In nuclear power plant (NPP) decommissioning, ventilation and purification of the building atmosphere are important to create a working environment, ensure worker safety, and prevent the release of gaseous radioactive materials into the environment. The heating, ventilation, and air conditioning (HVAC) system of each building is maintained, modified, or newly installed. In this study, based on APR1400, operation strategies were presented in case of ventilation abnormalities in the reactor containment building (RCB), where highly radioactive particles and high dust are most frequently generated during NPP decommissioning. For research, it was assumed that the entire RCB atmospheric ventilation during decommissioning would use the RCB purge system of the existing NPP and perform continuous ventilation. Additionally, it is assumed that areas where high radiation particles and high dust occur locally, such as reactor containers or internal segments, are sealed with tents and purified using a HEFA filter of a temporary portable HVAC, and a exhaust flow path is connected to the discharge duct of the existing RCB purge system. The possibility of abnormal occurrence was largely divided into two cases. First, when large amounts of uncontrolled pollutants are released into the atmosphere inside the RCB, discharge to the environment is stopped manually or automatically by a modified engineered safety function activation signal (ESFAS). Afterwards, the RCB purge system should be operated in recirculation mode to sufficiently purify the RCB atmosphere with a HEPA filter. Second, when the first train of the low volume purge system is not running due to a failure, standby train should be operated. If both low volume purge trains fail, a high volume purge system is used. Intermittent purge operation is preferred due to large capacity during high volume purge operation. In cases where it is not possible to operate all purge systems due to common issues such as power supply, atmospheric sampling is performed to determine whether to proceed with the work inside RCB.

The radiological characterization of SSCs (Structure, Systems and Components) plays one of the most important role for the decommissioning of KORI Unit-1 during the preparation periods. Generally, a regulatory body and laws relating to the decommissioning focus on the separation and appropriate disposal or storage of radiological waste including ILW (intermediate level waste), LLW (low level waste), VLLW (very low level waste) and CW (clearance waste), aligned with their contamination characteristics. The result of the preliminary radiological characterization of KORI Unit-1 indicated that, apart from neutron activated the RV (reactor vessel), RVI (reactor vessel internals), and BS (biological shielding concrete), the majorities of contamination were sorted to be less than LLW. Radiological contamination can be evaluated into two methods. Due to the difficulties of directly measuring contamination on the interior surfaces of the pipe, called CRUD, the assessment was implemented by modeling method, that is measuring contamination on the exterior surfaces of the pipes and calculating relative factors such as thickness and size. This indirect method may be affected by the surrounding radiation distribution, and only a few gamma nuclides can be measured. Therefore, it has limitation in terms of providing detailed nuclide information. Especially, α and β nuclides can only be estimated roughly by scaling factors, comparing their relative ratios with the existing gamma results. To overcome the limitation of indirect measurement, a destructive sampling method has been employed to assess the contamination of the systems and component. Samples are physically taken some parts of the systems or components and subsequently analyzed in the laboratory to evaluate detailed nuclides and total contamination. For the characterization of KORI Unit-1, we conducted the radiation measurement on the exterior surfaces of components using portable instruments (Eberline E-600 SPA3, Thermo G20-10, Thermo G10, Thermo FH40TG) at BR (boron recycle system) and SP (containment spray system) in primary system. Based on these results, the ProUCL program was employed to determine the destructive sample collection quantities based on statistical approach. The total of 5 and 8 destructive sample quantities were decided by program and successfully collected from the BR and SP systems, respectively. Samples were moved to laboratory and analyzed for the detail nuclide characteristics. The outcomes of this study are expected to serve as valuable information for estimating the types and quantities of radiological waste generated by decommissioning of KORI Unit-1.

The Derived Concentration Guideline Level (DCGL) is required to release the facility from the nuclear safety act at the stage of site restoration of the decommissioning nuclear power plant. In order to evaluate DCGL, there are various requirements, and among them, the selection of input parameters based on the application scenario is the main task. Especially, it is important to select input parameters that reflect site characteristics, and at this time, a single deterministic value or a probabilistic distribution can be applied. If it is inappropriate to apply a particular single value, it may be reasonable to apply various distributions, and the RESRAD code provides for evaluation using probabilistic methods. Therefore, this study aims to analyze the difference between the application of the deterministic method and the application of the probabilistic method to the area and thickness of the contaminated zone among the site characteristics data. This study analyzed the thickness and area of the contaminated zone, and in the case of thickness, the deterministic method was applied by changing the thickness at regular intervals from the minimum depth considered by MARSSIM to the thickness of the unsaturated zone identified in previous research data. In addition, a probabilistic analysis was performed by applying a distribution to the thickness of contaminated zone. Second, for the area of the contaminated zone, the dose was evaluated for each area in consideration of the areas to be considered when deriving Area Factor (AF), and the resulting change in DCGL was observed. As a result, the DCGL tends to decrease as the thickness increases, and it seems to be saturated when the thickness exceeds a certain thickness. Therefore, It was confirmed that the level of saturated values is similar to that of entering a probabilistic distribution, and in the case of a parameter that is reasonable to enter as a distribution rather than as a single value, it is sufficiently conservative to perform a probabilistic evaluation. In the case of area change, the DCGL evaluation result showed that the DCGL increased as the scale decreased. The magnitude of the change varies depending on the characteristics of each radionuclide, and in the case of radionuclides where external exposure gamma rays have a major exposure effect, the change is relatively small. It can be seen that the change in DCGL according to the area has the same tendency as the AF applicable to the survey unit for small survey units applied in the final status survey.

As unit 1 of Kori was permanently shut down in June 2017, domestic nuclear industry has entered the path of decommissioning. The most important thing in decommissioning is cost reduction. And volume reduction of radioactive waste is especially important. According to the IAEA report, more than 4,000 tons of metallic waste is generated during the decommissioning of a 1,000 MWe reactor and most of these wastes are LLW or VLLW. To reduce amount of metallic waste dramatically, we should choose efficient decontamination method. In this study, we conducted dry ice and bead blasting decontamination. We prepared Inconel-600 and STS-304 specimen with dimensions of 30 mm × 30 mm × 5 mm. Loose and fixed contamination was applied on the surface of specimen using SIMCON method. Bead and dry-ice blasting was conducted by spraying alumina and dry ice pellet at the same pressure and distance for the same time. The removal of loose contamination was observed using microscope. It was found that contaminants are significantly removed using both dry ice blasting and bead blasting. However, some abrasive material remained on the surface of specimen. The removal of fixed contamination was verified by weight comparison before and after experiment and cobalt concentration comparison before and after experiment using X-ray Fluorescence Spectroscope (XRF). At least 90% of the cobalt was removed, but some abrasive particle was also remained on the surface of specimen. In this study, it is confirmed that the effectiveness of manufacturing a large-scale abrasive decontamination facility, and it is expected that this technology can be used to effectively reduce the amount of metallic waste generated during decommissioning.

Domestic commercial low- and intermediate-level radioactive waste storage containers are manufactured using 1.2 mm thick cold-rolled steel sheets, and the outer surface is coated with a thin layer of primer of 10~36 μm. However, the outer surface of the primer of the container may be damaged due to physical friction, such as acceleration, resonance, and vibration during transportation. As a result, exposed steel surfaces undergo accelerated corrosion, reducing the overall durability of the container. The integrity of storage containers is directly related to the safety of workers. Therefore, the development of storage containers with enhanced durability is necessary. This paper provides an analysis of mechanical properties related to the durability of WC (tungsten carbide)-based coating materials for developing low- and intermediate-level radioactive waste storage containers. Three different WC-based coating specimens with varied composition ratios were prepared using HVOF (high-velocity oxy-fuel) technique. These different specimens (namely WC-85, WC-73, and WC-66) were uniformly deposited on cold-rolled steel surfaces ensuring a constant thickness of 250 μm. In this work, the mechanical properties of the three different WCbased coaitng materials evaluated from the viewpoints of microstructure, hardness, adheision force between substrate and coating material, and wear resistance. The cross-sectional SEM-EDS (Scanning Electron Microscope-Energy Dispersive X-ray Spectroscopy) images revealed that elements W (tungsten), C (carbon), Ni (nickel), and Cr (chromium) were uniformly distributed within the each coating layers which was approximately 250 μm thick. The average hardness values of HWC-85 and HWC-73 were found to be 1,091 Hv (Vickers Hardness) and 1,083 Hv, respectively, while the HWC-66 exhibited relatively lower hardness value of 883 Hv. This indicates that a higher WC content results in increased hardness. Adhesion force between and substrates and coating materials exceeded 60 MPa for all specimens, however, there were no significant differences observed based on the tungsten carbide content. Furthermore, a taber-type abrasion tester was used for conducting abrasion resistance tests under specific conditions including an H-18 load weight at 1,000 g with rotational speed set at 60 RPM. The abrasion resistance tests were performed under ambient temperatures (RT: 23±2°C) as well as relative humidity levels (RH: 50±10%). Currently, the ongoing abrasion resistance tests will include some results in this study.

As the acceptance criteria for low-intermediate-level radioactive waste cave disposal facilities of Korea Radioactive Waste Agency (KORAD) were revised, the requirements for characterization of whether radioactive waste contains hazardous substances have been strengthened. In addition, As the recent the Nuclear Safety and Security Commission Notice (Regulations on Delivery of Low- Medium-Level Radioactive Waste) scheduled to be revised, the management targets and standards for hazardous substances are scheduled to be specified and detailed. Accordingly, the Korea Atomic Energy Research Institute (KAERI) needs to prepare management methods and procedures for hazardous substances. In particular, in order to characterize the chemical requirements (explosiveness, ignitability, flammability, corrosiveness, and toxicity) contained in radioactive waste, it must be proven through documents or data that each item does not contain hazardous substances, and quality assurance for the overall process must be provided. In order to identify the characteristics of radioactive waste that will continue to be generated in the future, KAERI needs to introduce a management system for hazardous substances in radioactive waste and establish a quality assurance system. Currently, KAERI is thoroughly managing chelates (EDTA, NTA, etc.), but the detailed management procedures for hazardous substances related to chemical requirements in radioactive waste in the radiation management area specified above are insufficient. The KAERI’s Laboratory Safety Information Network has a total periodic regulatory review system in place for the purchase, movement, and disposal of chemical substances for each facility. However, there is no documents or data to prove that the hazardous substances held in the facility are not included in the radioactive waste, and there are no procedures for managing hazardous substances. Therefore, it is necessary to establish procedures for the management of hazardous substances, and we plan to prepare management procedures for hazardous substances so that chemical substances can be managed according to the procedures at each facility during preliminary inspection before receiving radioactive waste. The procedure provides definitions of terms and types of management targets for each characteristic of the chemical requirements specified above (explosiveness, ignition, flammability, corrosiveness, and toxicity). In addition, procedure also contains treatment methods of radioactive waste generated by using hazardous substances and management methods of in/out, quantity, history of that substances, etc. As the law is revised in the future, management will be carried out according to the relevant procedures. In this study, we aim to present the hazardous substance management procedures being established to determine whether radioactive waste contains hazardous substances in accordance with the revised the notice and strengthened acceptance criteria. Through this, we hope to contribute to improving reliability so that radioactive waste could be disposed of thoroughly and safely.

This study focuses on the development of coatings designed for storage containers used in the management of radioactive waste. The primary objective is to enhance the shielding performance of these containers against either gamma or neutron radiation. Shielding against these types of radiation is essential to ensure the safety of personnel and the environment. In this study, tungsten and boron cabide coating specimens were manufactured using the HVOF (High-Velocity Oxy Fuel) technuqe. These coatings act as an additional layer of protection for the storage containers, effectively absorbing and attenuating gamma and neutron radiation. The fabricated tungsten and boron carbide coating specimens were evaluated using two different testing methods. The first experiment evaluates the effectiveness of a radiation shielding coating on cold-rolled steel surfaces, achieved by applying a mixture of WC (Tungsten Carbide) powders. WC-based coating specimens, featuring different ratios, were prepared and preliminarily assessed for their radiation shielding capabilities. In the gamma-ray shielding test, Cs-137 was utilized as the radiation source. The coating thickness remained constant at 250 μm. Based on the test results, the attenuation ratio and shielding rate for each coated specimen were calculated. It was observed that the gammaray shielding rate exhibited relatively higher shielding performance as the WC content increased. This observation aligns with our findings from the gamma-ray shielding test and underscores the potential benefits of increasing the tungsten content in the coating. In the second experiment, a neutron shielding material was created by applying a 100 μm-thick layer of B4C (Boron Carbide) onto 316SS. The thermal neutron (AmBe) shielding test results demonstrated an approximate shielding rate of 27%. The thermal neutron shielding rate was confirmed to exceed 99.9% in the 1.5 cm thick SiC+B4C bulk plate. This indicates a significant reduction in required volume. This study establishes that these coatings enhance the gamma-ray and neutron shielding effectiveness of storage containers designed for managing radioactive waste. In the future, we plan to conduct a comparative evaluation of the radiation shielding properties to optimize the coating conditions and ensure optimal shielding effectiveness.

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 radwaste repository consists of a multi-barrier, including natural and engineered barriers. The repository’s long-term safety is ensured by using the isolation and delay functions of the multi-barrier. Among them, natural barriers are difficult to artificially improve and have a long time scale. Therefore, in order to evaluate its performance, site characteristics should be investigated for a sufficient period using various analytical methods. Natural barriers are classified into lithological and structural characteristics and investigated. Structural factors such as fractures, faults, and joints are very important in a natural barrier because they can serve as a flow path for groundwater in performance evaluation. Considering the condition that the radioactive waste repository should be located in the deep part, the drill core is an important subject that can identify deep geological properties that could not be confirmed near the surface. However, in many previous studies, a unified method has not been used to define the boundaries of structural factors. Therefore, it is necessary to derive a method suitable for site characteristics by applying and comparing the boundary definition criteria of various structural factors to boreholes. This study utilized the 1,000 m deep AH-3 and DB-2 boreholes and the 500 m deep AH-1 and YS- 1 boreholes drilled around the KURT (KAERI Underground Research Tunnel) site. Methods applied to define the brittle structure boundary include comparing background levels of fracture and fracture density, excluding sections outside the zone of influence of deformation, and confining the zone to areas of concentrated deformation. All of these methods are analyzed along scanlines from the brittle structure. Deriving a site-specific method will contribute to reducing the uncertainties that may arise when analyzing the long-term evolution of brittle structures within natural barriers.

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 occurrence of shear failure in a rock mass, resulting from the sliding of joint surfaces, is primarily influenced by the surface roughness and contact area of these joints. Furthermore, since joints serve as crucial conduits for the movement of water, oil, gas, and thermal energy, the aperture and geometric complexity of these joints have a significant impact on the hydraulic properties of the rock mass. This renders them critical factors in related industries. Therefore, to gain insights into the mechanical and hydraulic behavior of a rock mass, it is essential to identify the key morphological characteristics of the joints mentioned above. In this study, we quantified the morphological characteristics of tensile fractures in granitic rocks using X-ray CT imaging. To accomplish this, we prepared a cylindrical sample of Hwang-Deung granite and conducted splitting tests to artificially create tensile fractures that closely resemble rough joint surfaces. Subsequently, we obtained 2D sliced X-ray CT images of the fractured sample with a pixel resolution of approximately 0.06 mm. By analyzing the differences in CT numbers of the rock components (e.g., fractures, voids, and rock matrix), we isolated and reconstructed the geometric information of the tensile fracture in three dimensions. Finally, we derived morphological characteristics, including surface roughness, contact area, aperture, and fracture volume, from the reconstructed fracture.

Engineered Barrier Systems (EBS) are a key element of deep geological repositories (DGR) and play an important role in safely isolating radioactive materials from the ecosystem. In the environment of a DGR, gases can be generated due to several factors, including canister corrosion. If the gas production rate exceeds the diffusion rate, pore pressures may increase, potentially inducing structural deterioration that impairs the function of the buffer material. Therefore, understanding the hydraulic-mechanical behavior of EBS due to gas generation is essential for evaluating the longterm stability of DGR. This study employed X-ray computed tomography (CT) technology to observe cracks created inside the buffer material after laboratory-scale gas injection experiments. After CT scanning, we identified cracks more clearly using an image analysis method based on machine learning techniques, enabling us to examine internal crack patterns caused by gas injection. In the samples observed in this study, no cracks were observed penetrating the entire buffer block, and it was confirmed that most cracks were created through the radial surface of the block. This is similar to the results observed in the LASGIT field experiment in which the paths of the gas migration were observed through the interface between the container and the buffer material. This study confirmed the applicability of high-resolution X-ray CT imaging and image analysis techniques for qualitative analysis of internal crack patterns and cracks generated by gas breakthrough phenomena. This is expected to be used as basic data and crack analysis techniques in future research to understand gas migration in the buffer material.

The buffer is installed around the disposal canister, subjected to heating due to decay heat while simultaneously experiencing expansion influenced by groundwater inflow from the surrounding rock. The engineering barrier system for deep geological disposal require the evaluation of longterm evolution based on the verification of individual component performance and the interactions among components within the disposal environment. Thus, it is crucial to identify the thermalhydro- mechanical-chemical (THMC) processes of the buffer and assess its long- and short-term stability based on these interactions. Therefore, we conducted experimental evaluations of saturationswelling, dry heating, gas transport, and mineralogical alterations that the buffer may undergo in the heated-hydration environment. We simulated a 310 mm-thick buffer material in a cylindrical form, simulating the domestic disposal system concept of KRS+ (the improved KAERI reference disposal system for spent nuclear fuel), and subjected it to the disposal environment using heating cartridges and a hydration system. To monitor the thermal-hydro-mechanical behavior within the buffer material, load cells were installed in the hydration section, and both of thermal couples and relative humidity sensors were placed at regular intervals from the heat source. After 140 days of heating and hydration, we dismantled the experimental cell and conducted post-mortem analyses of the samples. In this post-mortem analysis, we performed functions of distance from the water contents, heat source, wet density, dry density, saturation, and X-Ray diffraction analysis (XRD). The results showed that after 140 days in the heated-hydration environment, the samples exhibited a significant decrease water contents and saturation near the heat source, along with very low wet and dry densities. XRD Quantitative Analysis did not indicate mineralogical changes. The findings from this study are expected to be useful for input parameters and THMC interaction assessments for the long-term stability evaluation of buffer in deep geological disposal.

Rock discontinuities in underground rock behave as weak planes and affect the safety of underground structures, such as high-level radioactive waste disposal and underground research facilities. In particular, rock discontinuities can be a main flow path of groundwater and induce large deformation caused by stress disturbance or earthquakes. Therefore, it is essential to investigate the characteristics of rock discontinuities considering in-situ conditions when constructing highlevel radioactive waste disposal, which needs to assure the long-term safety of the structure. We prepared Hwang-Deung granite rock block specimens, including a saw-cut rock surface, to perform multi-stage direct shear tests as a preliminary study. In the multi-stage direct shear tests, we can exclude possible errors induced by different specimens for obtaining a full failure envelope by using an identical specimen. We applied the initial normal stress of 3 MPa on the specimen and increased the normal stress to 5 and 10 MPa step by step after peak shear stress observation. We obtained the mechanical properties of saw-cut rock surfaces from the experiments, including friction coefficient and cohesion. Additionally, we investigated the effect of filling material between rock discontinuities, assuming the erosion and piping phenomenon in the buffer material of the engineering barrier system. When the filling material existed in the rock surfaces, the shear characteristics deteriorated, and the effect of bentonite was dominant on the shear behavior.

The engineered barrier system (EBS), composed of spent nuclear fuel, canister, buffer and backfill material, and near-field rock, plays a crucial role in the deep geological repository for high-level radioactive waste. Understanding the interactions between components in a thermo-hydro-mechanical -chemical (THMC) environment is necessary for ensuring the long-term performance of a disposal facility. Alongside the research project at KAERI, a comprehensive experimental facility has been established to elucidate the comprehensive performance of EBS components. The EBS performance demonstration laboratory, which installed in a 1,000 m2, consists of nine experimental modules pertaining to rock mechanics, gas migration, THMC characteristics, buffer-rock interaction, buffer & backfill development, canister corrosion, canister welding, canister performance, and structure monitoring & diagnostics. This facility is still conducting research on the engineering properties and complex interactions of EBS components under coupled THMC condition. It is expected to serve as an important laboratory for the development of the key technologies for assessing the long-term stability of engineered barriers

Due to the necessity of isolating spent nuclear fuel (SNF) from the human life zone for a minimum of 106 years, deep geological disposal (DGD) has emerged as a prominent solution for SNF management in numerous countries. Consequently, the resilience of disposal canisters to corrosion over such an extended storage period becomes paramount. While copper exhibits a relatively low corrosion rate, typically measured in millimeters per million years, in geological environment, special attention must be directed towards verifying the corrosion resistance of copper canister welds. This validation becomes inevitable during the sealing of the disposal canister once SNFs are loaded, primarily because the weld zone presents a discontinuous microstructure, which can accelerate both uniform and localized corrosion processes. In this research, we conducted an in-depth analysis of the microstructural characteristics of copper welds manufactured by TIG-based wire are additive manufacturing, which is ideal for welding relatively large structures such as a disposal canister. To simulate the welds of copper canister, a 12 mm thick oxygen-free plate was prepared and Y and V grooves were applied to perform overlay welding. Both copper welding zones were very uniform, with negligible defects (i.e., void and cracks), and contained relatively large grains with columnar structure regardless of groove types. For improving microstructures at welds with better corrosion resistance, the effect of preheat temperature also investigated up to 600°C.

The Barcelona Basic Model (BBM) is an elasto-plastic model used to describe the coupled thermo-hydro-mechanical behaviors of unsaturated soil. BBM is frequently adopted to model the unique swelling behavior of bentonite, which is generally considered as the buffer material between the host rock and the canister containing high-level radioactive waste in deep geological repositories, under the changing thermal, hydraulic, mechanical and chemical conditions during the lifetime of repository. Therefore, a variety of the continuum-based numerical methods tried to add the BBM for modelling the multi barrier systems of geological repository and succeeded to describe the elasto-plastic deformation of bentonite. However, to demonstrate the entire barrier systems the host rock should be modelled simultaneously with the buffer materials, and the continuum-based methods may be limited in their ability to reflect the fracture networks in the host rock which could be the major flow channels of groundwater. This research applies BBM in 3DEC, a three-dimensional block-based discrete element method, and validates the model by comparing the change of specific volume and mean effective stress during three numerical test cases. Discontinuum-based numerical methods with BBM can be extended to describe the coupled thermo-hydro-mechanical processes of multi-barrier systems in geological repositories, with a focus on the interaction between the host rock and bentonite.

The natural barrier system surrounding the geological repository for high-level radioactive waste plays a crucial role in preventing or delaying the leakage of radionuclides. Therefore, the natural barrier should ensure low permeability to prevent groundwater flow into the engineered barrier system throughout the repository’s lifetime. Crystalline rock, considered as the host rock for the geological repository in Korea, exhibits low intact rock permeability, but the crystalline rock often contains the multiple discontinuities due to its high brittleness that can allow the unexpected fluid flow. Therefore, the long-term hydraulic behavior of the discontinuity should be characterized while considering additional thermal, mechanical, and chemical effects. In comparison to thermal, hydraulic, and mechanical processes, the chemical processes on the discontinuities progress relatively slowly, resulting in limited researches to include these chemical processes. This research introduces mechanisms the involving coupled thermal-hydraulic-mechanicalchemical processes focusing on the rough fracture surfaces and asperities. The chemically-induced changes in mechanical and hydraulic properties are described based on pressure solution and precipitation concepts. A comprehensive review of laboratory tests, field tests, and numerical simulations is conducted related to the chemically-induced coupled processes in fractured rock. Laboratory tests, in particular, concentrate on microscopic changes in fracture asperities induced by pressure solution to analyze chemically-induced aperture changes. The TOUGHREACT, an integral finite difference method program for thermal-hydraulic-chemical simulations, is generally employed to model the chemical response of pressure solution and precipitation on fracture surfaces. The TOUGHREACT includes a module to describe effective porosity and permeability changes based on the modified cubic law, so the real-time change of the fracture permeability can be reflected during the flow simulation. Considering the coupled thermal-hydraulic-mechanicalchemical processes of discontinuity, it becomes evident that the chemical processes under repository conditions (long-term, high temperature, and high pressure) can disturb the hydraulic performance of the natural barrier, so further research is required to characterize the chemically-induced coupled processes for assessing the long-term performance of the natural barrier system.

The compacted bentonite buffer is a key component of the engineered barrier system in deep geological repositories for high-level radioactive waste disposal. Groundwater infiltration into the deep geological repository leads to the saturation of the bentonite buffer. Bentonite saturation results in bentonite swelling, gelation and intrusion into the nearby rock discontinuities within the excavation damaged zone of the adjacent rock mass. Groundwater flow can result in the erosion and transport of bentonite colloids, resulting in bentonite mass loss which can negatively impact the long-term integrity and safety of the overall engineered barrier system. The hydro -mechanicalchemical interactions between the buffer, surrounding host rock and groundwater influence the erosion characteristics of the bentonite buffer. Hence, assessing the critical hydro-mechanicalchemical factors that negatively affect bentonite erosion is crucial for the safety design of the deep geological repository. In this study, the effects of initial bentonite density, aperture, discontinuity angle and groundwater chemistry on the erosion characteristics of Bentonil WRK are investigated via bentonite extrusion and artificial fracture experiments. Both experiments examine bentonite swelling and intrusion into simulated rock discontinuities; cylindrical holes for bentonite extrusion experiments and plane surfaces for artificial fracture experiments. Compacted bentonite blocks and bentonite pellets are manufactured using a compaction press and granulation compactor respectively and installed in the transparent extrusion cells and artificial fracture cells. The reference test condition is set to be 1.6 g/cm3 dry density and saturation using distilled water. After distilled water or solution injection, the axial and radial expansion of the bentonite specimens into the simulated rock discontinuities are monitored for one month under free swelling conditions with no groundwater flow. Subsequent flow tests are conducted using the artificial fracture cell to determine the critical flow rate for bentonite erosion. The intrusion and erosion characteristics are modelled using a modified hydro-mechanicalchemical coupled dynamic bentonite diffusion model and a fluid-based hydro-mechanical penetration model.

Buffer materials play an important role in preventing the leakage of radionuclides from the residue. The mineralogical properties of these buffer materials are critical in repository design. This study presents the fundamental properties of Na-type MX80 and a novel Ca-type Bentonil- WRK. The CaO to MgO ratio in Bentonil-WRK was approximately 1:1, and the CaO to Na2O ratio was approximately 2.8:1. These results suggest that Bentonil-WRK demonstrates a lower swelling index compared to Gyeongju bentonite due to its CaO-to-MgO ratio’s proximity to 1:1, despite having a higher montmorillonite content than Gyeongju bentonite. The results of this research can provide useful foundational data for the evaluation of the thermal-hydraulic-mechanical-chemical behavior of buffer materials.