The primary objective of this study is to evaluate a systematic design’s effectivity in remediating actual uranium-contaminated soil. The emphasis was placed on practical and engineering aspects, particularly in assessing the capabilities of a zero liquid discharge system in treating wastewater derived from soil washing. The research method involved a purification procedure for both the uranium-contaminated soil and its accompanying wastewater. Notably, the experimental outcomes demonstrated successful uranium separation from the contaminated soil. The treated soil could be self-disposed of, as its uranium concentration fell below 1.0 Bq·g−1, a level endorsed by the International Atomic Energy Agency for radionuclide clearance. The zero liquid discharge system’s significance lay in its distillation process, which not only facilitated the reuse of water from the separated filtrate but also allowed for the self-disposal of high-purity Na2SO4 within the residues of the distilled filtrate. Through a comparative economic analysis involving direct disposal and the application of a remediation process for uranium-contaminated soil, the comprehensive zero liquid discharge system emerged as a practical and viable choice. The successful demonstration of the design and practicality of the proposed zero liquid discharge system for treating wastewater originating from real uranium-contaminated soil is poised to have a lasting impact.

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 safety of deep geological disposal systems has to be ensured to guarantee the isolation of radionuclides from human and related environments for over a million years. Over such a long timeframe, disposal systems can be influenced by climate change, leading to significant long-term impacts on the hydrogeological condition, including changes in temperature, precipitation and sea levels. These changes can affect groundwater flow, alter geochemical conditions, and directly/ indirectly impact the stability of the repository. Hence, it is essential to conduct a safety assessment that considers the long-term evolution induced by climate change. In this context, the Korea Atomic Energy Research Institute (KAERI) is developing the Adaptive Process-based total system performance assessment framework for a geological disposal system (APro). Currently, numerical modules for APro are under development to account for the longterm evolution that can influence groundwater flow and radionuclide transport in the far-field of the disposal system. This study focuses on the development of two numerical modules designed to model permafrost formation and buoyance force due to relative density changes. Permafrost is defined as a ground in which temperature remains below zero-isotherm (0°C) continuously for more than two consecutive years. In regions where permafrost forms, the relative permeability of porous media is significantly reduced. The changes in permeability due to permafrost formation are modelled by calculating the unfrozen fluid content within a porous medium. Meanwhile, buoyancy force can occur when there is a difference in density at the boundary of two distinct water groups, such as seawater (salt water) and freshwater. Sea level change associated with climate change can alter the boundary between seawater and freshwater, resulting in changes in groundwater flow. The buoyancy force due to relative density is modelled by adjusting concentration boundary conditions. Using the developed numerical modules, we evaluated the long-term evolution’s effects by analyzing radionuclide transport in the far-field of the disposal system. Incorporating permafrost and buoyancy force modelling into the APro framework will contribute valuable insights into the complex interactions between geological and climatic factors, enhancing our ability to ensure the secure isolation of radionuclides for extended periods.

Over the years, in the field of safety assessment of geological disposal system, system-level models have been widely employed, primarily due to considerations of computational efficiency and convenience. However, system-level models have their limitations when it comes to phenomenologically simulating the complex processes occurring within disposal systems, particularly when attempting to account for the coupled processes in the near-field. Therefore, this study investigates a machine learning-based methodology for incorporating phenomenological insights into system-level safety assessment models without compromising computational efficiency. The machine learning application targeted the calculation of waste degradation rates and the estimation of radionuclide flux around the deposition holes. To develop machine learning models for both degradation rates and radionuclide flux, key influencing factors or input parameters need to be identified. Subsequently, process models capable of computing degradation rates and radionuclide flux will be established. To facilitate the generation of machine learning data encompassing a wide range of input parameter combinations, Latin-hypercube sampling will be applied. Based on the predefined scenarios and input parameters, the machine learning models will generate time-series data for the degradation rates and radionuclide flux. The time-series data can subsequently be applied to the system-level safety assessment model as a time table format. The methodology presented in this study is expected to contribute to the enhancement of system-level safety assessment models when applied.

The operation time of a disposal repository is generally more than one hundred years except for the institutional control phase. The structural integrity of a repository can be regarded as one of the most important research issues from the perspective of a long-term performance assessment, which is closely related to the public acceptance with regard to the nuclear safety. The objective of this study is to suggest the methodology for quantitative evaluation of structural integrity in a nuclear waste repository based on the adaptive artificial intelligence (AI), fractal theory, and acoustic emission (AE) monitoring. Here, adaptive AI means that the advanced AI model trained additionally based on the expert’s decision, engineering & field scale tests, numerical studies etc. in addition to the lab. test. In the process of a methodology development, AE source location, wave attenuation, the maximum AE energy and crack type classification were subsequently studied from the various lab. tests and Mazars damage model. The developed methodology for structural integrity was also applied to engineering scale concrete block (1.3 m × 1.3 m × 1.3 m) by artificial crack generation using a plate jacking method (up to 30 MPa) in KURT (KAERI Underground Research Tunnel). The concrete recipe used in engineering scale test was same as that of Gyeongju low & intermediate level waste repository. From this study, the reliability for AE crack source location, crack type classification, and damage assessment increased and all the processes for the technology development were verified from the Korea Testing Laboratory (KTL) in 2022.

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.

The effectiveness of a crystalline natural barrier in providing sealing capabilities is based on the behavior of numerous fractures and their intersections within the rock mass. It is important to evaluate the evolving characteristics of fractured rock, as the hydro-mechanical coupled processes occurring through these fractures play a dominant role. KAERI is actively developing a true tri-axial compression test system and concurrently conducting hydro-mechanical experiments using replicated fractured rock samples. This research is focused on a comprehensive examination of coupled processes within fractures, with a particular emphasis on the development of true tri-axial testing equipment. The designed test system has the capability to account for three-dimensional stress conditions, including vertical and both maximum and minimum horizontal principal stresses, realizing the disposal conditions at specific underground depths. Notably, the KAERI-designed test system employs the mixed true tri-axial concept, also known as the Mogi-type, which allows for fluid flow into fractures under tri-axial compression conditions. This system utilizes a hydraulic chamber to maintain constant stress in one direction through the application of oil pressure, while the other two directional stresses are applied using rigid platens with varying magnitudes. Once these mechanical stress conditions are established, control over fluid flow is achieved through the rigid platens in contact with the specimen section. This pioneering approach effectively replicates in-situ mechanical conditions while concurrently observing the internal fluid flow patterns within fractures, thereby enhancing our capacity to study these coupled phenomena. As future research, numerical modeling efforts will be proceeding with experimental data-driven approaches to simulate the coupled behavior within the fractures. In these numerical studies, two distinct fracture geometry domains will be generated, one employing simplified rough-walled fractures and the other utilizing mismatched rough-walled fractures. These investigations mark the preliminary steps in the process of selecting and validating an appropriate numerical model for understanding the hydro-mechanical evolution within fractures.

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 high-level radioactive waste repository must ensure its performance for a long period of time enough to sufficiently reduce the potential risk of the waste, and for this purpose, multibarrier systems consisting of engineered and natural barrier systems are applied. If waste nuclides leak, the dominating mechanisms facilitating their movement toward human habitats include advection, dispersion and diffusion along groundwater flows. Therefore, it is of great importance to accurately assess the hydrogeological and geochemical characteristics of the host rock because it acts as a flow medium. Normally, borehole investigations were used to evaluate the characteristics and the use of multi-packer system is more efficient and economical compared to standpipes, as it divides a single borehole into multiple sections by installing multiple packers. For effective analyses and groundwater sampling, the entire system is designed by preselecting sections where groundwater flow is clearly remarkable. The selection is based on the analyses of various borehole and rock core logging data. Generally, sections with a high frequency of joints and evident water flow are chosen. Analyzing the logging data, which can be considered continuous, gives several local points where the results exhibit significant local changes. These clear deviations can be considered outliers within the data set, and machine learning algorithms have been frequently applied to classify them. The algorithms applied in this study include DBSCAN (density based spatial clustering of application with noise), OCSVM (one class support vector method), KNN (K nearest neighbor), and isolation forest, of which are widely used in many applications. This paper aims to evaluate the applicability of the aforementioned four algorithms to the design of multi-packer system. The data used for this evaluation were obtained from DB-2 borehole logging data, which is a deep borehole locates near KURT.

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

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.

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

Buffer materials are one of the engineering barrier components in high-level radioactive waste disposal facilities. Compacted bentonite has been known as the most suitable buffer material so far, and research is being conducted worldwide to determine the characteristics of suitable bentonite blocks in each country. Therefore, this study aims to compare and analyze various properties of different buffer material candidates, including thermal, hydraulic, and mechanical properties. Buffer material candidates for Korea disposal system, Kyungju Bentonite (KJ-I, KJ-II), and Bentonil- WRK were compared. The properties were compared and analyzed based on experimental and literature data. The data obtained from this report can be used to understand the behavior of buffer materials and assess whether they meet various criteria, such as temperature and saturation, and ultimately serve as an essential input variable database for safety evaluations of disposal systems.

Even though a huge amount of spent nuclear fuels are accumulated at each nuclear power plant site in Korea, our government has not yet started to select a final disposal site, which might require more than several km2 surface area. According to the second national plan for the management of high-level radioactive waste, the reference geological disposal concept followed the Finnish concept based on KBS-3 type. However, the second national plan also mentioned that it was necessary to develop the technical alternatives. Considering the limited area of the Korean peninsula, the authors had developed an alternative disposal concepts for spent nuclear fuels in order to enhance the disposal density since 2021. Among ten disposal concepts shown in the literature published in 2000’s, we narrowed them to four concepts by international experiences and expert judgements. Assuming 10,000 t of CANDU spent nuclear fuels (SNF), we designed the engineered barriers for each alternative disposal concept. That is, using a KURT geological conditions, the engineered barrier systems (EBS) for the following four alternative concepts were proposed: ① mined deep borehole matrix, ② sub-seabed disposal, ③ deep borehole disposal, and ④ multi-level dispoal. The quantitative data of each design such as foot prints, safety factors, economical factors are produced from the conceptual designs of the engineered barriers. Five evaluation criteria (public acceptance, safety, cost, technology readiness level, environmental friendliness) were chosen for the comparison of alternatives, and supporting indicators that can be evaluated quantitatively were derived. The AHP with domestic experts was applied to the comparison of alternatives. The twolevel disposal was proposed as the most appropriate alternative for the enhancement of disposal efficiency by the experts. If perspectives changes, the other alternatives would be preferred. Three kinds of the two-level disposal of CANDU SNF were compared. It was decided to dispose of all the CANDU spent nuclear fuels into the disposal holes in the lower-level disposal tunnels because total footprint of the disposal system for CANDU SNF was much smaller than that for PWR SNF. Currently, we reviewed the performance criteria related to the disposal canister and the buffer and designed the EBS for CANDU SNF. With the design, safety assessment and cost estimates for the alternative disposal system will be carried out next year.

Nuclear power is responsible for a large portion of electricity generation worldwide, and various studies are underway, including the design of permanent deep geological disposal facilities to safely isolate spent nuclear fuel generated as a result. However, through the gradual development of drilling technology, various disposal option concepts are being studied in addition to deep geological disposal, which is considered the safest in the world. So other efforts are also being made to reduce the disposal area and achieve economic feasibility, which requires procedures to appropriately match the waste forms generated from separation process of spent nuclear fuel with disposal option systems according to their characteristics. And safety issue of individual disposal options is performed through comparison of nuclide transport. This study briefly introduces the pre-disposal nuclide management process and waste forms, and also introduces the characteristics of potential disposal options other than deep geological disposal. And environmental conditions and possible pathways for nuclide migration are reviewed to establish transport scenarios for each disposal option. As such, under this comprehensive understanding, this study finally seeks to explore various management methods for high-level radioactive waste to reduce the environmental burden.

It is very important that the confinement of a spent fuel storage systems is maintained because if the confinement is damaged, the gaseous radioactive material inside the storage cask can leak out and have a radiological impact on the surrounding public. For this reason, leakage rate tests using helium are required for certificate of compliance (CoC) and fabrication inspections of spent fuel storage cask. For transport cask, the allowable leakage rate can be calculated according to the standardized scenario presented by the IAEA. However, for storage cask, the allowable leakage rate is determined by the canister, facility, and site specific information, so it is difficult to establish a standardized leakage rate criterion. Therefore, this study aims to establish a system that can derive system-specific leakage test criteria that can be used for leakage test of actual storage systems. First, the variables that can affect the allowable leakage rate for normal and accident conditions were derived. Unlike transportation systems, for storage systems, the dose from the shielding analysis and the dose from the confinement analysis are summed up to determine whether the dose standard is satisfied, and even the dose from the existing nuclear facilities is summed up during normal operation condition. For this reason, the target dose is used as an input variable when calculating the allowable leakage rate for the storage system. In addition, the main variables are the distance from the boundary of the exclusive area, the number of cask, the inventory of nuclide material in the cask, the free volume, and the internal and external pressure. Utilizing domestic and US NRC guidelines, we derived basic recommended values for the selected variables. The GASPARII computer code that can evaluate the dose to the public under normal operating conditions was utilized. Using the above variables, the allowable leakage rate is calculated and converted to the allowable criteria for helium leakage rate test. The developed system was used to calculate the allowable leakage rate for normal and accident conditions for a hypothetical storage system. The leakage rate criteria calculation system developed in this study can be useful for CoC and fabrication inspections of storage systems in the future, and a GUI-based program will be built for user convenience.

This paper describes the development and operation of an autonomous robotic system designed for pyroprocess automation. The unique challenges of pyroprocess automation, such as the need for a highly dry atmosphere to handle materials like chloride, are addressed through this system. For the experiments, a specialized dehumidifier and dry mock-up facility were designed to produce dry air condition. Performances of dry air conditioning for the various simulated situations were evaluated, including assessing worker access within a mock-up to determine the system’s feasibility. To enable automation, containers used for processing materials were modified to fit the gripper system of the gantry robot. The loading and unloading of materials in each equipment were automated to connect them with the robotic system. This gantry robot primarily utilized macro motions to approach waypoints containing process materials, reducing the need for precise approach motions. Its tapered jaw design allowed it to grip target objects even with imperfect positioning. The robot’s motions were programmed initially using a robot simulator for positioning and motion planning, and real-world accuracy was tested in a dry mock-up facility using the OPC platform. Finally, the paper discusses the potential application of XR (eXtended Reality) technology in this context, which could enhance the robot’s operation and provide valuable insights into the automation process. Further analysis of XR technology’s feasibility and benefits for this specific pyroprocess automation system are presented.

Since the Fukushima nuclear accident in 2011, the development of accident tolerant fuel (ATF) has been actively pursued as an alternative to improve the safety of nuclear power plants. In addition, nuclear power plants containing ATF have recently been included as green energy in the 2022 EU taxonomy bill, receiving a lot of attention. Many countries are considering increasing 235U enrichment from 5 to 10 235U % for higher burnup and long cycle operation with ATF improving safety. To utilize ATF, the applicability of fuel storage systems such as new fuel storage vault, Region 1, and Region 2 must be determined. The purpose of this paper is to confirm the applicability of applying ATF, which is being developed in Korea, to the nuclear fuel storage system of Korean nuclear power plants. The nuclear power plant model used in the analysis is APR-1400, a representative Korean nuclear power plant model, and ATF model used in the analysis is Mo microcell UO2 pellet with CrAl coating, which is being developed in Korea. MCNP 6.2 has been used for multiplication factor calculations, and the TRITON/NEWT and ORIGEN-S modules of the SCALE code have been used for depletion calculations. From the analysis results, solutions and additional analysis would be necessary to satisfy criticality regulatory requirements to utilize ATF with increased enrichment.