In order to dispose of spent nuclear fuel (SNF) in deep geological repository, source term evaluation considering its specification, enrichment, burnup, cooling time should be performed. In this study, the measured values of Takahama-3 pressurized water reactor SNF (WH 17×17) samples were analyzed with SCALE 6.1/ORIGEN-S and TRITON code calculation results for validation. Unlike the ORIGENS code, TRITON code calculations differed from two-dimensional neutron flux distribution by using the multi-group cross-section library. Both calculation results from ORIGEN-S and TRITON code showed higher errors in 234U, 239Pu, and 241Pu compared to other actinide nuclides. In the case of axial locations of fuel rods in fuel assembly, fuel rods located at the edge of the fuel assembly presented increased errors due to nuclear reaction cross-section. Overall, the ORIGEN-S predictions informed more accurate agreement with the measured results compared with TRITON results. Especially to 235U, 239Pu, and 240Pu radionuclides, ORIGEN-S errors were denoted more than twice as low as the TRITON results. Comparing the calculation results with experimental results implied that the ORIGENS code was more accurate code than the TRITON code for source term evaluation.

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.

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.

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.

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.

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.

To minimize the short-term thermal load on the repository facility, heat generating nuclides such as Cs-137 and Sr-90 should be separated from the spent nuclear fuel for efficiency of repository facility. In particular, Sr-90 must be separated because it generates high heat during the decay process. Recently, Korea Atomic Energy Research Institute (KEARI) is developing a waste burden minimization technology to reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides such as Cs, Sr, I, TRU/RE, and Tc/Se from spent nuclear fuel. Among the major nuclides, Sr nuclides dissolve in chloride phase during the chlorination process of spent nuclear fuel and recovered in the form of carbonate or oxide via reactive distillation. In this process, Ba nuclides are also recovered along with Sr nuclides due to their chemical similarity. In this study, we prepared group II nuclide ceramic waste form, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method by considering the various ratio of Sr/Ba nuclides generated from nuclide management process. The established waste form fabrication process was able to produce a stable waste form regardless of the ratio of Sr/Ba nuclides. To evaluate the stability of group II waste form, physicochemical properties such as leaching and thermal properties were evaluated. Also, the radiological properties of the Ba(x)Sr(1-x)TiO3 waste forms with various Sr/Ba ratios were evaluated, and the estimation of centerline temperature was carried out using the experimental thermal property data. These results provided fundamental data for long-term storage and management of group II nuclides waste form.

Three government bodies, that is, the Ministry of Science and ICT (MSIT), Ministry of Trade, Industry, and Energy (MOTIE), and Nuclear Safety and Security (NSSC), jointly established the Institute for Korea Spent Nuclear Fuel (iKSNF) in December 2020 to secure the management technologies for spent nuclear fuel (SNF). The objective of iKSNF is to successfully conduct the long-term research and development program of the 「Development of Core Technologies to Ensure Safety of Spent Nuclear Fuel Storage and Disposal System」. Our program, known as the first multi-ministry program in the nuclear field of Korea, mainly focuses on developing core technologies required for the long-term management of SNF, including those for safe storage and deep geological disposal of SNF. The program comprises three subprograms and seven key projects covering the storage, disposal, and regulatory sectors of SNF management. Our program will last from 2021 through 2029, with a budget of approximately four billion USD sponsored by MSIT, MOTIE, and NSSC.

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as 129I and 135Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM−1.

경주 방폐물 처분시설의 1단계 시설로 건설된 지하 사일로 구조는 2014년에 10만 드럼 규모로 완공되어 현재 운영중에 있다. 지하 사일로 구조는 지름 25m, 높이 50m로써 방폐물을 저장하는 실린더부분과 돔 부분으로 구성되어 있으며, 돔부분은 운영터널과 연결 되는 하부 돔 부분과 상부 돔 부분으로 구분할 수 있다. 지하 사일로 구조의 벽체는 철근콘크리트 라이너이고, 두께는 약 1m이다. 본 논문에서는 지하 사일로 구조의 건설과정 및 운영과정의 단계별 유한요소해석을 수행하였다. SMAP-3D 프로그램을 사용하여 2차원 축대칭 유한요소해석을 수행하였다. 2차원 축대칭 유한요소모델의 신뢰성을 검토하고자 3차원 유한요소해석도 수행하였다. 본 논문 에서는 지하 사일로 구조의 구조거동을 분석하고 구조적 안전성을 검토결과를 제시하였다.

In Malaysia, there are several industries processing mineral ores generate residues containing naturally occurring radioactive material (NORM) with activity concentrations above the control limits established by the Malaysian Atomic Energy Licensing Board (AELB). These industries use mineral ores or concentrated ores as their feed materials to produce or extract valuable sand minerals or rare earth compounds for use in another industries. The control limits for activity concentrations of Uranium-238 (U-238) and Thorium-232 (Th-232) and their decay series is 1.0 Becquerel per gram (Bq·g−1) while activity concentration of Potassium 40 (K-40) is 10.0 Bq·g−1. The management of residue containing NORM radioactivity above the control limits must be done in accordance with current rules and regulations including proper handling, storage, transportation and/or disposal. Where possible, appropriate mixture process with other non-radiological material would reduce the activity concentrations to below the control limits. Depending on specific characteristics of residue, appropriate approach to reuse or recycle should be encouraged as part of special waste management. For this case, an exemption to release it from radiological controls can be applied but require scrutiny review and approval process by AELB. In addition, the health and safety aspects and environmental issues should be assessed which to be done in accordance with the relevant rules and regulations. As a last resort, a disposal of residue containing NORM radioactivity shall be done at the landfill disposal facility approved by AELB and other relevant Authorities.

The spent filters stored in Kori Unit 1 are planned that compressed and disposed for volume reduction. However, shielding reinforcement is required to package high-dose spent filters in a 200 L drum. So, in this study suggests a shielding thickness that can satisfy the surface dose criteria of 10 mSv·h−1 when packaging several compressed spent filters into 200 L drums, and the number of drums required for the compressed spent filter packaging was calculated. In this study, representative gamma-emitting nuclides in spent filter are assumed that Co-60 and Cs-137, and dose reduction due to half-life is not considered, because the date of occurrence and nuclide information of the stored spent filter are not accurate. The shielding material is assumed to be concrete, and the thickness of the shielding is assumed to 18 cm considering the diameter of the spent filter and compression mold. Considering the height of the compressed spent filter and the internal height of the shielding drum, assuming the placement of the compressed spent filter in the drum in the vertical direction only, the maximum number of packaging of the compressed spent filter is 3. When applying a 18 cm thick concrete shield, the maximum dose of the spent filter can packaged in the drum is 125 mSv·h−1, so when packaging 3 spent filters of the same dose, the dose of a spent filter shall not exceed 41 mSv·h−1 and not exceed 62 mSv·h−1when packing 2 spent filters. Therefore, the dose ranges of spent filters that can be packaged in a drum are classified into three groups: 0–41 mSv·h−1, 41–62 mSv·h−1, and 62–125 mSv·h−1based on 41 mSv·h−1, 62 mSv·h−1, and 125 mSv·h−1. When 227 spent filters stored in the filter room are classified according to the above dose group, 207, 3 and 4 spent filters are distributed in each group, and the number of shielding drums required to pack the appropriate number of spent filters in each dose group is 75. Meanwhile, 8 spent filters exceeding 125 mSv·h−1 and 5 spent filters that has without dose information are excluded from compression and packaging until the treatment and disposal method are prepared. In the future, we will segmentation of waste filter dose groups through the consideration of dose reduction and horizontal placement of compressed spent filters, and derive the minimum number of drums required for compressed spent filter packaging.

Korea Research Reactor 1&2 (KRR-1&2), Korea’s first research reactor, began dismantling in 1997. As of 2022, the demolition of general areas such as offices has been completed, and contaminated areas such as reactor rooms remain. On the other hand, construction waste generated in contaminated areas of nuclear facilities cannot be disposed of as general industrial waste. It is predicted that about 5,000 tons of construction waste will be generated if the contaminated area of KRR-1&2 is demolished. In this study, the application plan for the demolition of contaminated area of KRR-1&2 was reviewed through a review of laws and cases related to domestic and overseas disposal. The only method for disposing of construction waste in contaminated areas that can be applied in Korea is clearance in accordance with Nuclear Safety Commission Notice No. 2020-06. In addition, there has been no case of demolishing large-scale nuclear facilities in Korea. Therefore, there are limitations in domestic laws and standards to be applied to the dismantling of contaminated areas of KRR-1&2. The IAEA and the United States specify comprehensive matters such as optimization of radiation protection and minimization of waste products. The EU recommends demolition after decontamination by removing contaminated areas before demolition of buildings. It also presents three options for reuse, recycling, and disposal of buildings and building waste. In particular, in the case of Germany, detailed radioactivity measurement methods for deregulation of buildings and building waste are presented in accordance with the EU’s guidelines. As a result of synthesizing this, it is judged that the EU and Germany building clearance plan will be suitable for domestic application.

The purpose of this study was to effectively purify U-contaminated soil-washing effluent using a precipitation/distillation process, reuse the purified water, and self-dispose of the generated solid. The U ions in the effluent were easily removed as sediments by neutralization, and the metal sediments and suspended soils were flocculated–precipitated by polyacrylamide (PAM). The precipitate generated through the flocculation–precipitation process was completely separated into solid–liquid phases by membrane filtration (pore size < 45 μm), and Ca2+ and Mg2+ ions remaining in the effluent were removed by distillation. Even if neutralized or distilled effluent was reused for soil washing, soil decontamination performance was maintained. PAM, an organic component of the filter cake, was successfully removed by thermal decomposition without loss of metal deposits including U. The uranium concentration of the residual solids after distillation is confirmed to be less than 1 Bq·g−1, so it is expected that the self-disposal of the residual solids is possible. Therefore, the treatment method of U-contaminated soil-washing effluent using the precipitation/distillation process presented in this study can be used to effectively treat the washing waste of U-contaminated soil and self-dispose of the generated solids.

Safety for the radioactive waste disposed of in the disposal facility should be secured through safety assessment in consideration of the various situations. In this study, the influence and correlation of EDTA and ISA, which are the factors that can impede the safety of the disposal facility, were analyzed using the PHREEQC computational code. Thermodynamic database (TDB) of Andra, specific ion interaction theory (SIT) model as ionic strength correction model, radionuclides (Ni, Am, Pu) were adopted to perform the calculation on the distribution of chemical species by pH. According to the results, EDTA dominated the system and the effect of ISA is relatively small for the distribution of the chemical species of divalent and trivalent cations in neutral and weak base conditions. In the case of the tetravalent cations, the effect of ISA increased compared to the previous case especially in the strong base conditions. In conclusion, EDTA has a more significant effect on the system than ISA under the environment of the domestic disposal facility. Furthermore, when EDTA and ISA are present simultaneously in the system, the effects of two materials are inversely proportional and this characteristic should be considered during the safety assessment.

Radioactive waste disposal facility in Korea, radioactive waste packaged in 200 L drums is placed in a concrete disposal container and disposed of at an underground silo type (cave) disposal facility. At this time, the disposal container cover is seated on the top of the disposal container, and if the disposal container and the cover are not completely combined, the container cover is raised up from the top of the disposal container, so safety problems may occur when stacking the disposal container. Therefore, various methods exist to secure a margin for the pure height inside the disposal container. The disposal container cover only covers the upper surface of the container to shield radiation, and structural performance is not required. Therefore, the method of processing the cover, such as a method of making the cover of the disposal container thin, is the easiest method to apply. In this study, a method to reduce the thickness of the cover of a concrete disposal container was devised, and structural performance under usability conditions such as lifting and seating was analyzed. In addition, the disposal container cover has a reinforced concrete form in which dissimilar materials (concrete and steel) are combined, an integrated analysis was performed to secure the reliability of the analysis results for this, and the analysis results were described. It was found that the proposed disposal container cover structure can improve usability by reducing the stress concentration phenomenon.

When the decommissioning of a nuclear power plant begins in earnest, starting with Kori Unit 1, it is necessary to dispose of intermediate-level wastes such as high-dose waste filters and waste resin stored in the power plant, as well as the internal structures of the reactor. However, there are no intermediate-level waste disposal facilities in Korea, and the maintenance of acceptance criteria considering the physical, chemical, and radiological characteristics of intermediate-level waste is insufficient. In this paper, in preparation for the establishment of domestic intermediate-level waste treatment/disposal and acceptance standards, the following major foreign countries’ legal and institutional standards for intermediate-level waste are reviewed, and based on this, factors to be considered when establishing domestic intermediate-level waste treatment/disposal standards were derived. First, although the USA does not define and manage intermediate-level wastes separately, low-level wastes were separated into Class A, B, and C, where land disposal is allowed, and GTCC, which does not allow land disposal. However, it was recently confirmed that the position was changed to recognize the possibility of land disposal of GTCC waste under the condition that the dose to inadvertent intruders does not exceed 5 mSv·yr−1 and a barrier against inadvertent intrusion valid for 500 years is installed. Second, Sweden classifies intermediate-level wastes into short-lived and longlived intermediate-level wastes. The maximum dose rate permitted on packages are different for each vault and a silo of the SFR where short-lived wastes; 100 mSv·h−1 or less is disposed of in BMA, 10 mSV·h−1 or less in BTF, 2 mSv·h−1 or less in BLA and 500 mSv·h−1 or less in silo. Meanwhile, a repository for long-lived low and intermediate level waste, SFL, which could contains significant amounts of nuclides with a half-life greater than 31 years, operations are planned to commence in 2045. Third, France also manages short-lived intermediate-level wastes and long-lived intermediatelevel wastes separately, and the short-lived intermediate-level wastes were disposed of together with short-lived low-level wastes at the La Manche and L’Aube repository. France announced the Cigéo Project, a high- and medium-level long-lived waste plan in 2012, and submitted the creation authorization application for in 2021 with the goal of operating a repository in 2025. Finally, the UK defines intermediate-level waste as “waste whose activity level exceeds the upper limit for low-level waste but does not require heating, which is considered in the design of storage or disposal facilities” and established NIREX to provide deep disposal of intermediate-level radioactive waste. In Finland, wastes with radioactive concentrations of 1 MBq/kg to 10 GBq·kg−1 are classified as intermediatelevel wastes, and a repository was constructed and operated in a bedrock of about 110 m underground. Because the domestic classification standard simply classifies intermediate-level waste as waste exceeding the activity level of low-level waste limit, not high-level wastes, it is necessary to establish treatment and disposal standards by subdividing them by dose rate and long-lived radionuclides concentration to safely and efficiently dispose of intermediate-level waste for. Additionally, there is a need to decide whether or not to reflect safety by inadvertent intruders when evaluating the safety of intermediate-level disposal.

The disposing method of the low-intermediate-level radioactive waste, near-surface disposal facilities are generally used. This disposal method refers to a method of constructing a concrete structure on the surface of the ground, putting radioactive waste in it, and covering it with an engineered barrier to isolate human life. Among these, engineered barriers mean covering multiple layers of heterogeneous materials such as sand, clay, and gravel. Engineering barriers have the purpose of delaying the release of radioactive materials into the natural environment as much as possible, and maintaining the isolation of radioactive waste and human life for as long as possible. In this study, the design and construction method of the facility to demonstrate the performance of the engineered barrier that isolates the surface disposal facility from nature was described. In addition, the design and construction method of monitoring technology that can monitor the safety of engineered barriers by measuring information such as moisture, temperature, and slope safety in real time was also explained.

The natural barrier, a component of the deep disposal system, has site-specific characteristics depending on the site of the repository, and is one of the main considerations for long-term safety evaluation after closure along with the engineered barrier among the multiple barrier systems of the repository. The natural barrier is defined in Korea as the natural underground and surface structures that can restrict the exposure of radioactive waste, human intrusion or groundwater infiltration into a disposal facility, and the transfer of radionuclides. It includes bedrocks and soils surrounding the engineered barriers of radioactive wastes [Notice of the NSSC, No. 2020021]. This study analyzed foreign regulatory requirements related to natural barriers, requirements for natural barrier and performance target of Sweden and Finland (safety functions and target characteristics of natural barriers, e.g. natural barrier composition, geological characteristics, hydrogeological characteristics). Overseas regulations and cases referenced to derive regulations of general safety requirements on natural barrier are IAEA SSG-14, SSMFS 2008:21 in Sweden, STUK/Y/4/2018 in Finland, and POSIVA SKB Report 01, a joint report between POSIVA and SKB. The repository site and repository depth should be chosen so that the geological formation provides adequately stable and favorable conditions to ensure that the repository barriers perform as intended over a sufficient period of time. The conditions intended primarily concern temperature- related, hydrological, mechanical (for example, rock mechanics and seismology) and chemical (geochemistry, including groundwater chemistry) factors. Furthermore, the repository site should be located at a secure distance from natural resources exploited today or which may be exploited in the future [SSMFS 2008:21]. Finland regulations also suggests similar requirements [STUK Y-4-2018]. According to the above regulations, POSIVA SKB report 01 mentions both the host rock and the underground opening as natural barriers and requires a safety function, and the main safety functions of the host rock and underground opening are as follows: (1) Isolation from the surface environment; (2) Favorable thermal conditions; (3) Mechanically stable conditions; (4) Chemically favorable conditions; and (5) Favorable hydrogeological conditions with limited transport of solutes. Such safety functions would provide insight for understanding of the natural barrier of deep geological disposal system.

Deep geological disposal (DGD) of spent nuclear fuels (SNF) at 500 m–1 km depth has been the mainly researched as SNF disposal method, but with the recent drilling technology development, interest in deep borehole disposal (DBD) at 5 km depth is increasing. In DBD, up to 40SNF canisters are disposed of in a borehole with a diameter of about 50 cm, and SNF is disposed of at a depth of 2–5 km underground. DBD has the advantage of minimizing the disposal area and safely isolating highlevel waste from the ecosystem. Recently, due to an increasing necessity of developing an efficient alternative disposal system compared to DGD domestically, technological development for DBD has begun. In this paper, the research status of canister operation technology and plans for DBD demonstration tests, which subjects are being studied in the project of developing a safety-enhancing high-efficiency disposal system, are introduced. The canister operation technology for DBD can be divided into connection device development and operation technology. The developing connection device, emplacing and retrieving canisters in borehole, adopted the concept of a wedge thus making replacement equipment at the surface unnecessary. The new connection device has the advantage of being well applied with emplacement facilities only by simple mechanical operation. The technology of operating a connection device in DBD can be divided into drill pipe, coiled tubing, free-drop, and wireline. The drill pipe is a proven method in the oil industry, but requiring huge surface equipment. The coiled tubing method uses a flexible tube and shares disadvantages as the drill pipe. The free-drop is a convenient method of dropping canister into a borehole, but has a weakness in irretrievability in an accident. Finally, the wireline method can be operational on a small scale using hydraulic cranes, but the number of operated canisters at once is limited. The test facility through which the connection device is to be tested consists of dummy canister, borehole, lifting part, monitoring part, and connecting device. The canister weight is determined according to the emplacement operation unit. The lifting part will be composed following wireline consisting of a crane, a wire and a winding system. The monitoring part will consist of an external monitoring system for hoists and trolleys, and an internal monitoring system for the connection device’s location, pressure, and speed. In this project, a demonstration test will be conducted in a borehole with 1km depth, 10 cm diameter provided by KAERI to verify operation in the actual drilling environment after design improvement of the connecting device. If a problem is found through the demonstration test, the problem will be improved, and an improved connection device will be tested to an extended borehole with a 2 km depth, 40 cm diameter.