranium to sorbed at further sites then it should. For further study, 3 main tasks were proposed. First, precipitation/ dissolution will be added to the reaction part. Second, multiphase flow will be considered instead of single phase Darcy flow. Last, the effect of redox potential will be considered.

In the geological disposal system whose host rock is crystalline rock, fractures play a significant role in the safety assessment as they are the main pathway of the radionuclide migration. From the perspective of long-term safety assessment, the properties of fractures can be changed by tectonic movement such as earthquake, uplift, etc. In general, methods for simulating fractures include Discrete Fracture Network (DFN), which directly simulates the fracture surface, and Equivalent Continuous Porous Media (ECPM), which is equivalent to the ratio of the fractures in a certain rock volume. DFN is generally appropriate for deterministic fractures with large scale and high flow velocity, but ECPM may be more appropriate for small scale and sporadically distributed stochastic fractures because the flow velocity is slow and thus the rock matrix diffusion needs to be considered. In fact, several commercial software, such as FracMan, are already in use to convert DFN to ECPM. However, in order to consider the change in properties of fractures due to tectonic movement in the long-term safety assessment, a model that converts DFN to ECPM needs to be modularized and embedded into the safety assessment model. In this study, therefore, an in-house MATLAB code was developed to convert DFN to ECPM, which can be used as a submodule. The algorithm of converting from DFN to ECPM basically followed the Oda’s method. As the first step of the algorithm, in order to obtain the volume ratio of the fracture in a certain mesh element, the cross-sectional area of the fracture and the mesh element was calculated. Then, porosities of each mesh element were calculated as the volume fraction of fractures passing through the mesh element. Based on the Oda’s method, the permeability tensors of each mesh element were calculated by using an empirical fracture tensor which is weighted by the cross-sectional area and transmissivity of each fracture. Finally, the newly developed module was verified by a benchmark test, in which the ECPM results converted from a certain DFN data by using the numerical module developed in this study were compared with those by using FracMan. The newly developed module will be installed in the process-based total system performance assessment framework (APro) being developed by KAERI.

Currently, the most widely accepted disposal concept for long-term isolation of high level radioactive waste including spent nuclear fuels is to disposal in a deep geological repository designed and constructed with multiple barriers composed of engineered and natural barriers so that the waste can be completely isolated in a stable deep geological environment. In this concept, an important consideration is the heat generated from the waste due to the large amount of fission products present in the high level waste loaded in the disposal container. For safe and complete isolation of high level radioactive waste in the deep geology, the disposal concepts that meet the thermal requirements for the disposal system design have been developed by harmonizing the thermal characteristics of engineered and natural barriers in Korea. In this paper, the deposition hole configuration and the decay heat dissipation area (surface area) of disposal container were considered for the efficient thermal management in the deep geological disposal concept. Heat transfer through the waste form, its container and surrounding components and the rock will be mainly by conduction. Heat transfer by radiation and convection can be negligible after backfilling. When considering heat conduction, according to Fourier’s law, if the thermal conductivity of the repository components is the same, the greater the heat dissipation area and the adjacent temperature gradient, the greater the conduction effect. Therefore, rather than the conventional concept of loading 4 PWR spent fuel assemblies per disposal container and placing one disposal container in a deposition hole, it is better to load one assembly per disposal container and place 4 disposal containers in a deposition hole. In this case, it was found that the disposal area could be reduced through efficient thermal management. Considering this thermal management method as an alternative to the concept of deep geological disposal, additional research is needed.

The underground environment has an advantage to minimize the external influences because it is isolated space with surrounded rock medium. Therefore, underground rock has been used recently as the target for a disposal system of spent fuel with high-level radioactive. The disposal system mainly consists of natural barrier (i.e., surrounded rock medium) and engineered barrier (i.e., concrete lining, plug, backfill, canister, and buffer). In particular, the engineered barrier is important for long-term storage because it has to preferentially block the leakage of radioactive nuclide. Non-destructive technologies (NDT) have been utilized to monitor the state of disposal system for considering the limitation in deep depth conditions such as limited environment for direct damage inspection. Acoustic emission (AE) monitoring technique is an effective method to monitor the damage (crack) magnitude, history (i.e., crack evolution), and location using high-frequency elastic waves. To apply the AE monitoring method in the disposal system, the characteristics of damaged materials should be considered. The concrete lining has multi-failure behavior (i.e., brittle and ductile) resulted from composition as cement and reinforcing steel bar. Therefore, it important to investigate the AE characteristics according to the failure level of reinforced concrete for damage monitoring of the disposal systems. In this study, the four-point bending tests were carried out to measure the AE signals from the cracking of reinforce concrete specimens in laboratory. The test specimens were prepared with different strength. After the experiment, the AE characteristics were analyzed using the AE parameters with loading and failure state in the curve of time-stress. This study will be helpful for damage monitoring using AE technique in the field of high-level radioactive disposal system.

Compacted bentonite buffer materials are a key component of the engineered barrier system for high-level radioactive waste disposal. The bentonite buffer is saturated via groundwater flow through the excavation damaged zone in the adjacent rock mass. Bentonite saturation results in bentonite swelling, gelation and intrusion into the nearby rock discontinuities. Groundwater flow can cause bentonite erosion and transportation of bentonite colloids. This bentonite mass loss can negatively impact the long-term integrity of the engineered barrier system. Hence, it is necessary to understand the effects of erosion on the properties of the bentonite buffer. In this study, a series of artificial fracture erosion experiments are conducted to investigate the erosion characteristics of compacted Ca-bentonite buffer materials for different initial dry density conditions. Compacted bentonite blocks and bentonite pellets were manufactured using the cold isostatic pressing technique and granulation compactor respectively. The specimens were placed in a custommade transparent artificial fracture cell and the bentonite intrusion characteristics were monitored for two months under free swelling conditions with no groundwater flow. The radial expansion of the bentonite specimens within the artificial fracture was measured using a digital camera. In addition, the swelling pressure, displacement, and saturation were determined using a load cell-piston system, LVDT, and electrical resistivity electrodes respectively. A hydro-mechanical-chemical coupled dynamic bentonite diffusion model was applied to model the bentonite erosion characteristics using COMSOL Multiphysics.

The reliable information on the hydraulic characteristics of rock mass is one of the key site factors for design and construction of deep subsurface structures such as geological radioactive nuclear waste disposal repository, underground energy storage facility, underground research laboratory, etc. In order to avoid relying on foreign field test technology in future projects, we have independently designed and made integrated type main frame, 120 bar waterproof downhole sonde, and 1,200 m wireline cable winch through a series of R&D activities. They are core apparatuses of the Deep borehole Hydraulic Test System (DHTS). Integration of individual test equipment into a single main frame allows safe and efficient work in the harsh field condition. The DHTS was developed aiming primarily for constant pressure (head) injection test and pulse test in deep impermeable rock mass. The maximum testing depth of the DHTS is about 1,050 m from the surface. Using this system, it is possible to make precise stepwise control of downhole net injection pressure in less than 2.0 kgf/cm2 with dual hydraulic volume controller and also to inject and measure the very low flow rate below 0.01 l/min with micro flow rate injection/control module. Over the past two years, we have successfully completed more than 50 in situ hydraulic tests at 5 deep boreholes located in the Mesozoic granite and sedimentary rock regions in Korea. Among them, the deepest testing depth was more than 920 m. In this paper, the major characteristics of the DHTS are introduced and also some results obtained from the high precision field tests in the deep and low permeable rock mass environment are briefly discussed.

Filtering nuclides in high-level nuclear waste using rotating plasmas is a physical separation method based on mass difference. Since it is not chemical separation or metallurgical separation, the elements are separated regardless of their chemical composition. Accordingly, the more the number of chemical elements present in the waste or the more difficult when using the differences in solubility, the more advantageous. However, to be economically competitive, new concepts for rotating plasmas are needed to improve the poor separation rates despite high energy and installation costs. In this work, we revisited a counter rotating plasma centrifuge to examine its potential as an effective device for separating nuclear waste efficiently and economically.

Many countries have been developing their own FEP (Feature, Event, Process) lists to formulate radionuclide release scenarios in deep disposal repository of spent nuclear fuels and to assess the safety. The main issue in developing a FEP list is to ensure its completeness and comprehensiveness in examining all plausible scenarios of radionuclide release in a repository of interest. To this end, the NEA International FEP (IFEP) list as a generic reference have been developed and updated through long-term international collaborations. Leading countries advanced in the research field of deep geologic disposal of spent nuclear fuels have comparatively mapped their project-specific FEP (PFEP) lists with the IFEP list. Recently in 2019, NEA has published an updated version of IFEP list (ver. 3.0) which has a different classification system: the IFEP version 3.0 has the five main categories including the waste package, repository, geosphere, biosphere and external factors while the previous IFEP versions were mainly classified into the external, environmental, and contaminant factors. Most leading countries in this field, Finland and Sweden, recently succeeded to obtain the design and/or construction licenses for deep geologic disposal of spent nuclear fuel. Therefore, their PFEP lists should be good benchmark cases to the following countries. However, their PFEP lists have not comparatively mapped with the most recent version of IFEP and thus some gaps may exist in showing completeness and comprehensiveness in comparison to the IFEP version 3.0. In this study, we comparatively map the PFEP lists of Finland and Sweden to the IFEP version 3.0. The comparatively mapped PFEP list could be used as the basis for verifying the comprehensiveness and completeness of the domestic PFEP list currently under development in Korea.

Despite the increasing interest in Deep Borehole Disposal (DBD) for its capability of minimizing disposal area, detailed research about DBD operation system design should be conducted before the DBD can be implemented. Recently, DBD operation system applying wireline emplacement (WE) technique is under study due to its high flexibility and capability of minimizing surface equipment. In this study, a conceptual WE system, and operation procdure is introduced. The conceptual WE system consists of 3 main stations, which from the top are hoisting station (HS), canister connection station (CCS) and basement (BS). In HS, WE is controlled and monitored. The WE is controlled using wireline drum winch and sheaves, and load on wireline is measured using a load cell. HS also has a pressure control system (PCS), which monitors internal pressure of the system, and a lubricator, which act as housing for joint device, allowing the joint device to be easily inserted into the borehole. The joint device is used to connect the disposal canister to wireline for emplacement/retrieval. In CCS, a rail transporter brings a transport cask containing disposal canisters, then the transport cask is connected to the hoisting system and a PCS in the BS. The main component located at canister station are a sliding shielding door (SSD), and a slip. The SSD is used to prevent canister from falling into borehole during the connecting operation and prevent radiation from BS to affect the workers. The slip is located beneath the SSD and is used to hold the disposal canister before it is lowered into the borehole. In BS, PCS is installed to prevent overflow and blowout of borehole fluid. The PCS consists of wireline pressure valve, christmas tree and BOP, which all are a type of pressure valve to seal the borehole and release pressure inside the borehole. The WE procedure starts with transporting transport cask to CCS. The transport cask is connected to lubricator, and PCS. Joint device is lowered down to be connected with disposal canisters, then pulled up to check the load on the wireline. After the check-up, SSD is opened, and disposal canister is lowered into the borehole. When desired depth is reached, joint device is disconnected and retrieved for next emplacement. In this study, the conceptual deep borehole disposal system design implementing WE technique is introduced. Based on this study, further detailed design could be derived in future, and feasibility could be tested.

Copper is used for deep geological disposal canisters of spent nuclear fuels, because of excellent corrosion resistance in an oxygen-free environment. However, sulfide formation during the long-term exposure under deep geological disposal condition can be harmful for the integrity of copper canisters. Sulfur around the canisters can diffuse along grain boundaries of copper, causing grain boundary embrittlement by the formation of copper sulfides at the grain boundaries. The development of copper alloys preventing the formation of copper sulfides along grain boundaries is essential for the longterm safety of copper canisters. In this research, the mechanisms of copper sulfide formation at the grain boundary are identified, and possible alloying elements to prevent the copper sulfide formation are searched through the first principle calculations of solute atom-vacancy binding energy and the molecular dynamics calculation of grain boundary segregation energy. The comparison with the experimental literature results on the mitigation of copper embrittlement confirmed that the theoretically identified mechanisms of copper sulfide formation and the selected alloy elements are valid. Thereafter, binary copper alloys were prepared by using a vacuum arc melting furnace. Sulfur was added during casting of the copper alloys to induce the sulfide formation. The cast alloys were cold-rolled into a plate after homogenization heat treatment. The microstructure and mechanical property of each alloy were investigated after recrystallization in a vacuum tube heat treatment furnace. The copper alloys developed in this study are expected to contribute in increasing the long-term safety of deep geological disposal copper canisters by reducing the embrittlement caused by the sulfide formation.