Properties of bentonite, mainly used as buffer and/or backfill materials, will evolve with time due to thermo-hydro-mechanical-chemical (THMC) processes, which could deteriorate the long-term integrity of the engineered barrier system. In particular, degradation of the backfill in the evolution processes makes it impossible to sufficiently perform the safety functions assigned to prevent groundwater infiltration and retard radionuclide transport. To phenomenologically understand the performance degradation to be caused by evolution, it is essential to conduct the demonstration test for backfill material under the deep geological disposal environment. Accordingly, in this paper, we suggest types of tests and items to be measured for identifying the performance evolution of backfill for the Deep Geological Repository (DGR) in Korea, based on the review results on the performance assessment methodology conducted for the operating license application in Finland. Some of insights derived from reviewing the Finnish case are as follows: 1) The THMC evolution characteristics of backfill material are mainly originated from hydro-mechanical and/or hydrochemical processes driven by the groundwater behavior. 2) These evolutions could occur immediately upon installation of backfill materials and vary depending on characteristics of backfill and groundwater. 3) Through the demonstration experiments with various scales, the hydro-mechanical evolution (e.g. advection and mechanical erosion) of the backfill due to changes in hydraulic behavior could be identified. 4) The hydro-chemical evolution (e.g. alteration and microbial activity) could be identified by analyzing the fully-saturated backfill after completing the experiment. Given the findings, it is judged that the following studies should be first conducted for the candidate backfill materials of the domestic DGR. a) Lab-scale experiment: Measurement for dry density and swelling pressure due to saturation of various backfill materials, time required to reach full saturation, and change in hydraulic conductivity with injection pressure. b) Pilot-scale experiment: Measurement for the mass loss due to erosion; Investigation on the fracture (piping channel) forming and resealing in the saturation process; Identification of the hydro-mechanical evolution with the test scale. c) Post-experiment dismantling analysis for saturated backfill: Measurement of dry density, and contents of organic and harmful substances; Investigation of water content distribution and homogenization of density differences; Identification of the hydro-chemical evolution with groundwater conditions. The results of this study could be directly used to establishing the experimental plan for verifying performance of backfill materials of DGR in Korea, provided that the domestic data such as facility design and site characteristics (including information on groundwater) are acquired.

The effect of various physicochemical processes, such as seawater intrusion, on the performance of the engineered barrier should be closely analyzed to precisely assess the safety of high-level radioactive waste repository. In order to evaluate the impact of such processes on the performance of the engineered barrier, a thermal-hydrological-chemical model was developed by using COMSOL Multiphysics and PHREEQC. The coupling of two software was achieved through the application of a sequential non-iterative approach. Model verification was executed through a comparative analysis between the outcomes derived from the developed model and those obtained in prior investigations. Two data were in a good agreement, demonstrating the model is capable of simulating aqueous speciation, adsorption, precipitation, and dissolution. Using the developed model, the geochemical evolution of bentonite buffer under a general condition was simulated as a base case. The model domain consists of 0.5 m of bentonite and 49.5 m of granite. The uraninite (UO2) was assigned at the canister-bentonite interface as the potential source of uranium. Assuming the lifetime of canister as 1,000 years, the porewater mixing without uranium leakage was simulated for 1,000 years. After then, the uranium leakage through the dissolution of uraninite was initiated and simulated for additional 1,000 years. In the base case model, where the porewater mixing between the bentonite and granite was the only considered process, the gypsum tended to dissolve throughout the bentonite, while it precipitated in the vicinity of bentonite-granite boundary. However, the precipitation and dissolution of gypsum only showed a limited effect on the performance of the bentonite. Due to the low solubility of uraninite in the reduced environment, only infinitesimal amounts of uranium dissolved and transported through the bentonite. Additional cases considering various environmental processes, such as seawater or cement porewater intrusion, will be further investigated.

The increasing accumulation of spent nuclear fuel has raised interest in High-Level Waste (HLW) repositories. For example, Sweden is under construction of the KBS-3 repository. To ensure the safety of such HLW repository, various countries have been developing assessment models. In the Republic of Korea, the Korea Atomic Energy Research Institute has been developing on the AKRS model. However, traditional safety assessment models have not considered the fracture growth in the far-field host rock as a function of time. As repository safety assessments guarantee safety for million years, sustained stress naturally leads to the progressive growth of fractures as time goes on. Therefore, it becomes essential to account for fracture growth in the surrounding host rock. To address this, our study proposes a new coupling scheme between the Fracture growth model and the radionuclide transport model. That coupling scheme consists of the Cubic Law model as a fracture growth function and the GoldSim code which is a commercial software for radionuclide transport calculations. The model that adopting such fracture growth functions showed an increase of up to 15% in the release of radionuclide compared to traditional assessment models. our observations indicated that crack growth as a function of time led to an increase in hydraulic conductivity that allowed more radionuclide transport. Notably, these findings show the significance of adopting fracture growth models as a critical element in evaluating the safety of nuclear waste repositories.

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.

With the importance of permanent disposal of high-level radioactive waste (HLW) generated in Korea, the deep geological disposal system based on the KBS-3 type is being developed. Since the deep geological repository must provide the long-term isolation of HLW from the surface environment and normal habitats for humans, plants, and animals, it is essential to assess the longterm performance of the disposal facility considering thermal-hydraulic-mechanical-chemical (TH- M-C) evolution. Decay heat dissipated from HLW contained in the canister causes an increase in temperature in the adjacent area. The requirement for the maximum temperature is established in consideration of the possibility of bentonite degradation. Therefore, when designing the repository, the temperature in the region of interest should be identified in detail through the thermal evolution assessment to ensure that the design requirement is satisfied. In the thermal evolution analysis, it is needed to evaluate the temperature distribution over the entire area of the disposal panel to consider the heat generated from both a single canister and adjacent canisters. Computational fluid dynamics (CFD) codes are widely used for detailed temperature analysis but are limited to simulating a wide range. Accordingly, in this study, we developed an analytical solution-based program for efficiently calculating the temperature distribution throughout the deposition panel, which is based on threedimensional heat conduction equations. The code developed can assess the temperature distribution of engineered and natural barrier systems. Principal parameters to be inputted are as follows: (a) geometry of the panel (e.g. width, length, height, spacing between canisters), (b) geometry of the canister (e.g. diameter, height), (c) thermal properties of bentonite and host-rock, (d) initial conditions (e.g. residual heat, temperature), and (e) time information (e.g. canister emplacement rate, time-interval, period). Through the calculation for the conceptual problem of a deposition panel capable of accommodating 900 (i.e. 30×30) canisters, it was confirmed that the program can adequately predict when and where the maximum temperature will occur. It is expected that the overall temperature distribution within the panel can be obtained by the evaluation of the entire region using this program reflecting the detailed design of the repository to be developed in the future. In addition, the thermal evolution analysis considering the influence of other canisters can be performed by applying the results as boundary conditions in the CFD analysis.

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.

In the design of HLW repositories, it is important to confirm the performance and safety of buffer materials at high temperatures. Most existing models for predicting hydraulic conductivity of bentonite buffer materials have been derived using the results of tests conducted below 100°C. However, they cannot be applied to temperatures above 100°C. This study suggests a prediction model for the hydraulic conductivity of bentonite buffer materials, valid at temperatures between 100°C and 125°C, based on different test results and values reported in literature. Among several factors, dry density and temperature were the most relevant to hydraulic conductivity and were used as important independent variables for the prediction model. The effect of temperature, which positively correlates with hydraulic conductivity, was greater than that of dry density, which negatively correlates with hydraulic conductivity. Finally, to enhance the prediction accuracy, a new parameter reflecting the effect of dry density and temperature was proposed and included in the final prediction model. Compared to the existing model, the predicted result of the final suggested model was closer to the measured values.

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.

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.

In Korea, 483,102 assemblies of spent fuel have been discharged and stored in sites, as of 2019. However, total capacity for site storage is 529,748 assemblies, and more than 90% is already saturated. Wolsong site, the most saturated site, started to construct more dry storage to extend the capacity in 2020. Spent fuel and high-level waste (HLW) is a big concern in Korean nuclear industry. Then, master plan for management of spent fuel is once announced by Ministry of Trade, Industry and Energy (MOTIE) in 2016 and reviewed by civil committee in 2019. The core contents of the plan are establishing schedule for construction of HLW management facility in one area, and construction of temporary dry storage in each site, if unavoidable. For HLW management facility, there are three following schedules: siting of Underground Research Laboratory (URL) and Interim Storage by 2020, operation of facilities initiated by 2030, and operation of final disposal facility initiated by 2050. Final repository will be designed as deep geological repository. The concept of deep geological disposal is that spent nuclear fuel is placed in disposal containers that can withstand corrosion and pressure in long-term, permanently isolated from the human sphere of life, and dumped in deep geological media, such as crystalline rocks and clay layer, at a depth of 300 to 1,000 meters underground. The safety assessment of waste disposal sites focuses on determining whether the disposal sites meet the safety requirements of national regulatory authority. This safety assessment evaluates the potential radiation dose of radionuclides from the disposal site to humans or the environment. In this case, the calculation is performed assuming that all engineering barriers of the disposal site have collapsed in a long-term period. Then radionuclides are released from the waste, and migrated in groundwater. The dose resulting from the release and migration of radionuclides on the concentration of nuclides in groundwater. In general, metallic nuclides may exist in water in various ionic states, but some form colloids. This colloid allows more nuclides to exist in water than in solubility. Therefore, more doses may occur than we know generally predict. To determine the impact of colloids, we performed the safety assessment of the Yucca Mountain repository as an example.

Backfill is one of the main components of engineered barrier in a high-level waste repository. The material selection of the backfill determines the barrier performance of the backfill. Overseas, its related research has been carried out mainly in Sweden, Finland, Canada, and Japan. However, Korea has recently started backfill research, and it is urgent to select a potential material for establishing the concept of backfill material and conducting backfill research. This study reviews NEA report, potential materials for overseas backfill research, advantages and disadvantages of single and mixed backfill materials, cases of license applications in Finland and Sweden for the selection of potential materials for backfill in Korea’s high-level waste repository. The review results indicated that it is reasonable to carry out backfill research according to the following plan: Both single and mixed materials are considered as potential materials for backfill research; experiments and performance studies are conducted with these materials; and, based on the results, a potential material or candidate material for the backfill suitable for the HLW repository in Korea is determined. For this plan, the single material is tentatively selected, as in Sweden, as bentonite with a montmorillonite content of about 40-50%. Then, if the selection criteria for montmorillonite content are determined through experiments and performance studies, we determine the final potential backfill material. As for the mixed backfill material, the bentonite/crushed rock mixture seems to be more advantageous than the bentonite/sand mixture considering the disposing problem of crushed rock generated from tunnel excavation and economic feasibility through its recycling. It is thought that the bentonite used in the bentonite/crushed rock mixture should have a higher montmorillonite content than bentonite used as a single backfill material since the crushed rock acts as an inert material in the mixture. The results of this study can be used as basic data for selecting the backfill material to be applied to the high-level waste repository in Korea, and can be used as a guideline for selecting the potential material required for backfill experiments and performance studies to be carried out in the future.

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.

Safety evaluation of high-level radioactive waste disposal facilities including spent nuclear fuel is a very urgent and critical issue, and in order to do so, it is very important to develop a safety case that includes Feature, Event, Process (FEP) analysis, scenario development, and scenario uncertainty evaluation. In the case of Korea, the disposal of spent nuclear fuel is recognized as an unavoidable option, and in the end, Korea’s specific FEP (SFEP) development and safety evaluation according to the scenario should be conducted. Because each country’s situation and environment are different, it is necessary to develop an SFEP based on a generic FEP (International FEP). To this end, an understanding of IFEP is essential. In this study, about 1,000 major terms appearing in the OECD/NEA IFEP are classified to where each of them belongs among F, E, and P, and which FEP each word belongs to, and the correlation between the frequency of occurrence and each term is analyzed. This result will serve as a reference for the results of SFEP analysis such as POSIVA and SKB, which our research team will analyze later. In addition, each term belongs to which academic field, and the most appropriate translation for translating each term into Korean is also described.

본 연구의 목적은 고준위폐기물 처분기술 개발과 관련하여 현장실증 연구를 위해 사용될 공학규모 이상의 균질 완충재 블록 을 제작하기 위한 새로운 방법론을 제시하는 것이다. 이와 관련하여 플롯팅 다이(floating die) 방식의 프레스 재하 및 냉간 등방압프레스(CIP; Cold Isostatic Press) 기법을 국내 최초로 완충재 제작에 적용하였다. 또한 소요 밀도기준을 충족하는 완충재 블록을 생산하기 위한 최적의 제작조건(프레스 및 CIP의 소요 압력)과 현장 적용성을 분석하였다. 상기 기법의 적용을 통해 완충재 블록 내 밀도분포 편차가 현저히 감소하였으며, 이와 동시에 평균 건조밀도가 소폭 상승하고 약 5%의 크기가 감소하였다. 또한 CIP 적용을 통해 응력이완(stress release) 현상이 감소하고, 이로 인해 시간 경과에 따른 표면균열 발생이 현저히 저감됨을 시험제작을 통해 확인하였다. 본 연구에서 제시된 방법론은 공학규모 이상의 균질한 완충재 블럭을 성형할 수 있으며, 또한 이는 선진핵주기 고준위폐기물처분시스템(AKRS; Advanced Korea Reference Disposal System of HLW)의 완충재 소요 밀도기준을 충족하는 것으로 분석되었다.

본 연구에서는 벤토나이트의 변질을 모사하기 위해 TOUGHREACT를 이용하여 열-수리-화학적 개념 모델링을 수행하였다. 모델링 결과 벤토나이트의 포화도는 지속적으로 증가하여 약 10 년 후에 포화상태에 도달하였다. 또한 온도는 급격히 증가 하여 0.5 년 이후에는 구리관으로부터 거리에 따라 일정한 온도 구배가 유지되었다. 이러한 열-수리 조건 변화에 따라 화학 적으로는 경석고와 방해석의 변질이 주로 발생하였다. 경석고와 방해석은 지하수가 유입됨에 따라 지속적으로 용해되었으 나, 온도가 높은 구리관 인근에서는 침전하는 경향을 보여주었다. 또한 경석고와 방해석의 침전으로 인해 구리관 인근의 공 극률과 투과도가 감소하였다. 확산 상수 변화에 대한 모델링 결과 경석고와 방해석의 변질은 확산 상수에 매우 민감하였으 며, 이는 결과적으로 수리적 특성인 공극률과 투과도에 영향을 미치고 있었다. 본 연구는 고준위 방사성폐기물 처분장 안전 성 연구에 기초적인 자료를 제공해 줄 것으로 판단된다.

고준위폐기물처분장에서 완충재는 공학적방벽의 주요 구성요소 중 하나이다. 본 연구에서는 국·내외의 완충재 요구사항과 성능기준을 분석하고, 우리나라 고준위폐기물처분장에 적합한 완충재 개념 도출을 위한 접근방안을 제시하였다. 완충재의 주요 성능기준 인자항목으로, 수리전도도, 핵종 저지능, 팽윤압, 열전도도, 역학적 특성치(mechanical properties), 유기물함 량(organic carbon content), 일라이트화 속도(illitization rate) 등을 고려하였다. 우리나라 고준위폐기물처분장 완충재 물질 로서 국산 벤토나이트(Ca-벤토나이트)와 대안재로 MX-80 벤토나이트(Na-벤토나이트)를 제안하였다. 완충재의 기술사양은 Ca-벤토나이트 경우엔 우리나라의 성능기준을, Na-벤토나이트의 경우는 스웨덴의 성능기준을 보수적으로 만족하는 값으로 설정하였다. 완충재의 두께는 전단거동, 핵종 유출, 열전도의 측면에서 평가하여 결정하였으며, 평가결과 완충재의 두께는 0.25 ~ 0.5 m 사이가 적절하였다. 그러나 최종적인 완충재의 두께는 향후 보다 심도 있는 열-수리-역학적 평가와 경제적, 공 학적 측면을 고려하여 결정하여야 할 것이다.

현재까지 개발된 고준위폐기물 심지층처분장의 열-수리-역학적 복합거동 해석을 위한 전산 코드의 현황을 조사하고, 문헌에 보고된 각 코드에 의한 계산치와 현장실험 측정치의 비교 결과를 이용하여, 기존 전산 코드들의 신뢰도를 분석하였다. 개발된 전산코드들은 완충재가 없는 처분장에서는 붕괴열에 따른 암반의 열-수리-역학적 거동을 비교적 잘 모사하였으나, 포화 경암층에 위치한 완충재가 존재하는 처분장의 공학적방벽시스템 내에서 일어나는 열-수리-역학적 복합거동의 예측은 만족스럽지 못하였다. 현재 제안된 열-수리-역학적 복합거동 해석모델을 고준위폐기물 처분장 공학적방벽시스템의 거동 해석에 적용하기위해서는 완충재 내의 수분함량 및 전 압력 분포를 보다 정교하게 모사할 수 있도록 수학적 모델의 개선이 필요하다.