Among nuclear power plants in the Republic of Korea, Kori Unit 1 and Wolsong Unit 1 have been permanently shut down, and Kori Unit 1 is preparing to be decommissioned. According to the decommissioning plan (DP) of Kori Unit 1, a radioactive waste processing complex will be built on the Kori site to reduce radioactive waste generated during decommissioning actively, and various types of decommissioning waste are expected to be treated in the complex. It is judged that matters related to the safety assessment of the complex are not included in the DP since the equipment and treatment processes have not been determined. IAEA GSR Part 5 states that radioactive waste processing complex shall be operated according to national regulations and the conditions imposed by the regulatory body. However, it has been confirmed that separate regulatory requirements for the complex have not yet been established in Korea. It is expected that the Regulation on Technical Standards for Nuclear Facilities, etc. will be applied mutatis mutandis. Liquid and gaseous radioactive materials can be expected to be released into the sea or atmosphere during the operation of the complex. Accordingly, it should be proved that standards such as discharge limits of radioactive effluents are met. Although the assessment of radioactive effluent discharged from nuclear power plants to the environment is systematically conducted, it has been confirmed that the safety assessment framework for radioactive effluents discharged from the complex has not yet been established. Currently, the SAFRAN Tool is based on SADRWMS (Safety Assessment Driving Radioactive Waste Management Solutions), an IAEA safety assessment methodology for pre-disposal management, which uses Pathway Dose Factors (PDFs) derived from generic environmental models. Therefore, in order to conduct a more detailed safety assessment of the complex on a specific site, site characteristic data should be reflected. Although safety assessment using the SAFRAN Tool was conducted at the Thailand Institute of Nuclear Technology (TINT) facility, detailed data were not provided, and PDFs reflecting site characteristic data were not applied. Also, no other studies that considered many types of waste and provided detailed data on the safety assessment were not confirmed. Therefore, this study developed K-CRAFT (Kyung Hee – Comprehensive RAdioactive waste treatment Facility safety assessment Tool), this tool that can derive PDFs by reflecting site characteristic data based on the SADRWMS methodology and conducted preliminary safety assessment for the complex which will be built in Kori site by this tool.

Nuclear power plants in Korea stores approximately 3,800 drums of paraffin solidification products. Due to the lack of homogeneity, these solidification products are not allowed to be disposed of. There is therefore a need for the separation of paraffin from the solidification products. This work developed an equipment for a selective separation of paraffin from the solidification product using the vacuum evaporation and condensational recovery method in a closed system. The equipment mainly consists of a vacuum evaporator and a condensational deposition recovery chamber. Nonisothermal vacuum TGAs, kinetic analyses and kinetic predictions were conducted to set appropriate operation conditions. Its basic operability under the established conditions was first confirmed using pure paraffin solid. Simulated paraffin solidification product fixing dried boric acid waste including nonradioactive Co and Cs were then fabricated and tested for the capability of selective separation of paraffin from the simulated waste. Paraffin was selectively separated without entertainment of Co and Cs. It was confirmed that the developed equipment could separate and recover paraffin in the form of nonradioactive waste.

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

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.

The post-closure safety assessment of a repository is typically conducted over an extensive timescale from ten thousand to a million years. Considering that biosphere ecosystems may undergo significant changes over such lengthy periods, it is essential to incorporate the long-term evolution of the biosphere into the safety assessment. Climate change and landscape development are identified as critical drivers with the potential to impact the hydrogeological and hydrogeochemical characteristics of the biosphere. These changes can subsequently alter the migration patterns of radionuclides through the biosphere and influence human exposure doses. Therefore, this study formulates scenarios within the context of long-term biosphere evolution. We examine biosphere assessment processes employed in other countries and conduct a comparative study on scenario conditions. For example, biosphere assessment in Finland has identified sea-level changes and land-use alterations as significant factors in the long-term evolution of the biosphere. These factors are linked to Features, Events, and Processes (FEPs) associated with climate change and human activities. Sea-level changes are related to FEPs regarding climate change, land uplift, and shoreline displacement, while land-use changes are based on human activity-related FEPs (e.g., crop type, livestock and forest management, well construction, and demographics). Based on the literature review, this study has configured long-term evolution scenarios for the safety assessment of a deep geological repository for spent fuels.

Safety assessments for geological disposal systems extend over tens of thousands of years, taking into account the radiotoxicity decay period of spent nuclear fuel. During this extensive period, the biosphere experiences multiple glacial cycles, and fluctuations in seawater amounts, attributed to the formation and melting of glaciers, lead to global sea level changes known as eustacy. These sea level changes can directly influence the land-sea interface and groundwater flow dynamics, consequently affecting the pathways of radionuclide transport - an essential element of dose assessment. Therefore, this study aims to investigate how glacial cycles and sea level changes impact radionuclide transport within geological disposal systems, especially in the biosphere. To achieve this objective, we obtained climate evolution data including sea level changes for the Korean Peninsula over a 200,000-years, simulated by a General Circulation Model (GCM). These data were then employed to predict site and hydrology evolutions. The study site was conceptualized biosphere of Artificial Disposal System (ADioS), and we utilized the Soil and Water Assessment Tool (SWAT) to simulate hydrological evolution. These datasets, encompassing climate, site, and hydrology evolution, were collectively employed as inputs for the biosphere module of Adaptive Process-Based Total System Performance Assessment Framework (APro). Subsequently, the APro’s biosphere module calculated radionuclide transport in groundwater flow and its release into surface water bodies, considering the influences of glacial cycles and sea level changes. The results show that hydrologic changes due to sea level change are relatively minor, while the impact of sea level change on groundwater flow and discharge is significant. Additionally, we identified that among the water bodies within ADioS, including rivers, lakes, and oceans, the ocean exhibits the most substantial radionuclide outflow throughout the entire period. The spatiotemporal distributions of radionuclides computed within APro will be further processed into a grid format and used as input for the dose assessment module. Through this study, it was possible to determine the impact of long-term glacial cycles and sea level changes on radionuclide transport. Additionally, this module can serve as a valuable tool for providing the spatiotemporal variability of radionuclides required for enhanced dose assessments.

Advanced countries in the field of nuclear research and technology are currently examining the feasibility of deep geological disposal as the most appropriate method for the permanent management of high-level radioactive waste, with no intention of future retrieval. Deep geological disposal involves the placement of such waste deep underground within a stable geological formation, ensuring its permanent isolation from the human environment. To guarantee the enduring isolation and retardation of radionuclides with half-lives spanning tens of thousands to millions of years from the broader ecosystem, it is imperative to comprehend the long-term evolution of deep disposal systems, especially the role of natural barriers. These natural barriers, typically consisting of bedrock, encase the repository and undergo long-term evolutions due to tectonic movements and climate variations. For the effective disposal of high-level radioactive waste, a thorough assessment of the site’s long-term geological stability is essential. This necessitates a comprehensive understanding of its tectonic evolution and development characteristics, including susceptibility to seismic and magmatic events like earthquakes and intrusions. Furthermore, a detailed analysis of alterations in the hydrogeological and geochemical environment resulting from tectonic movements over extended time frames is required to assess the potential for the migration of radionuclides. In this paper, we have examined international evaluation methodologies employed to elucidate the predictive long-term evolution of natural barriers within disposal systems. We have extracted relevant methods from international case studies and applied a preliminary scenario illustrating the long-term evolution of the geological environment at the KURT (KAERI Underground Research Tunnel) site. Nevertheless, unlike international instances, the scarcity of quantitative data limits the depth of our interpretation. To present a dependable scenario in the future, it is imperative to develop predictive technologies aimed at comprehensively studying the geological evolution processes in the Korean peninsula, particularly within the context of radioactive waste disposal.

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

The Korea Atomic Energy Research Institute (KAERI) is currently developing a process-based performance assessment model known as APro. Distinguished from the previous system-level safety assessment model developed by KAERI, APro exhibits the capacity to encompass a threedimensional biosphere domain, evolving over the long term. In this study, we elucidate the methodology employed in developing the dose assessment module of APro and present the module’s functionalities. The procedural steps underlying radiation dose calculations within the APro framework can be succinctly outlined as follows: 1) Definition of a landscape model, utilizing information derived from a specified snapshot period provided by the APro biosphere transport module; 2) Generation of unit biotope objects spanning the landscape; 3) Evaluation of radionuclide transfer within the soil medium; 4) Calculation of activity concentration for flora and fauna groups; 5) Assessment of the distribution of effective dose among representative human groups; 6) Progressing through successive time steps. The APro dose calculation module exhibits notable capabilities that encompass: 1) Accounting for radionuclide decay and ingrowth; 2) Facilitating transfer through unsaturated porous media; 3) Considering sorption effects; 4) Addressing the inheritance of radioactivity between various landscape models; 5) Offering customizable ecosystem parameters; 6) Providing flexibility for user-defined exposure pathways. Leveraging these functionalities of the dose assessment module, APro is proficient in evaluating the distribution of radiological doses and associated risks for representative population groups, all while accounting for the dynamic, long-term evolution of the biosphere, including alterations in land cover.

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

With the importance of permanent disposal of high-level radioactive waste (HLW) generated in Korea, the deep geological disposal system based on the KBS-3 type is being developed. Since the deep geological repository must provide the long-term isolation of HLW from the surface environment and normal habitats for humans, plants, and animals, it is essential to assess the longterm performance of the disposal facility considering thermal-hydraulic-mechanical-chemical (TH- M-C) evolution. Decay heat dissipated from HLW contained in the canister causes an increase in temperature in the adjacent area. The requirement for the maximum temperature is established in consideration of the possibility of bentonite degradation. Therefore, when designing the repository, the temperature in the region of interest should be identified in detail through the thermal evolution assessment to ensure that the design requirement is satisfied. In the thermal evolution analysis, it is needed to evaluate the temperature distribution over the entire area of the disposal panel to consider the heat generated from both a single canister and adjacent canisters. Computational fluid dynamics (CFD) codes are widely used for detailed temperature analysis but are limited to simulating a wide range. Accordingly, in this study, we developed an analytical solution-based program for efficiently calculating the temperature distribution throughout the deposition panel, which is based on threedimensional heat conduction equations. The code developed can assess the temperature distribution of engineered and natural barrier systems. Principal parameters to be inputted are as follows: (a) geometry of the panel (e.g. width, length, height, spacing between canisters), (b) geometry of the canister (e.g. diameter, height), (c) thermal properties of bentonite and host-rock, (d) initial conditions (e.g. residual heat, temperature), and (e) time information (e.g. canister emplacement rate, time-interval, period). Through the calculation for the conceptual problem of a deposition panel capable of accommodating 900 (i.e. 30×30) canisters, it was confirmed that the program can adequately predict when and where the maximum temperature will occur. It is expected that the overall temperature distribution within the panel can be obtained by the evaluation of the entire region using this program reflecting the detailed design of the repository to be developed in the future. In addition, the thermal evolution analysis considering the influence of other canisters can be performed by applying the results as boundary conditions in the CFD analysis.

In 2012, POSIVA selected a bentonite-based (montmorillonite) block/pellet as the backfilling solution for the deposition tunnel in the application for a construction license for the deep geological repository of high-level radioactive waste in Finland. However, in the license application (i.e. SC-OLA) for the operation submitted to the Finnish Government in 2021, the design for backfilling was changed to a granular mixture consisting of bentonite (smectite) pellets crushed to various sizes, based on NAGRA’s buffer solution. In this study, as part of the preliminary design of the deep geological repository system in Korea, we reviewed history and its rationale for the design change of Finland’s deposition tunnel backfilling solution. After the construction license was granted by the Finnish Government in 2015, POSIVA conducted various lab- and full-scale in-situ tests to evaluate the producibility and performance of two design alternatives (i.e. block/pellet type and granular type) for backfilling. Principal demonstration tests and their results are summarized as follows: (a) Manufacturing of blocks using three types of materials (Friedland, IBeco RWC, and MX-80): Cracking and jointing under higher pressing loads were found. Despite adjusting the pressing process, similar phenomena were observed. (b) 1:6 scale experiment: Confirmation of density difference inhomogeneity due to the swelling of block/pellet backfill and void filling due to swelling behavior into the mass loss area of block/pellet. (c) FISST (Full-Scale In situ system Test): Identification of technical unfeasibility due to the inefficient (too manual) installation process of blocks/pellets and development of an efficient granular in-situ backfilling solution to resolve the disadvantage. (d) LUCOEX-FE (Large Underground Concept Experiments – Full-scale Emplacement) experiment: Confirmation of dense/homogeneous constructability and performance of granular backfilling solution. In conclusion, the simplified granular backfill system is more feasible compared to the block/ pellet system from the perspective of handling, production, installation, performance, and quality control. It is presumed that various experimental and engineering researches should be preceded reflecting specific disposal conditions even though these results are expected to be applied as key data and/or insights for selecting the backfilling solution in the domestic deep geological repository.

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.

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.

The WRK (Waste Repository Korea bentonite) compacted bentonite medium has been considered as the appropriate buffer material in the Korean SNF (Spent nuclear fuel) repository site. In this study, hydraulic properties of the WRK compacted bentonite core (4.5 cm in diameter and 1.0 cm in length) as the buffer material were investigated in laboratory experiments. The porosity and the entry pressure of the water saturated core at different confining pressure conditions were measured. The average velocity of water flow in the WRK compacted bentonite core was calculated from results of the breakthrough curves of the CsI aqueous solution and the hydraulic conductivity of the core was also calculated from the continuous flow core experiments. Because various gases could be generated by continuous SNF fission, container corrosion and biochemical reactions in the repository site, the gas migration property in the WRK compacted bentonite core was also investigated in experiments. The gas permeability and the average of gas (H2) in the core at different water saturation conditions were measured. Laboratory experiments with the WRK Compacted bentonite core were performed under conditions simulating the DGR environment (confining pressure: 1.5- 20.0 MPa, injection pressure: 1.0-5.0 MPa, water saturation: 0-100%). The WRK Compacted bentonite core was saturated at various confining pressure conditions and the porosity ranged from 27.5% to 43.75% (average: 36.75%). The calculated hydraulic conductivity (K) of the core using experimental results was 8.69×10-11 cm/s. The gas permeability of the core when the water saturation 0~58 % was ranged of 19.81~3.43×10-16 m2, representing that the gas migration in the buffer depends directly on the water saturation degree of the buffer medium. The average gas velocity in the core at 58% of water saturation was 9.8×10-6 m/s, suggesting that the gas could migrate fast through the buffer medium in the SNF repository site. Identification of the hydraulic property for the buffer medium, acquired through these experimental measurements is very rare and is considered to have high academic values. Experimental results from this study were used as input parameter values for the numerical modeling to simulate the long-term gas migration in the buffer zone and to evaluate the feasibility of the buffer material, controlling the radionuclide-gas migration in the SNF repository site.

According to the analysis of the Korean Radioactive Waste Society, saturation of nuclear power plant temporary storage is expected sequentially from 2031, and accordingly, the need for highlevel radioactive waste disposal facilities has emerged. In order to establish a repository for high-level radioactive waste, the performance and safety analysis of the repository must be conducted in compliance with regulatory requirements. For safety analysis, it needs a collection of arguments and evidence. and IAEA defined it as ‘Safety case’. The Systematic method, which derives scenarios by systematizing and combining possible phenomena around the repository, is widely used for developing Safety case. Systematic methods make use of the concept of Features, Events and Processes (FEP). FEP identifies features that affect repository performance, events that can affect a short period of time, and processes that can have an impact over a long period of time. Many countries, such as Finland, Sweden, Japan, United States, etc., are in process of licensing disposal facilities by using ‘Safety case’. And they then develop their own project-specified FEP lists and employ them for performing safety assessments. However, the systematic procedure for generating scenarios for safety evaluation is not clearly defined. According to the International Atomic Energy Agency (IAEA) Safety Standards Series (SSG- 23), the bottom-up method is an approach for conducting safety analysis using Features, Events, and Processes (FEPs). However, the process of how each FEP is utilized to establish a scenario for safety evaluation remains unclear. Additionally, there exists not only a bottom-up approach for generating scenarios using FEPs, but also a hybrid scenario development method that incorporates a top-down approach based on safety functions. Each country address scenario derivation in accordance with the adopted hybrid method. Nevertheless, a challenge arises in its application due to discrepancies between their approach and the hybrid approach specific which we are going on. Hence, this study introduces the FEP integration methodology for generating scenarios based on the hybrid scenario development method using the FEP list.

Raman characteristics of various minerals constituting natural rocks collected from uranium deposits in Okcheon metamorphic zone in Korea are presented. Micro-Raman spectra were measured using a confocal Raman microscope (Renishaw in Via Basis). The focal length of the spectrometer was 250 mm, and a 1800 lines/mm grating was installed. The outlet of the spectrometer was equipped with a CCD (1,024256 pixel) operating at -70°C. Three objective lenses were installed, and each magnification was 10, 50, and 100 times. The diameter of the laser beam passing through the objective lens and incident on the sample surface was approximately 2 m. The laser beam power at 532 nm was 1.6 mW on the sample surface. Raman signal scattered backward from the sample surface was transmitted to the spectrometer through the same objective lens. To accurately determine the Raman peak position of the sample, a Raman peak at 520.5 cm-1 measured on a silicon wafer was used as a reference position. Since quartz, calcite, and muscovite minerals are widely distributed throughout the rock, it is easy to observe with an optical microscope, so there is no difficulty in measuring the Raman spectrum. However, it is difficult to identify the uraninite scattered in micrometer sizes only with a Raman microscope. In this case, the location of uraninite was first confirmed using SEM-EDS, and then the sample was transferred to the Raman microscope to measure the Raman spectrum. In particular, a qualitative analysis of the oxidation and lattice conditions of natural uraninite was attempted by comparing the Raman properties of a micrometer-sized natural uraninite and a laboratory-synthesized UO2 pellet. Significantly different T2g/2LO Raman intensity ratio was observed in the two samples, which indicates that there are defects in the lattice structure of natural uraninite. In addition, no uranyl mineral phases were observed due to the deterioration of natural uraninite. This result suggests that the uranium deposit is maintained in a reduced state. Rutile is also scattered in micrometer-sizes, similar to uraninite. The Raman spectrum of rutile is similar in shape to that of uraninite, making them confused. The Raman spectral differences between these two minerals were compared in detail.

Deep borehole drilling is essential not only to select the host rock type for deep geological disposal of high-level radioactive waste (HLW), but also to identify the characteristics of the disposal site during the site selection process. In particular, since the disposal depth of HLW is considered to be over 300 m, deep borehole drilling must be performed. In deep borehole drilling, drilling design, excavation, and operation may vary depending on the rock type, drilling depth, and drilling purpose etc. This study introduced cases in which Korea was divided into four geotectonic structures and four representative rock types and conducted with a goal of 750 m drilling depth. Prior to this, a review of deep drilling cases conducted at domestic and abroad was presented. If sufficient time and cost are available, several drilling holes can be excavated for various purposes, but if not, one or two drilling holes should be used to achieve the objectives of various fields related to HLW disposal. The presence of bedding, strata or fault zones depending on the type of rock, etc. may affect drilling deviation or circulating water management. In addition, unlike drilling in general geotechnical investigation drilling, the use of polymers or grouting agents is limited to determine hydraulic and geochemical characteristics. This report introduces the experience considered during the design and drilling process of deep drilling in granite, gneiss, sedimentary rock, volcanic rock, etc., and is expected to be used as basic data when carrying out future HLW projects.

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.