Wolsong Unit 1 nuclear power plant, which was permanently shut down in 2019, has a 678 MWe calandria vessel of the CANDU-6 type pressurized heavy-water reactor model. The calandria inside the vault is a horizontal cylindrical vessel made of stainless steel with a length of 7.8 m and a thickness of 28.6 mm. For the entire dismantling processes of a nuclear power plant, dismantling works cannot be performed using only one cutting technology and method, and when performing dismantling of a calandria vessel, various systems and components can be used for cutting and dismantling. The calandria vessel is located in a concrete compartment called a vault, and in order to safely dismantle the calandria vessel, the spread of radioactive contaminants from inside of the vault to the outside must be prevented. We designed dismantling processes using the laser cutting method to dismantle the calandria vessel and end shields. We must minimize the risk of internal radiation exposure to workers from aerosols derived from the thermal cutting processes. Therefore, we need a way to prevent secondary contamination from spreading outside the vault and within the reactor building. The path through which radioactive contaminants move is that the flying airborne products generated during the cutting process inside the vault where the calandria is located do not stay in place but spread outward through the opening of the RM-Deck structure at the top. Therefore, facilities or devices are needed to effectively prevent the spread of radioactive contaminants by blocking the expected movement path. By using these facilities or devices, it is possible to prevent the movement of radioactive aerosol particles between the location of the worker and the location of the cutting area where the calandria is located, thereby preventing internal exposure through the worker’s breathing. In addition, by using these, the cutting area where airborne pollutants are generated can be designed as an isolated work space to prevent the spread of radioactive contaminants. In this study, we propose a method of facilities for confining radioactive aerosol particles and preventing the spread of contamination when thermal cutting of the calandria vessel within the vault.

During the operation of nuclear power plant (NPP), the concentrates and spent resin are generated. They show relatively high radioactivity compared to other radioactive waste, such as dry active waste, charcoals, and concrete wastes. The waste acceptance criteria (WAC) of disposal facility defines the structure and property of treated waste. The concentrates and spent resin should be solidified or packaged in high integrity container (HIC) to satisfy the WAC in Korea. The Kori NPP has stored history waste. The large concrete package with solidified concentrates and spent resin. The WAC requires identification of 18 properties for the radioactive waste. Since some of the properties are not clearly identified, the large concrete packages could not satisfy the WAC in this moment. The generation of the large concrete package (rectangular type and cylindrical type), pretreatment of the package, treatment of inner drum, process development for clearance waste, etc. will be discussed in this paper. In addition, the conceptual design of whole treatment process will be discussed.

Hydrogen isotope separation involves the separation of hydrogen, deuterium, tritium, and their isotopologues. It is an essential technology for removing radioactive tritium contamination and for obtaining valuable hydrogen isotope resources. Among various hydrogen isotope separation technologies, water electrolysis technology exhibits a high separation factor. Consequently, the electrolysis of tritiated water is of paramount importance as a tritium enrichment method for treating tritium-contaminated water and for analyzing tritium in environmental samples. More recently, hydroelectrolysis technology, which utilizes proton exchange membranes (PEM) to reduce water inventory, has gained favor over traditional alkaline hydroelectrolysis. Nevertheless, it is crucial to decrease the hydrogen permeability of the PEM in order to mitigate the explosion risk associated with tritium hydrogen electrolysis devices. Additionally, efforts are needed to enhance the hydrogen isotope selectivity of the PEM and optimize the manufacturing process of the membrane-electrode assembly (MEA), thereby improving both hydrogen isotope separation performance and water electrolysis efficiency. In this presentation, we will delve into two key aspects. Firstly, we’ll explore the reduction of hydrogen permeability and the enhancement of the hydrogen isotope separation factor in PEM through the incorporation of 2D nanomaterial additives. Secondly, we’ll examine the influence of various MEAs preparation methods on electrolysis and isotope separation performances. Lastly, we will discuss the effectiveness of the developed system in separating deuterium and tritium.

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.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR 1&2), the first research reactors in South Korea, began in 1997 and the decommissioning status is currently proceeding with phase 3. It is expected that more than 5,000 tons of dismantled wastes will be generated as the contaminated building is demolished. Since these dismantled wastes must be disposed of in an efficient method considering economic feasibility, it is desirable to clearance extremely low-level wastes whose contamination is so minimal that the radiological risk is negligible. In Korea, in order to approve the clearance of radioactive waste, it must be proven that the nuclide concentration standards are met or that the dose to individuals and collectives is below the allowable dose value. At the KRR 1&2 decommissioning site, dismantled wastes have been steadily being disposed of through clearance procedure since 2021. Clearance was approved by the Korean Institute of Nuclear Safety (KINS) for one case of concrete waste in 2021 and two cases of metal waste in 2022. In 2023, the clearance of metal waste and asbestos waste has been approved so far, and in particular, this is the first case in Korea for asbestos waste. In this study, we compared the dose assessment methods and results of clearance wastes at the KRR 1&2 decommissioning site from 2021 to present. Dose assessment was conducted by applying the landfill scenario for concrete and asbestos and the recycling scenario for metal waste. The calculation codes used were RESRAD-onsite 7.2 and RESRAD-recycle 3.10. The dose conversion factors (DCF) for each age group (infant, 1y, 5y, 10y, 15y, adult) of the target nuclide used the values presented in ICRP-72, and in particular, geo-hydrological data of the actual landfill site was used as an input factor when evaluating landfill scenarios. As a result of the dose assessment, when landfilling concrete wastes in 2020, the personal dose and collective dose were evaluated the most at 2.80E+00 μSv/y and 4.83E-02 man·Sv/y, respectively.

At domestic nuclear power plant, concrete containers are stored to store waste generated before waste acceptance criteria (WAC) was established. Concrete container store concentrated waste liquid and waste resin. In order to disposal radioactive waste to a disposal site, it is necessary to conduct a characteristic evaluation inside the waste to check whether it satisfies the WAC. Two types of concrete containers are stored: round and square. The round type is filled with one 200-liter drum, and the square type is filled with four 200-liter drums. In the case of a round shape, the top lid is fastened with bolts, so it is possible to collect samples after opening the top lid without the need for additional equipment. However, in the case of a square shape, there is no top lid, and concrete is poured to cure the lid, so the separate equipment for characteristic evaluation is required. It is necessary to install a workstation for sample collection on the top of the concrete container, equipment for coring the top of the concrete container, and a device to prevent concrete dust scattering. Currently, the design of equipment for evaluating the characteristics of concrete containers has been completed, and equipment optimization through mock-up test will be performed in the future.

Republic of Korea is preparing to decommission Kori Unit 1 and Wolsong Unit 1. Decommissioning of a nuclear power plant proceeds in the following stages: shutdown, transition period, decontamination, cutting, waste treatment, and site restoration. When nuclear power plant is decommissioned, It is expected that approximately 80,000 drums of radioactive waste will be generated per nuclear power plant. Therefore, various technologies are being researched and developed to reduce this to approximately 14,500 drums. Technologies for waste volume reduction are largely mechanical and electrical/thermal methods. Representative examples of mechanical volume reduction technologies include super compactors and electrical/thermal volume reduction technologies include induction and plasma torch furnaces. Both technologies are effective reduction technologies, but the reduction ratio varies depending on the type or condition of waste before treatment. For example, as a result of testing waste reduction using a super compactor at NUKEM in Germany, the reduction ratio was found to be between 1.3 and 7 depending on the type or condition of waste such as chips, ash, scrap metal, sand, etc. And according to IAEA-TECDOC-1527, when reducing the volume of metals, aluminum, lead, copper, brass, etc. using induction melting, the waste volume reduction ratio is 5 to 20. In this paper, referring to these results, a melting test was conducted using a previously developed plasma torch with an output of more than 100 kW. And volume reduction characteristics of this plasma torch was considered depending on waste type or condition.

In general, radioactive waste with high radioactivity is made into a solid form with performance such as leaching restriction, shape retention, and structural stability so that radioactive waste does not affect humans and the environment as much as possible. This should be applied equally to radioactive waste, whether homogeneous or heterogeneous. The requirements are stipulated in the “Low and Intermediate Level Radioactive Waste Delivery Regulations” notice of the Korea Nuclear Safety and Security Commission. On the other hand, the waste acceptance criteria for domestic disposal facilities require immobilization of heterogeneous waste when the activity concentration is above a certain level, but do not provide specific immobilization performance requirements. In this study, the immobilization requirements applied to heterogeneous radioactive waste in various overseas countries operating low and intermediate-level radioactive waste disposal facilities were studied. First, the IAEA’s safety standards for radioactive waste immobilization, domestic regulations, and disposal facility waste acceptance criteria were reviewed. Countries operating surface disposal facilities such as the United States, France, Spain, and Japan and countries operating underground disposal facilities such as Sweden and Finland were divided to review the current status of immobilization application to heterogeneous waste in overseas countries. When reviewing overseas cases, each country’s disposal methods, types of disposal waste, and waste treatment criteria were also reviewed. It was found that the immobilization requirements for heterogeneous radioactive waste vary depending on the disposal method and the type of barrier used to ensure disposal safety in each country. The common point is to surround heterogeneous radioactive waste within a concrete lining of a certain thickness, and to apply the thickness, compressive strength, and diffusion coefficient of the concrete lining as immobilization performance requirements. Through this study, the immobilization performance requirements for heterogeneous radioactive waste in various overseas countries that stably operate low- and intermediate-level radioactive waste were confirmed, which is expected to contribute to specifying the performance requirements for immobilization of heterogeneous radioactive waste in domestic disposal facilities.

Domestic nuclear power plants can affect the environment if multiple devices are operated on one site and even a trace amount of pollutants that may affect the environment after power generation are simultaneously discharged. Therefore, not only radioactive substances but also ionic substances such as boron should be discharged as minimally as possible. We adopted pilot CDI and SD-ELIX sytem to separating and concenrating of boron containing nulcear power plant discharge water. The boron concentration of the initial inflow water tended to decrease over time. The water quality of concentrated water also reached its peak until the initial 60 minutes, but tended to decrease in line with the decrease in the inflow water concentration. The boron removal rate was in the range of 85 to 99% with respect to the initial boron concentration of 15 to 25 mg/L. On the other hand, performance degradation due to the use of electrochemical modules is also observed, and regeneration through low ion-containing water cleaning effective. We shortened processing time by considering the optimal flow rate conditions and conductivity conditions and converting electrochemical modules into series or parallel.

Nuclear power plants use ion exchange resins to purify liquid radioactive waste generated while operating nuclear power plants. In the case of PHWR, ion exchange resins are used in heavy water and dehydration systems, liquid waste treatment systems, and heavy water washing systems, and the used ion exchange resins are stored in waste resin storage tanks. The C-14 radioactivity concentration in the waste resin currently stored at the Wolseong Nuclear Power Plant is 4.6×106 Bq/g, exceeding the low-level limit, and if all is disposed of, it is 1.48×1015 Bq, exceeding the total limit of 3.04×1014 Bq of C-14 in the first stage disposal facility. Therefore, disposal is not possible at domestic low/medium-level disposal facilities. In addition, since the heavy water reactor waste resin mixture is stored at a ratio of about 20% activated carbon and zeolite mixture and about 80% waste resin, mixture extraction and separation technology and C-14 desorption and adsorption technology are required. Accordingly, research and development has been conducted domestically on methods to treat heavy water waste resin, but the waste resin mixture separation method is complex and inefficient, and there are limitations in applying it to the field due to the scale of the equipment being large compared to the field work space. Therefore, we would like to introduce a resin treatment technology that complements the problems of previous research. Previously, the waste resin mixture was extracted from the upper manhole and inspection hole of the storage tank, but in order to improve limitations such as worker safety, cost, and increased work time, the SRHS, which was planned at the time of nuclear power plant design, is utilized. In addition, by capturing high-purity 14CO2 in a liquid state in a high-pressure container, it ensures safety for long-term storage and is easy to handle when necessary, maximizing management efficiency. In addition, the modularization of the waste resin separation and withdrawal process from the storage tank, C-14 desorption and monitoring process, high-concentration 14CO2 capture and storage process, and 14CO2 adsorption process enables separation of each process, making it applicable to narrow work spaces. When this technology is used to treat waste resin mixtures in PHWR, it is expected to demonstrate its value as customized, high-efficiency equipment that can secure field applicability and safety and reflect the diverse needs of consumers according to changes in the working environment.

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.

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

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.

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.