KORAD (Korea Radioactive Waste Agency, http://www.korad.or.kr) has stored slightly contaminated ascon (asphalt coated concrete mixture) that was introduced to Gyeongju repository about a decade ago waiting for a final disposal. It is believed to be mainly contaminated by radioisotope 137Cs due to impurities introduced from the outside during the ascon manufacturing process. We studied characteristics of the radioactive waste to see whether this material would be proper enough to be disposed in Gyeongju LILW repository or be other ways to reduce the disposal volume including self-disposal before its final disposal otherwise. KORAD looked into the properness of characteristics of ascon in terms of WAC (Waste Acceptance Criteria) documented by KORAD that includes general chemical and physical properties of asphalt, density, size of grains, content of organic material and possibility of existence of chelate materials that qualitatively limited to be disposed by the criteria. And other associated characteristics such as gas generation and bio degradation were also investigated. Based on the data obtained from the study, we proposed various plausible solutions in associated with operational and disposal safety and economic view points. This study will be used for KORAD’s decision on how to control and safely dispose the spent ascon within a reasonable time period. And also those experiences may be applied for other LILW issues that require treatment or conditioning of radioactive wastes in the future.

The deep geological repository for high-level radioactive waste requires careful consideration due to its exceptionally long-term implications, making long-term impact assessments essential. However, evaluating the long-term effects of deep geological repositories using performance assessment models is accompanied by various sources of uncertainty, including uncertainties about the future, model uncertainties, and uncertainties associated with input data. These multifaceted uncertainties arise from factors such as a lack of current knowledge, contributing to a complex web of unpredictability. Managing, mitigating, and ultimately eliminating these uncertainties is crucial for ensuring the performance and safety of deep geological repositories. Currently, the Korea Radioactive Waste Agency (KORAD) is developing a complex behavior model that incorporates Thermal-Hydraulic-Mechanical-Chemical (THMC) phenomena within the disposal system using PFLOTRAN. To address model uncertainties and furthermore input data uncertainties for this intricate model, an automated sensitivity analysis system has been developed. This automated system operates without human intervention, facilitating tasks such as automatic parameter adjustments and the quantification of uncertainties. Furthermore, this system aids in identifying key factors characterized by substantial uncertainties. Through this system, it is possible to examine concentration distributions in each components of the deep disposal facility in response to changes in input data and to identify factors with significant uncertainties. The sensitivity results and key uncertainty factors obtained through this system are intended to be used for optimizing uncertainties in future research and development.

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.

Bentonite, primarily composed of montmorillonite, plays a crucial role as one of the engineering barrier materials in a deep geological repository (DGR). The advantageous properties of montmorillonite, such as its swelling capacity, low permeability, and low thermal conductivity, make it a key component as a buffer material for isolating high-level radioactive waste from the surrounding natural environment. It has been known that the stability of montmorillonite is influenced by a wide range of pressure-temperature-composition (P-T-X) conditions relevant to the DGR environment. When considering potential geological events, of notable concerns are its interactions with groundwater or seawater at elevated temperatures, leading to safety hazards within the system. In this study, therefore, we investigated the hydration and dehydration reactions of Ca-montmorillonite with saline fluids such as NaCl and KCl solutions at elevated pressures and temperatures by conducting in-situ X-ray diffraction experiments using both a capillary sample heating cell and a resistive-heated diamond anvil cell. As a result, we observed different hydration states of montmorillonite depending on the chemical composition of fluids, i.e., tri-hydrated layers in NaCl and bi-hydrated layers in KCl solutions, respectively. Furthermore, we identified that montmorillonite undergoes distinct stepwise dehydration with increasing temperature, and the dehydration temperature of montmorillonite significantly increases with increasing water pressure. Consequently, this study would provide insights into the stability of hydrated montmorillonite in a seawater-dominated fluid environment and its implications for the long-term safety of the disposal system.

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.

High level radioactive waste (HLW) final disposal repository is faced thermos-hydro-mechanical - radioactive condition because it is placed over 500 m in depth and waste emits decay heats for decades. Repository will be operated around 100 years and will be closed after all the wastes are disposed. The integrity of engineered barriers including buffer, backfill, concrete plug and canister and natural barrier (natural rock mass) will be stood during operating periods. Monitoring sensors for concrete and rock mass is conducted using piezo based sensors such as accelerometer or acoustic emission (AE) sensors. Typical accelerometer for harsh conditions is commonly expensive and data/power cable can be a potential groundwater inflow and nuclide outflow path. The fiber optic accelerometer whose data and power cable are united and has limited volume. Therefore, it can be a potential alternative sensor of piezo based sensors. The temperature limits and accelerated tests for fiber optic sensors are conducted. Most of sensors gives a malfunction around 130°C. The results of these experimental tests give a possibility of communications in compacted bentonite buffer and will be utilized for the design of monitoring systems for the repository.

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.

The high-level nuclear waste (HLW) repository disposes of high-level nuclear waste at a depth of 500 m to 1,000 m underground. Structural health monitoring must be accompanied by the complex environmental conditions of high temperature, high humidity, radiation, and mechanical stress. A thermocouple for measuring temperature, total stress meter and pore pressure meter for measuring stress and water pressure, relative hygrometer and electrical resistivity sensor (TDR or SUS) for measuring humidity, accelerometer for measuring crack signals, and strain gauge for measuring displacement are used. For safety, after disposing of HLW in the HLW repository, access to the disposal tunnel gets blocked, making it impossible to replace or remove the monitoring sensors. So, it is necessary to evaluate the effect of the HLW repository’s environmental conditions on the monitoring sensors and enhance their durability through quantitative life evaluation and shielding. Before evaluating the life of accelerometers and strain gauges used in the HLW repository, an experimental study is conducted to determine failure modes and failure mechanisms under radiation conditions, which are unique environmental conditions of the HLW repository.

The seven-year research project entitled “Development of workflow for integrated 3D geological site descriptive modeling” is being carried out from 2023. This research is funded by Ministry of Trade, Industry, and Energy (MOTIE). Progress of the research is discussed here. The integrated 3D geological SDM (site descriptive model; GSDM hereafter) consists of three part; 1) three dimensional representation of geologic elements, 2) database for material properties and modeling results from SDMs of other disciplines (e.g., rock mechanics), and 3) a visualization tool for geology, material properties and modeling results. The GSDM is comparable to the GDSMs of SKB and POSIVA in its representation of geology by volume of geologic elements. However, our GSDM is different in that extra information of material properties and an extra tool for visualization is included in the GDSM. The rationale for incorporating material properties and a visualization tool into the GSDM is to expedite the development of the GSDM and SDMs of other disciplines by allowing single institution to integrate database and visualization with the GSDM. SKUA-GOCAD is used for representation of geologic surfaces for ductile and brittle shear zones, and also for surfaces for delineation of volumes of rock units. We have adopted SKUAGOCAD because the program offers powerful functions of interpolation including borehole data and geophysical prospecting. So far, we have tested the program for five different geologies, including sedimentary, high-grade metamorphic, and intrusive igneous geology. The test results are promising. Incorporation of data and modeling results for the SDMs of other disciplines is at conceptual stage. The working conceptual model involves the following steps, 1) to provide the modeler of other disciplines with surface information representing geologic elements, 2) the modeler returns not only material properties but the results of numerical analysis, and 3) incorporation of material properties and modeling results into database. Since the numerical codes in other disciplines adopt different types of formats for 3D geology, we plan to adopt the widely used FEM format prepared by Gmsh. The visualization tool will also adopt Gmsh for graphical representation of 3D geology as well as database for material properties and modeling results. When the working model of GSDM becomes available, rapid and significant progress is expected in the SDMs of other disciplines and related areas, for example, geotechnical investigation for deep geological repository.

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 the nuclear environment, sensors ensure safety, monitoring, and operational efficiency under various operating conditions. These sensors come in various forms, each tailored to specific purposes, including nuclear safety and security, waste treatment and storage, gas leak detection, temperature and humidity monitoring, and corrosion detection. Ensuring the longevity of sensors without the need for frequent replacements is a vital goal for researchers in this field. This paper explores materials that can act as shields to protect sensors from harsh environmental conditions (high radiation and temperatures) to enhance their lifetime. The types of material that had been explored were divided into categories: metal and non-metal. Fourteen types of metal and seven different plastic materials were studied and focused on their characteristics and current applications. Considering properties like melting point, intensity, and conductivity, plastic materials are chosen to be examined as sensor shielding material. A preliminary experiment was conducted to verify signal characteristics changes by shielding material. Metal material and plastic material each were placed in the middle of the granite and the target sensor. The result showed that when metal is between the granite and the sensor, the density and impedance are higher in granite than in the metal. This leads to signal attenuation and a shift in resonance frequency, while plastic does not. Therefore, PPS (Polyphenylene sulfide) and PAI (Polyamide-imide) have lower density and impedance than granite while also possessing heat, moisture, and radiation resistance for effective shielding.

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 order to ensure the long-term safety of a deep geological repository, the performance assessment of the Engineered Barrier System (EBS) considering a thermal process should be performed. The maximum temperature at the side wall of a disposal canister for the technical design requirement should not exceed 100°C. In this study, the thermal modelling was conducted to analyze the effects of the thermal process from a disposal canister to the surrounding near-field host rock using the PFLOTRAN code. The mesh was generated using the LaGriT code and the material properties were assigned by applying the FracMan code. Initial conditions were set as the average geothermal gradient (25.7°C/km) and an average surface temperature (14.7°C) in Korea. The highest temperature was observed at the middle of the canister side wall. The temperature of the buffer was lower than that of the canister, and the temperature increase of the deposition tunnel and the host rock was insignificant due to the lower effect of the heat source. The result of the thermal evolution of the EBS represented the highest thermal effects in the vicinity of the canister. In addition, the thermal effects were largely decreased after 10 years of the entire simulation period. It demonstrated that the model took 3 years to heat up the buffer around the canister. The temperature at the canister side wall increased until 3 years and then decreased after that time. This is because that the radioactive decay heat from the heat source was emitted enough to raise the overall temperature of the EBS by 3 years. However, the decay heat rate of the canister decreased exponentially with the disposal time and then its decay heat was not emitted enough after 3 years. In conclusion, the peak temperature results of the EBS were lower than 70°C to meet the technical design requirement.

According to the second high-level radioactive waste management national basic plan announced in December 2021, the reference geological disposal concept for spent nuclear fuels (SNF) in Korea followed the Finnish concept based on KBS-3 type. Also, the basic plan required consideration of the development of the technical alternatives. Accordingly, Korea Atomic Energy Research Institute is conducting analyses of various alternative disposal concepts for spent nuclear fuels and is in the final selection stage of an alternative disposal concept. 10 disposal concepts including reference concept were considered for analysis in terms of disposal efficiency and safety. They were reference concept, mined deep borehole matrix, sub-seabed disposal, deep borehole disposal, multi-level disposal, space disposal, sub-sea bed disposal, long-term storage, deep horizontal borehole disposal, and ice-sheet disposal. Among them, first 4 concepts, mined deep borehole matrix, sub-seabed disposal, deep borehole disposal, multi-level disposal, were selected as candidate alternative disposal concepts by the evaluation of qualitative items. And then, by the evaluation of quantitative and qualitative items with specialists, multi-level disposal concept was being selected as a final alternative disposal concept. Design basis and performance requirements for designing alternative disposal systems were laid in the previous stage. Based on this, the design strategy and main design requirements were derived, and the engineered barrier system of a high-efficiency disposal concept was preliminary designed accordingly. In addition, as an alternative disposal concept, performance targets and related requirements were established to ensure that the high-efficiency repository system and its engineered barrier system components, such as disposal containers, buffer bentonites, and backfill perform the safety functions. Items that qualitatively describe safety functions, performance goals, and related requirements at this stage and items whose quantitative values are changed according to future test results will be determined and updated in the process of finalizing and specifically designing an alternative highefficiency disposal system.

This study aimed to provide better understanding of the bedrock aquifer bacterial communities and their functions in deep geological repository (DGR) environment. Two study sites of uranium deposits in the Ogcheon Metamorphic Belt were selected: Boeun and Guemsan. From two study sites, six groundwater samples were obtained with different boreholes and depths: OB1 (Boeun, 25 m), OB3 (Boeun, 80 m), GS1 (Guemsan, 25 m), GS2 (Guemsan, 85-90 m), GS3-I (Guemsan, 32- 38 m), GS3-II (Guemsan, 70-74 m). The physicochemical properties of groundwater were analyzed by multi-parameter sensors, ion chromatography (IC), and inductively coupled plasma optical emission spectroscopy (ICP-OES). Illumina Miseq sequencing was performed to investigate bacterial community in six groundwater samples. In addition, the number of sulfate-reducing bacteria (SRB) was quantified by a quantitative PCR (qPCR). Bacterial community composition varied in response to boreholes and depths. A total of 14 different phyla and 36 classes were detected from six groundwater samples. Overall, Proteobacteria, Actinomycetota, and Bacteroidota were dominant in the phylum level. SRB and iron-reducing bacteria (IRB) were detected in all groundwater samples even though organic carbon sources were not abundant (0.7-3.3 mg-total organic carbon/L). This result shows a potential to immobilize uranium in DGR environment. In particular, SRB, Desulfosporosinus fructosivorans and Humidesulfovibrio mexicanus were mainly detected in GS1 and GS2 groundwater samples, which attributed to higher dissimilatory sulfite reductase functional gene copy number in GS1 and GS2 groundwater samples. Statistical analysis was performed to understand the correlation between environmental factors and core bacterial species. Dissolved oxygen (DO), Fe, and Mn concentrations were positively correlated with Curvibacter fontanus while Undibacterium rivi had a negative correlation with pH. These results indicate that bacterial community could be changed in response to environmental variation. Further study with a greater number of samples is necessary to obtain statistically reliable and meaningful results for a safe DGR system.

The final disposal of Spent Nuclear Fuel (SNF) will take place in a deep geological repository. The metal canister surrounding the SNF is made of cast iron and copper, designed to provide longterm containment of radionuclides. Canister is intended to be safeguarded by a multiple-barrier disposal system comprising engineered and natural barriers. Colloids and gases are mediators that can accelerate radionuclide migration and influence radionuclide behavior when radionuclides leak from the canister at the end of its service life. It is very important to consider these factors in the assessment of the long-term stability of deep dispoal repository. An experimental setup was designed to observe the acceleration of nuclide behavior due to gas-mediated transport in a simulated environment with specific temperature and pressure conditions, similar to those of a deep disposal repository. In this study, experiments were conducted to simulate gas flow within an engineered barrier under conditions reflecting 1000 years post repository closure. The experiment utilized bentonite WRK with a dry density of 1.61 g/cm³ after compaction. The compacted bentonite was subsequently saturated under a water pressure of 5 MPa, equivalent to the hydrostatic pressure found 500 meters underground. Gas was introduced into the saturated bentonite at different pressures to assess the permeation behavior of the bentonite relative to gas pressure variations. Consequently, it was observed that under specific pressures, gas permeated the saturated bentonite, ascending in the form of bubbles. Furthermore, it was noted that when a continuous flow was initiated within the bentonite, erosion took place, leading to the buoyant transportation of eroded particles upward by the bubbles. The particles transported by the bubbles had a relatively small particle size distribution, and cesium also tended to be transported by the bubbles and moved upward. When high-pressure gas is generated at the interface of the canister and the buffer, flow through the buffer can occur, and cationic nuclides such as cesium and strontium can be attached to the gas bubble and migrate. However, the pressure of the gas to break through the saturated buffer is very high, and the amount of cesium transported by the gas bubbles is very limited.

Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SFs with their high radiotoxicity, heat generating, and long-lived radioactivity are actually the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own methodologies and management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SF has also been conducted during last couple of decades since around 1997, during which several various alternative type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide. The world first Finnish repository which is expected to begin to operate sooner or later will be also this type. Since the characteristics of SF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development has been being roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in nearfield, far-field, and even biosphere in and around the conceptual repository system. Through the study such scenarios and models has been implemented to development of a safety assessment by utilizing GoldSim development tool for a rough quantitative comparison with existing disposal options and simple illustration purpose as well as for showing how to develop and implementation of the model to GoldSim templet.

The thermal evaluations for the conceptual design of the deep geological repository considering the improved modeling of the spent fuel decay heat were conducted using COMSOL Multiphysics computational program. The maximum temperature at the surface of a disposal canister for the technical design requirement should not exceed 100°C. However, the peak temperature at the canister surface should not exceed 95°C considering the safety margin of 5°C due to several uncertainties. All thermal evaluations were based on the time-dependent simulation from the emplacement time of the canister to 100,000 years later. In particular, the heat source condition was set to the decay heat rate and axial decay heat profile of the PLUS7 fuel with 4.0wt% U-235 and 45 GWD/MTU. The thermal properties of the granitic rock in South Korea were applied to the host rock region. For the reference design case, the cooling time of the SNF was set to 40 years, the distance between the deposition holes 8 meters and that between the deposition tunnels 30 meters. However, the peak temperature at the canister surface at 10 years was 95.979°C greater than 95°C. This design did not meet the thermal safety requirement and needed to be modified. For the first modified case, when the distance between the deposition tunnels was set to 30 meters, three cooling time cases of 40, 50 and 60 years and five distances of 6, 7, 8, 9 and 10 meters between the deposition holes were considered. The design with the distances of 9 and 10 meters between the deposition holes for the cooling time of 40 years and all five distances for 50 and 60 years were less than 95°C. For the second modified case, when the distance between the deposition holes was set to 8 meters, three cooling time cases of 40, 50 and 60 years and five distances of 20, 25, 30, 35 and 40 meters between the deposition tunnels were considered. The design with the distances of 35 and 40 meters between the deposition tunnels for the cooling time of 40 years, the distances of 25, 30, 35 and 40 meters for 50 years and all five distances for 60 years were less than 95°C. As a result, the peak temperature at the canister surface decreased as the cooling time and the distance between the deposition holes and the tunnels increased.

Spent nuclear fuel management is a high-priority issue in South Korea, and addressing it is crucial for the country’s long-term energy sustainability. The KORAD (Korea Radioactive Waste Agency) is leading a comprehensive, long-term project to develop a safe and effective deep geological repository for spent nuclear fuel disposal. Within this framework, we have three primary objectives in this work. First, we conducted statistical analysis to assess the inventory of spent nuclear fuel in South Korea as of 2021. We also projected future generation rates of spent nuclear fuels to identify what we refer to as reference spent nuclear fuels. These reference spent nuclear fuels will be used as the design basis spent fuels for evaluating the safety of the repository. Specifically, we identified four types of design basis reference spent nuclear fuels: high and low burnup from PLUS7 (with a 16×16 array) and ACE7 (with a 17×17 array) assemblies. Second, we analyzed radioactive nuclides’ inventory, activities, and decay heats, extending up to a million years after reactor discharge for these reference spent nuclear fuels. This analysis was performed using SCALE/TRITON to generate the burnup libraries and SCALE/ORIGEN for source term evaluation. Third, to assess the safety resulted from potential radioactive nuclides’ release from the disposal canister in future work, we selected safety-related radionuclides based on the ALI (Annual Limit of Intake) specified in Annex 3 of the 2019-10 notification by the NSSC (Nuclear Safety and Security Commission). Conservative assumptions were made regarding annual water intake by humans, canister design lifetime, and aquifer flow rates. A safety margin of 10-3 of the ALI was applied. We selected 56 radionuclides that exceed the intake limits and have half-lives longer than one year as the safety-related radionuclides. However, it is crucial to note that our selection criteria focused on ALI and half-lives. It did not include other essential factors such as solubility limits, distribution coefficients, and leakage processes. So, some of these nuclides can be removed in a specific analysis area depending on their properties.