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.

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.

Effective containment and disposal of high-level radioactive waste is critical to ensure long-term environmental and human safety. Especially bentonite, which is widely used as a buffer material due to its favorable characteristics such as swelling ability and low permeability, plays an important role in preventing the migration of radioactive waste into the surrounding environment. However, the long-term performance of bentonite buffer remains an area of ongoing investigation, with particular attention focused on erosion mechanisms induced by swelling and groundwater flow. The erosion of the bentonite buffer can significantly impact the integrity of buffer and lead to the formation of colloids, which could potentially facilitate the transport of radionuclides through groundwater. Therefore, quantification of bentonite buffer erosion based on an understanding of the underlying mechanisms and factors that influence bentonite buffer erosion, is essential for the safety assessment of high-level radioactive waste repositories. In this study, we aimed to develop a bentonite buffer erosion model using the Adaptive Processbased total system performance assessment framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). The impact of bentonite erosion on performance assessment can be broadly divided into bentonite property degradation by the penetration of the bentonite buffer into rock fractures and the formation of pseudocolloids. To simulate this phenomenon, Two-region model based on a dynamic bentonite diffusion model is adopted, which can quantify the extent of bentonite intrusion and loss by erosion. Using this Tworegion model, a numerical model was developed to simulate the degradation of bentonite properties based on the amount of bentonite intrusion, as well as to simulate the migration of pseudocolloids in the near-field by deriving the amount of pseudocolloid production based on the loss of bentonite and the sorption rate of radionuclides. To check the applicability of the developed numerical model, preliminary analysis was performed for the effect of bentonite erosion in terms of process-based performance assessment. It is anticipated that this comprehensive model developed in this study will contribute to the accurate and reliable assessment of the long-term performance and safety of high-level radioactive waste repositories.

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.

Bentonite is a widely used buffer material in high-level radioactive waste repositories due to its favorable properties, including its ability to swell and low permeability. Bentonite buffers play an important role in safe disposal by providing a low permeability barrier and preventing radionuclides migration into the surrounding rock. However, the long-term performance of the bentonite buffer is still an area of research, and one of the main concerns is the erosion of the buffer due to swelling and groundwater flow. Erosion of the bentonite buffer can have a significant impact on repository safety by reducing the integrity of the buffer and forming colloids that can transport radionuclides through groundwater, potentially increasing the risk of radionuclide migration. Therefore, understanding the mechanisms and factors that influence the erosion of the bentonite buffer is critical to the safety assessment of high-level radioactive waste repositories. In this study, we attempted to develop the bentonite buffer erosion model using Adaptive Processbased total system performance assessment framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). First, the erosion phenomenon was divided into two stages: bentonite buffer penetration into rock fractures and colloid formation. As an initial step in the development of the buffer erosion model, a bentonite buffer intrusion into the fracture and consequent degradation of buffer property were considered. For this purpose, a tworegion model based on the dynamic bentonite diffusion model was adopted which is one of the methods for simulating bentonite buffer intrusion. And, it was assumed that the buffer properties, such as density, porosity and permeability, thermal conductivity, modulus of elasticity, and mechanical strength, are degraded as the buffer erodes. The bentonite buffer degradation model developed in this study will serve as a foundation for the comprehensive buffer erosion model, in conjunction with the colloidal formation model in the future.

In order to reduce the area of the high-level radioactive waste (HLW) repository, a buffer material with high thermal conductivity is required. This is because if the thermal conductivity of the buffer material is high, the distance between the disposal tunnels and the deposition holes can be reduced. Sand, which is a natural material and has higher thermal conductivity than bentonite, is added to bentonite to develop an enhanced buffer material. For the sand-bentonite mixture, it is important which sand to use and how much to add because an enhanced buffer material should satisfy both hydraulic (H) and mechanical (M) performance criteria while improving thermal conductivity (T). In this study, we would like to show what type of sand and how much sand should be added to develop an enhanced buffer material by adding sand to Gyeongju bentonite, a representative bentonite in Korea. For this purpose, the thermal conductivity, hydraulic conductivity, and swelling pressure of the sand-Gyeongju bentonite mixture according to the sand addition rate were measured. It is more efficient to use silica sand with smaller particles than Jumunjin sand which is a representative sand in Korea as an additive for an enhanced buffer material than using the Jumunjin sand. In order for the sand-Gyeongju bentonite buffer material to satisfy both the hydraulic and mechanical performance criteria as a buffer material while increasing the thermal conductivity, it is judged that the optimum dry density is 1.7 g/cm3 at least and the optimum sand addition rate is 10% at most.

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.

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.

Increasing thermal conductivity of buffer materials makes the disposal tunnel and hole spacings in high-level radioactive waste (HLW) repositories decrease, so that the area of the HLW repository decreases, which gives more choices to choose the HLW repository site and economical cost savings to construct HLW repositories. Thus, developing enhanced buffer materials with improved thermal conductivity is needed. One of the methods to develop enhanced buffer materials is to add additives to the bentonite which is main material for buffer materials. Most additives have high thermal conductivity, but most additives do not swell or less swell than bentonite, so that the swelling pressure of the enhanced buffer materials by additives decreases compared to the swelling pressure of pure bentonite buffer materials. Swelling pressure is an important performance criterion to design buffer materials. Thus, it is important to confirm the swelling pressure of the enhanced bentonite. However, it is not simple to measure the swelling pressure of the buffer materials and furthermore, it takes several days to measure the swelling pressure of the buffer materials. For these reasons, swelling index can be considered to predict the swelling pressure of the enhanced buffer materials relatively. In this study, it was investigated through tests how the swelling index of bentonite-sand mixtures change according to the amount of sand and it was found that the linear relationship between swelling index and sand amount in the bentonite-sand mixtures.

The design of the high-level radioactive waste (HLW) repository is made for isolating the HLW from the groundwater system by using artificial and natural barriers. Granite is usually considered to be a great natural barrier for the HLW repository in various countries including Sweden, Canada, and Korea due to its low hydraulic permeability. However, many fractures that can act as conduits for groundwater and radionuclides exist in granite. Furthermore, the decay heat generated by the HLW can induce groundwater acceleration through the fracture. Since the direction, magnitude, and lasting time of the heat-induced groundwater flow can be differed depending on the fracture geometry, the effect of fracture geometry on the groundwater flow around the repository should be carefully analyzed. In this study, groundwater models were conducted with various fracture geometries to quantify the effect of various properties of fractures (or fracture networks) on the heat-induced groundwater flow. In all models, the pressure around the repository only lasted for a short period after it peaked at 0.1 years. In contrast, the temperature lasted for 10,000 years after the disposal inducing the convective groundwater flow. Single fracture models with different orientations were conducted to evaluate the variations in groundwater velocities around the repository depending on the fracture slope. According to the results, the groundwater velocity on the fracture was the fastest when the regional groundwater flow direction and the fracture direction coincided. In double fracture models, various inclined fractures were added to the horizontal fracture. Due to the intersecting, the groundwater flow velocity showed a discontinuous change at the intersecting point. Lastly, the discrete fracture network models were conducted with different fracture densities, length distributions, and orientations. According to the modeling results, the groundwater flow was significantly accelerated when the fracture network density increased, or the average fracture length increased. However, the effect of the fracture orientation was not significant compared to the other two network properties.

In recent years, the importance of the thermo-hydraulic-mechanical-chemical coupled processes is increasing in the performance assessment (PA) of the high-level radioactive waste repository. In the case of mechanical behavior, it is very important because it can affect fluid flow and radionuclide transport by changing the porosity and permeability of the medium. In particular, Excavation Damaged Zone (EDZ) should be considered essential in PA because the migration of radionuclide is affected by the enhanced hydraulic transmissivity and altered geomechanical behavior of EDZ. Furthermore, due to various thermo-hydraulic behaviors such as decay heat generated from radioactive waste, pore water pressure increase, and swelling pressure of bentonite buffer material, mechanical evolution is occurred which may change the size and physical properties of EDZ. Therefore, to solve this problem, analysis of coupled thermal-hydraulic-mechanical (THM) processes with the effect of long-term evolution of EDZ due to the mechanical behavior should be accompanied. In this study, numerical model for the long-term evolution due to mechanical behavior considering EDZ using the Adaptive Process-based total system performance analysis framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). In the case of EDZ, the concept of Mazars’ damage evolution model was applied to simulate the behavior using the continuum model, and the change in hydraulic properties according to the degree of damage was considered. To investigate the importance of mechanical behavior in PA, the results were compared by performing numerical analysis according to the presence or absence of mechanical analysis. Finally, numerical analysis considering the mechanical evolution of EDZ was conducted using the model developed in this study to investigate the effect of EDZ.

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.