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.

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.

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.

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.

We conducted safety assessments for the disposal of spent resin mixed waste after the removal of beta radionuclides (3H, 14C) in a landfill facility. The spent resin tank of Wolsong nuclear power plant is generated by 8:1:1 weight ratio of spent ion exchange resin, spent activated carbon and zeolite. Waste in the spent resin tank was classified as intermediate-level radioactive waste due to 14C. Other nuclides such as 60Co and 137Cs exhibit below the low-level radioactive waste criteria. The techniques for separating mixed waste and capturing 14C have been under development, with a particular focus on microwave-based methods to remove beta radionuclides (3H, 14C) from spent activated carbon and spent resin within the mixed waste. The spent resin and activated carbon within the waste mixture exhibits microwave reactivity, heated when exposed to microwaves. This technology serves as a means to remove beta isotopes within the spent resin, particularly by eliminating 14C, allowing it to meet the low-level radioactive waste criteria. Using this method, the waste mixture can meet disposal requirements through free water and 3H removal. These assessments considered the human intrusion scenarios and were carried out using the RESRAD-ONSITE code. The institutional management period after facility closure is set at 300 years, during which accidental exposures resulting from human intrusion into the disposal site are accounted for. The assessment of radiation exposure to intruders in a landfill facility included six human intrusion scenarios, such as the drilling scenario, road construction scenario, post-drilling scenario, and post-construction scenario. Among the six human intrusion scenarios considered, the most conservative assessment about annual radiation exposure was the post-drilling scenario. In this scenario, human intrusion occurs, followed by drilling and residence on the site after the institutional management period. We assumed that some of the vegetables and fruits grown in the area may originate from contaminated regions. Importantly, we confirmed that radiation doses resulting from post-institutional management period human intrusion scenarios remain below 0.1 mSv/y, thus complying with the annual dose limits for the public. This research underscores the importance of effectively managing and securing radioactive waste, with a specific focus on the safety of beta radionuclide-removed waste during long-term disposal, even in the face of potential human intrusion scenarios beyond the institutional management period.

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.

Copper, mainly used as a material for outer canister, generates various corrosion products under aerobic and anaerobic conditions in the operational and/or post-closure phases of the deep geological repository. These products could affect performance of engineering barrier system (EBS) through interaction with surrounding bentonite that makes up the buffer and backfill materials. Accordingly, in this study, we suggested research items to be conducted to minimize degradation of EBS due to copper corrosion products, based on the phenomenological review results for copper corrosion mechanisms and interaction between resultant product and bentonite in the deep geological disposal environment. During the post-closure phase, condition in the disposal facility changes form aerobic to anaerobic over time, and thereby, causes and products of copper corrosion vary. Under aerobic condition, copper corrosion is mainly induced by oxygen (O2) in the repository, chloride (Cl-) and carbonate (CO3 2-) ions from groundwater flowing into the facility, resulting in corrosion products such as cuprite (Cu2O), tenorite (CuO), atacamite (CuCl2·3Cu(OH)2) and malachite (Cu2CO3(OH)2). And, copper corrosion under anaerobic condition is primarily due to hydrogen sulfide (H2S) and sulfate (SO4 2-) in groundwater flowing into the facility, leading to formation of chalcocite (Cu2S) and covellite (CuS) as corrosion products. Depending on environment of the disposal facility, copper corrosion products are dissolved and ionized to Cu2+ in groundwater, and subsequently adsorbed on the nearby smectite. Then, it causes a cation exchange reaction with exchangeable cations in the interlayer of smectite. As a result of reviewing the previous experiments, it was confirmed that Cu2+-exchanged bentonite has a slightly reduced basal spacing and swelling capacity. From the results as above, there is a possibility that performance of EBS may be degraded due to copper corrosion products. To minimize its effect of degradation in the domestic facility, items to be further studied are as follows: (a) Method for reducing copper corrosion such as selection of appropriate material and structure for the canister, and (b) How to control dissolution of copper canister product into groundwater through predicting type and ionization process. The results of this study could be directly used to developing design concept of EBS for the domestic disposal facility and to establishing roadmap of future R&D programs.

Conducting a TSPA (Total System Performance Assessment) of the entire spent nuclear fuel disposal system, which includes thousands of disposal holes and their geological surroundings over many thousands of years, is a challenging task. Typically, the TSPA relies on significant efforts involving numerous parts and finite elements, making it computationally demanding. To streamline this process and enhance efficiency, our study introduces a surrogate model built upon the widely recognized U-network machine learning framework. This surrogate model serves as a bridge, correcting the results from a detailed numerical model with a large number of small-sized elements into a simplified one with fewer and large-sized elements. This approach will significantly cut down on computation time while preserving accuracy comparable to those achieved through the detailed numerical model.

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.

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.

APro, a process-based total system performance assessment (TSPA) tool for a geological disposal system, has a framework for simulating the radionuclide transport affected by thermal, hydraulic, mechanical or geochemical changes occurred in the disposal system. APro aims to be applied for the TSPA to long-term (> 100,000) evolution scenarios in real-world repository having more than 10,000 boreholes. In this large-scale TSPA, it is important not only to develop a high-performance numerical approach, but also to apply an efficient post-processing approach to massive spatiotemporal data. The post-processing refers to validating numerical analysis results, analyzing and evaluating target systems through data processing or visualization. Since APro uses COMSOL interface, the postprocessing function in COMSOL can be used. However, when the data size increases due to largescale numerical analysis, the time for the COMSOL post-processing increases, resulting in a problem that the analysis and evaluation are not performed effectively. In this case, it is possible to extract necessary data using the COMSOL exporting function and importing it into an external postprocessing program for the analysis and evaluation. In this study, the efficiency of external post-processing with extracted data from COMSOL was reviewed. And, we derived a proper data extraction approach (format and structure) that can increase efficiency of external post-processing.

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.

The presence of technological voids in deep geological repositories for high-level radioactive nuclear waste can have negative effects on the hydro-mechanical properties of the engineered barrier system when groundwater infiltrates from the surrounding rock. This study conducted hydration tests along with image acquisition and X-ray CT analysis on compacted Korean bentonite samples, which simulated technological voids filling to investigate the behavior of fracturing (piping erosion) and cracking deterioration. We utilized a dual syringe pump to inject water into a cell consisting of a bentonite block and technological voids at a consistent flow rate. The results showed that water inflow to fill technological voids led to partial hydration and self-sealing, followed by the formation of an erosional piping channel along the wetting front. After the piping channel generated, the cyclic filling-piping stage is characterized by the repetitive accumulation and drop of water pressure, accompanied by the opening and closing of piping channels. The stoppage of water inflow leads to the formation of macro- and micro cracks in bentonite due to moisture migration caused by high suction pressure. These cracks create preferential flow paths that promote longterm groundwater infiltration. The experimental test and analysis are currently ongoing. Further experiments will be conducted to investigate the effects of different dry density in bentonite, flow rate, and chemical composition of injected water.

Nuclear fuel assemblies are exposed to high temperature and high pressure environments underwater for long periods of time in a reactor, leading to deterioration of the assembly structure. These assembly consists of fuel rods, grids, a top nozzle, a bottom nozzle and guide tubes. In particular, the integrity of the guide tube made of Zircaloy-4 is a very important part in handling the assembly. In the Post Irradiation Examination Facility (PIEF), there are 14×14 Westinghouse STD assemblies that have lost their handleability due to the top nozzle being removed for damaged fuel rod test. To handle these assemblies, it is reasonable to use cut guide tubes whenever possible. Therefore, it is necessary to determine the irradiation embrittlement state of the guide tube before designing or manufacturing parts that can connect the top nozzle and the guide tubes. Therefore, in this paper, the location for installing the top nozzle-guide tube connection parts was selected in the height range of 3,460 to 3,713 mm, and guide tube specimens were made within that range. Offset strain was derived from the load-displacement curve obtained through compression testing to confirm whether the ductility of guide tubes was maintained. As a result, there was no significant difference in strength and ductility of the guide tube within the above length range. In addition, it was confirmed that the ductility was maintained enough to install the top nozzle-guide tube connection parts. Therefore, it is judged that there will be no problem even if the top nozzle-guide tube connection parts are installed in the guide tube to restore the handleability of the assemblies.