Understanding the long-term geochemical evolution of engineered barrier system is crucial for conducting safety assessment in high-level radioactive waste disposal repository. One critical scenario to consider is the intrusion of seawater into the engineered barrier system, which may occur due to global sea level rise. Seawater is characterized by its high ionic strength and abundant dissolved cations, including Na, K, and Mg. When seawater infiltrates an engineered barrier, such dissolved cations displace interlayer cations within the montmorillonite and affect to precipitation/ dissolution of accessory minerals in bentonite buffer. These geochemical reactions change the porewater chemistry of bentonite buffer and influence the reactive transport of radionuclides when it leaked from the canister. In this study, the adaptive process-based total system performance assessment framework (APro), developed by the Korea Atomic Energy Research Institute, was utilized to simulate the geochemical evolution of engineered barrier system resulting from seawater intrusion. Here, the APro simulated the geochemical evolution in bentonite porewater and mineral composition by considering various geochemical reactions such as mineral precipitation/dissolution, temperature, redox processes, cation exchange, and surface complexation mechanisms. The simulation results showed that the seawater intrusion led to the dissolution of gypsum and partial precipitation of calcite, dolomite, and siderite within the engineered barrier system. Additionally, the composition of interlayer cation in montmorillonite was changed, with an increase in Na, K, and Mg and a decrease in Ca, because the concentrations of Na, K, and Mg in seawater were 2-10 times higher than those in the initial bentonite porewater. Further studies will evaluate the geochemical sorption and transport of leaked uranium-238 and iodine-129 by applying TDB-based sorption model.

In 2012, POSIVA selected a bentonite-based (montmorillonite) block/pellet as the backfilling solution for the deposition tunnel in the application for a construction license for the deep geological repository of high-level radioactive waste in Finland. However, in the license application (i.e. SC-OLA) for the operation submitted to the Finnish Government in 2021, the design for backfilling was changed to a granular mixture consisting of bentonite (smectite) pellets crushed to various sizes, based on NAGRA’s buffer solution. In this study, as part of the preliminary design of the deep geological repository system in Korea, we reviewed history and its rationale for the design change of Finland’s deposition tunnel backfilling solution. After the construction license was granted by the Finnish Government in 2015, POSIVA conducted various lab- and full-scale in-situ tests to evaluate the producibility and performance of two design alternatives (i.e. block/pellet type and granular type) for backfilling. Principal demonstration tests and their results are summarized as follows: (a) Manufacturing of blocks using three types of materials (Friedland, IBeco RWC, and MX-80): Cracking and jointing under higher pressing loads were found. Despite adjusting the pressing process, similar phenomena were observed. (b) 1:6 scale experiment: Confirmation of density difference inhomogeneity due to the swelling of block/pellet backfill and void filling due to swelling behavior into the mass loss area of block/pellet. (c) FISST (Full-Scale In situ system Test): Identification of technical unfeasibility due to the inefficient (too manual) installation process of blocks/pellets and development of an efficient granular in-situ backfilling solution to resolve the disadvantage. (d) LUCOEX-FE (Large Underground Concept Experiments – Full-scale Emplacement) experiment: Confirmation of dense/homogeneous constructability and performance of granular backfilling solution. In conclusion, the simplified granular backfill system is more feasible compared to the block/ pellet system from the perspective of handling, production, installation, performance, and quality control. It is presumed that various experimental and engineering researches should be preceded reflecting specific disposal conditions even though these results are expected to be applied as key data and/or insights for selecting the backfilling solution in the domestic deep geological repository.

For the deep geological repository, engineering barrier system (EBS) is installed to restrict a release of radionuclide, groundwater infiltration, and unintentional human intrusion. Bentonite, mainly used as buffer and backfill materials, is composed of smectite and accessory minerals (e.g. salts, silica). During the post-closure phase, accessory minerals of bentonite may be redistributed through dissolution and precipitation due to thermal-hydraulic gradient formed by decay heat of spent nuclear fuel and groundwater inflow. It should be considered important since this cause canister corrosion and bentonite cementation, which consequently affect a performance of EBS. Accordingly, in this study, we first reviewed the analyses for the phenomenon carried out as part of construction permit and/or operating license applications in Sweden and Finland, and then summarized the prerequisite necessary to apply to the domestic disposal facility in the future. In previous studies in Sweden (SKB) and Finland (POSIVA), the accessory mineral alteration for the post-closure period was evaluated using TOUGHREACT, a kind of thermal-hydro-geochemical code. As a result of both analyses, it was found that anhydrite and calcite were precipitated at the canister surface, but the amount of calcite precipitate was insignificant. In addition, it was observed that precipitate of silica was negligible in POSIVA and there was a change in bentonite porosity due to precipitation of salts in SKB. Under the deep disposal conditions, the alteration of accessory minerals may have a meaningful influence on performance of the canister and buffer. However, for the backfill and closure, this is expected to be insignificant in that the thermal-hydraulic gradient inducing the alteration is low. As a result, for the performance assessment of domestic disposal facility, it is confirmed that a study on the alteration of accessory minerals in buffer bentonite is first required. However, in the study, the following data should reflect the domestic-specific characteristics: (a) detailed geometry of canister and buffer, (b) thermal and physical properties of canister, bentonite and host-rock in the disposal site, (c) geochemical parameters of bentonite, (d) initial composition of minerals and porewater in bentonite, (e) groundwater composition, and (f) decay heat of spent nuclear fuel in canister. It is presumed that insights from case studies for the accessory mineral alteration could be directly applied to the design and performance assessment of EBS, provided that input data specific to the domestic disposal facility is prepared for the assessment required.

Spent nuclear fuel temporary storage in South Korea is approximately 70% of total storage capacity as of the 4th quarter of 2022 amount is stored. In addition, according to the analysis of the Korean Radioactive Waste Society, saturation of nuclear power plant temporary storage is expected sequentially from 2031, and accordingly, the need for high-level radioactive waste disposal facilities has emerged. Globally, after the conclusion of the EU Taxonomy, for nuclear energy in order to become an ecofriendly energy, it is necessary to have a high-level radioactive waste disposal site and submit a detailed operation plan for high-level radioactive waste disposal site by 2050. Finland and Sweden have already received permission for the construction of high-level radioactive waste disposal facilities, and other countries, such as Switzerland, Japan, the United States, and Canada, are in the process of licensing disposal facilities. In order to establish a repository for high-level radioactive waste, the performance and safety analysis of the repository must be conducted in compliance with regulatory requirements. For safety analysis, it needs a collection of arguments and evidence. and IAEA defined it as ‘Safety case’. The Systematic method, which derives scenarios by systematizing and combining possible phenomena around the repository, is widely used for developing Safety case. Systematic methods make use of the concept of Features, Events and Processes (FEP). FEP identifies features that affect repository performance, events that can affect a short period of time, and processes that can have an impact over a long period of time. Since it is a characteristic of the Systematic method to compose a scenario by combining these FEP, the Systematic method is the basic premise for the development of FEP. Completeness is important for FEP, and comprehensiveness is important for scenarios. However, combining all the FEP into one scenario is time-consuming and difficult to ascertain the comprehensiveness of the scenario. Therefore, an Integrated FEP list is being developed to facilitate tracking between FEP and scenarios by integrating similar FEP. In this study, during the integrated FEP development process, a method for utilizing experts that can be used for difficult parts of quantitative evaluation and a quantitative evaluation process through the method were presented.

Bentonite, a material mainly used in buffer and backfill of the engineering barrier system (EBS) that makes up the deep geological repository, is a porous material, thus porewater could be contained in it. The porewater components will be changed through ‘water exchange’ with groundwater as time passes after emplacement of subsystems containing bentonite in the repository. ‘Water exchange’ is a phenomenon in which porewater and groundwater components are exchanged in the process of groundwater inflow into bentonite, which affects swelling property and radionuclide sorption of bentonite. Therefore, it is necessary to assess conformity with the performance target and safety function for bentonite. Accordingly, we reviewed how to handle the ‘water exchange’ phenomenon in the performance assessment conducted as part of the operating license application for the deep geological repository in Finland, and suggested studies and/or data required for the performance assessment of the domestic disposal facility on the basis of the results. In the previous assessment in Finland, after dividing the disposal site into a number of areas, reference and bounding groundwaters were defined considering various parameters by depth and climate change (i.e. phase). Subsequently, after defining reference and bounding porewaters in consideration of water exchange with porewater for each groundwater type, the swelling and radionuclides sorption of bentonite were assessed through analyzing components of the reference porewater. From the Finnish case, it is confirmed that the following are important from the perspective of water exchange: (a) definition of reference porewater, and (b) variations in cation concentration and cation exchange capacity (CEC) in porewater. For applying items above to the domestic disposal facility, the site-specific parameters should be reflected for the following: structure of the bedrock, groundwater composition, and initial components of bentonite selected. In addition, studies on the following should be required for identifying properties of the domestic disposal site: (1) variations in groundwater composition by subsurface depth, (2) variations in groundwater properties by time frame, and (3) investigation on the bedrock structure, and (4) survey on initial composition of porewater in selected bentonite The results of this study are presumed to be directly applied to the design and performance assessment for buffer and backfill materials, which are important components that make up the domestic disposal facility, given the site-specific data.

Currently, there are 25 nuclear power plants (NPPs) in operation in Korea, including 22 pressurized water reactors (PWRs) and three pressurized heavy water reactors (PHWRs). Two NPPs, including Kori Unit 1 and Wolsong Unit 1, are permanently shut down and awaiting decommissioning. If Kori Unit 2, which is expected to be permanently shut down soon, is included, the number of decommissioning NPPs will be increased to three. Spent fuels (SFs) are continuously generated during the NPP operation, which are stored in an SF storage pool in NPPs to cool down the decay heat emitted from SFs. For safe NPP operation, SFs must be regarded as waste, and a disposal site must be selected to isolate SFs. However, an appropriate site has yet to be selected in Korea. SFs contain long-lived nuclides with a high specific activity. For disposal, it is important to characterize the nuclides in the fuels and delay the migration of the nuclides to the environment when SFs are placed in a future disposal facility. If the disposal container is broken, the nuclides in the fuels escape from the filling material, such as bentonite. These escaped nuclides are dissolved in groundwater and migrate to the surface of the earth. Thus, it is possible to assess the radiological impact, such as the exposure dose during and after the disposal, if the types and characteristics of nuclides in SFs are known. This study investigated the nuclides in SFs and identified exposure scenarios that may occur in the disposal process of SFs and migration characteristics when the nuclides leak into groundwater to propose a dose assessment methodology for workers and the public.

As Korea has relatively small land area and large population density compared to other countries considering the DGD concept such as Finland and Sweden, improvements of disposal efficiency in the viewpoint of the disposal area might be needed for the current disposal system to alleviate the difficulties of site selection for the HLW repository. In this research, we conduct a numerical investigation of the disposal efficiency enhancement for a high-level radioactive waste (HLW) repository through three design factors: decay heat optimization, increased thermal limit of buffer, and double-layer concept. In the optimized decay heat model, seven SNFs with the maximum and minimum decay heat depending on actual burn-up and cooling time are iteratively combined in a canister. Thermal limit of buffer is assumed as 100°C and 130°C for reference and high-efficiency repository concepts, respectively. By implementing an optimized decay heat model and a single-layer concept with a thermal limit of buffer set at 100°C, the disposal efficiency increases to 2.3 times of the improved Korean Reference disposal System (KRS+). Additionally, incorporating either an increased thermal limit of buffer to 130°C or a double-layer concept leads to a further 50% improvement in disposal efficiency. By integrating all three design factors, the disposal efficiency can be enhanced up to five times that of the KRS+ repository. Our analysis of rock mass stability reveals that increasing the thermal limit of buffer can generate rock spalling failure in a wider area. However, when accounting for the effect of confining stress by swelling of buffer and backfill using the Mohr-Coulomb failure criteria, the rock mass failure only occurred at the corner between the disposal tunnel and deposition hole when the thermal limit of buffer was increased and a single-layer concept was applied. The results given in this study can provide various options for designing the high-efficiency repository in accordance with the target disposal area and quality of the rock mass in the potential repository site.

Several countries have been operating radioactive waste disposal (RWD) programs to construct their own repositories and have used natural analogues (NA) studies directly or indirectly to ensure the reliability of the long-term safety of deep geological disposal (DGD) systems. A DGD system in Korea has been under development, and for this purpose a generic NA study is necessary. The Korea Atomic Energy Research Institute has just launched the first national NA R&D program in Korea to identify the role of NA studies and to support the safety case in the RWD program. In this article, we review some cases of NA studies carried out in advanced countries considering crystalline rocks as candidate host rocks for high-level radioactive waste disposal. We examine the differences among these case studies and their roles in reflecting each country’s disposal repository design. The legal basis and roadmap for NA studies in each country are also described. However because the results of this analysis depend upon different environmental conditions, they can be only used as important data for establishing various research strategies to strengthen the NA study environment for domestic disposal system research in Korea.

Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300–1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.

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.

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.

To decrease area of the repository for high-level radioactive waste, enhancing the disposal efficiency is needed for public acceptance. Previous studies regarding the performance assessment of KRS and KRS+ repository did not consider area-based variations of the geothermal gradient and rock thermal properties in Korea. This research estimated deposition hole spacing based on performance assessment of a repository using the distribution of geothermal gradient and rock thermal properties in Korea to increase disposal efficiency. Distributions of geothermal gradient, rock thermal properties were investigated based on 2019 Korea geothermal atlas published by Korea Institute of Geoscience and Mineral Resources (KIGAM). Effect of thermal performance parameters was analyzed using coupled thermal-hydraulic numerical simulations, and effect of rock thermal conductivity and deposition hole spacing on the maximum temperature of buffer was relatively large. In addition, distribution maps of thermal performance of a repository and deposition hole spacing were plotted using thermal performance parameters-maximum temperature of buffer regression equations and GIS data given by KIGAM. In the regions showing the highest maximum temperature of buffer in Korea, required deposition hole spacings were 10.5 m, 10.0 m, 10.1 m, respectively for KJ-II, MX-80, and FEBEX bentonite cases, and thereby additional disposal area of 40%, 33.3%, and 34.7% were required compared to that of the KRS+ repository. On the other hand, high disposal efficiency can be obtained in the regions showing the low maximum temperature of bentonite buffer. The methodology provided in this research can be used as one of the references for the selection of domestic candidate repository sites. Additional mechanical performance analysis should be conducted using distributions of mechanical properties of rock mass in Korea.

고준위방사성폐기물의 처분은 고심도 암반내에 처분시스템을 구축하는 심층 처분방법이 고려된다. 심층 처분은 처분용기, 완충재, 뒷채움재, 근계암반의 설계 요소인 공학적방벽과 천연 방벽으로 구성된다. 공학적방벽 중에서 벤토나이트 완충재는 암반으로부터 유입되는 지하수 흐름을 최소화하고 핵종 유출을 저지하는 기능을 한다. 지하수 유입으로 인한 완충재의 수리전도도 특성 규명은 처분장 공학적방벽의 안정성 및 건전성에 대한 성능 평가에 있어 중요한 사안이다. 본 연구에서는 경주 벤토나이트를 이용하여 다양한 건조밀도와 온도 조건에 따라 포화 수리전도도 실험을 수행하였으며, 120개의 실험 결과 를 다중 회귀 분석을 통해 수리전도도 추정 모델을 제시하였다. 실험 결과에서는 건조밀도가 커질수록 수리전도도가 감소하는 경향이 나타났다. 또한, 온도가 증가할수록 수리전도도가 증가하였다. 이러한 실험 결과들을 종합한 다중 회귀 분석 결과에서는 수리전도도 추정식의 결정계수(R2)가 0.93으로 높게 나타났다. 본 연구에서 제시된 수리전도도 추정식은 벤토나이트 완충재의 성능과 연관된 건조밀도와 온도의 영향을 고려하여 처분시스템의 공학적방벽 설계에 활용 될 것으로 판단된다.

국내 고준위 방사성폐기물 심층처분시스템에 대한 프로세스 기반의 종합성능평가체계(APro) 개발을 위하여 사용자 편의성이 향상된 모델링 인터페이스를 구축하였다. APro의 모델링 인터페이스는 프로그래밍 언어인 MATLAB을 이용하여 구축되었고, 다중물리현상 모사가 가능한 COMSOL과 지화학반응 계산이 가능한 PHREEQC를 계산 엔진으로 활용하여 연산 자분리 방식을 적용하였다. APro는 모델링 영역을 기존의 정형화된 처분시스템으로 제한함으로써 모델의 자유도는 낮지만, 사용자 편의성을 향상시켰다. 처분시스템에서 고려되는 주요 현상들을 모듈화하였고, 이를“Default process”와 다수의“Alternative process”로 구분하여 사용자가 선택할 수 있도록 함으로써 모델의 유연성을 높였다. APro는 크게 입력자료 부분과 계산실행 부분으로 구성된다. 기본 입력자료는 하나의 EXCEL 파일에 일정한 포맷으로 정리되고, 계산실행 부분은 MATLAB을 이용하여 코딩되었다. 최종적인 전체 계산 결과는 독립적인 COMSOL 파일 형태로 생성되도록 하여 COMSOL을 이용한 계산 결과의 후처리가 가능하도록 하였다.

고준위방사성폐기물 심층처분에서 처분안전성의 신뢰도를 향상시킬 수 있는 방안으로 그리고 처분 프로그램 개발 및 인허가를 위해 많은 나라들에서 자국에 적합한 safety case를 개발하고 있다. 본 연구에서는 방사성폐기물 처분을 위한 safety case 의 의의, 필요성, 개발과정들을 정리하고 소개하였다. 그리고 처분안전성을 safety case의 다양한 측면에서 논의하였다. 아울러 스위스, 일본, 미국, 스웨덴, 핀란드 등 해외의 safety case 개발 현황과 현재 KAERI에서 개발 중인 safety case의 개발 전략을 간략히 소개하였다. 고준위방사성폐기물 처분안전성의 신뢰도 향상을 위해 safety case 기반 하에서 어떤 노력들이 필요한지를 분석하였다. 그리고 국내 상황을 반영하여 신뢰할 수 있는 정보자료의 구축, 안전성 관련 과정들의 이해, 안전성 평가의 불확실성 저감, 이해당사자와의 의사소통, 공정성과 투명성 확보 등의 실행 방안을 제안하고 논의하였다. 본 논문에 제시된 내용들은 심층처분 safety case를 이해하고, 국내에서 개발하고 있는 고준위방사성폐기물 처분 safety case 개발을 통한 처분안전성 신뢰도 향상에 기여할 수 있을 것으로 기대한다.

원자력발전소에서 전기를 생산하고 난 후 발생하는 사용후핵연료 또는 이들 사용후핵연료의 재처리/재활용 공정으로부터 발생하는 고준위폐기물은 인간환경으로부터 안전하게 장기간 격리시켜야 한다. 최근 심부시추공 굴착기술의 획기적인 발전 으로 인하여, 방사성폐기물의 심부시추공 처분기술에 대한 연구가 의미 있게 진행되고 있다. 본 논문에서는 이러한 심부시추 공을 활용하여 고준위 방사성폐기물을 지하 3~5 km 심도에 격리시키는 심부시추공 처분기술의 국내 적용 가능성을 분석하 기 위하여 국내 심부 지하환경 특성에 대하여 예비분석 하였다 이를 위하여, 미국 및 유럽권 국가 연구사례와 기술개발 현황 을 검토하고, 실제 국내의 심부 지질조건을 검토하기 위하여 고지열 분포지역에 개발 중인 지열 탐사공을 대상으로 3~4 km 심도까지의 암석, 지온 등 특성 자료를 수집, 분석하였다. 결정질 암반 심도 및 지온경사 등 분석 결과와 국내 발생 사용후 핵연료를 바탕으로 심부시추공 처분시스템 구성요소인 처분용기, 밀봉시스템 등에 대하여 예비단계의 개념을 제안하였다.

사용후핵연료를 포함하는 고준위 방사성폐기물을 지질학적 조건이 안정적인 지하 3~5 km의 심도에 처분할 수 있다면 다음 과 같은 많은 장점이 있는 것으로 평가되고 있다. 즉, (1)암반 수리전도도가 매우 낮아 지하수가 생태계까지 도달하는데 속 도가 현저히 감소되며, (2)상부층 두께로 인하여 생태계와의 이격거리 확보에 유리하고, (3)지하수가 환원상태이므로 핵종 의 용해도가 매우 낮을 뿐만 아니라 (4)오랜 연령의 지하수에서는 핵종이 흡착된 콜로이드 생성과 이동이 극히 제한된다는 점이다. 이와 관련하여 심부시추공 처분(Deep Borehole Disposal) 연구는 심층 처분(Deep Geological Disposal) 시스템에 대한 이상적인 처분 대안기술로서 꾸준하게 진행되어 왔다. 본 논문에서는 최근 심부 시추기술이 비약적으로 발전됨에 따 라 의미있게 연구가 진행되고 있는 심부시추공 처분시스템을 국내 적용하기 위한 초기 단계로서 해외의 심부시추공 처분시 스템 기술개발 사례를 분석하였다. 이를 통하여 심부시추공 처분에 대한 일반적인 개념과 심부시추공 처분시스템 개념을 도 출한 연구사례를 국가별로 정리하였다. 이들 분석결과는 향후 심부시추공 처분기술의 국내 적용을 위한 입력자료로서 유용 하게 활용될 수 있을 것이다.

고준위폐기물 처분과 관련하여, 최근 저장소 형태의 처분장 개념에 대한 대안으로 검토되고 있는 시추 공 처분 개념에 대한 연구 현황을 정리하고 시추공 처분 개념의 국내 적용 가능성과 필요한 연구 항목에 대해 논의하였다. 현재 미국과 스웨덴을 중심으로 논의된 시추공 처분 개념은 심부시추공을 설치하여 지 하 3 - 5km 구간에 고준위폐기물을 처분하는 것을 의미하며, 현재까지의 연구 결과에 의하면 이 처분 개 념은 심부지하수의 층상구조, 작은 규모의 지표시설 등으로 인해 처분 및 비용 효율이 클 것으로 예상된 다. 이에 반해 국내에는 축적된 심부 지질 자료가 없어 적용 가능성에 대한 논의할 여지가 없다. 이에 저 장소 형태의 처분장 개념에 대한 대안으로 시추공 처분 개념을 검토하기 위해서는 향후 심지층 자료 확 보, 공학적 방벽 연구, 수치모의모델 개발, 처분 기술 개발 등의 연구가 필요하다.