본 연구를 통해 국내미기록속인 까마귀나비속의 푸른줄까마귀왕나비를 최초로 보고한다. 푸른줄까마귀왕나비의 암컷 형태와 생식기, 채집 지 등의 정보를 제공하였다. 또한 신속한 종 동정을 위한 DNA 바코드 염기서열을 정보를 제공하였다.

Electroanalytical study for the rotating cylinder electrode in molten LiCl-KCl eutectic salt (58– 42mol%) containing MgCl2 (0.1wt%) at 600°C is conducted. The researches of rotating cylinder electrode have been widely conducted for the century. The advantage of the electrode is that it can mitigate the unintended natural convection by providing a controlled diffusion boundary layer thickness. However, the experimental data for the high temperature molten salts is barely existed. The study adopts the electrochemical techniques such as cyclic voltammetry for the static cell and linear sweep voltammetry for the dynamic cell to calculate the diffusion coefficient. The peak current density and limiting current density are measured according to the scan rate. In order to evaluate the mass transfer under hydrodynamic flow condition, the revolution speeds of cylindrical electrode are varied from 10 rpm to 500 rpm which are corresponded to the Reynolds number of 4 and 185 respectively. The flow regime covers from the laminar to semi-turbulent regime (transient) as the critical Reynolds number Recrit is 200. The limiting current density shows a linear trend with the revolution speed and agrees well with the existing mass transfer correlations. For the extended flow regime, a new mass transfer correlation is suggested as the relation of non-dimensional numbers (Sh = aRebScc) based on the dimensionless analysis.

According to Article 4 and 5 of the Nuclear Safety and Security Commission (NSSC) Notice No. 2020-6, radioactive waste packages should be classified by radioactive levels, and finally permanently shipped to underground or surface disposal facilities. The level of the radioactive waste package is determined based on the concentrations of the radionuclides suggested in Article 8 of NSSC Notice No. 2021-26. Since most of the radionuclides in radioactive wastes are beta nuclides, chemical separation and quantification of the target nuclides are essential. Conventional methods to classify chemically non-volatile radionuclides such as Tc-99, Sr-90, Nb- 94, Fe-55 take a lot of time (about 5 days) and have low efficiency. An automated non-volatile nuclide analysis system based on the continuous chemical separation method of radionuclides has been developed to compensate for this disadvantages of the conventional method in this study. The features of the automated non-volatile nuclide separation system are as follows. First, the amount of secondary waste generated during the chemical separation process is very small. That is, by adopting an open-bed resin column method instead of a closed-bed resin column method, additional fittings and connector are unnecessary during the chemical separation. In addition, because the peristaltic pump is supplied for the sample and solution respectively, it is great effective to prevent cross-contamination between radioactive samples and the acid stock solution for analysis. Second, the factors that may affect results, such as solution amount, operating time and flow rate, are almost constant. By mechanically controlling the flow rate precisely, the operating time and additional factors required during the separation process can be adjusted and predicted in advance, and the uncertainty of the chemical separation process can be significantly reduced. Finally, it is highly usable not only in the continuous separation process but also in the individual separation process. It can be applied to the individual separation process because the user can set the individual sequence using the program. As a result of the performance evaluation of the automation system, recovery rates of about 80–90% and reproducibility within 5% were secured for all of the radionuclides. Furthermore, it was confirmed that the actual work time was reduced by more than 50% compared to the previous manual method. (It was confirmed that the operation time required during the separation process was reduced from 6 days to 3 days.) Based on these results, the automation system is expected to improve the safety of workers in radiation exposure, reduce human error, and improve data reliability.

Appropriateness of the minimum detectable activity in the analysis of gamma radionuclides is very important. This is reason determine the time factor among the conditions of the analysis when it is rationally determined has the advantage that radioactivity analysis can be performed accurately and quickly. In this study, 100 mL of an unknown sample was diluted in Marinelli Beaker 1L to obtain, review data on gamma radiation analysis results and minimum detectable activity for each measurement time. The measurement was used High Purity Germanium detector, target nuclides are Co-57, Co-58, Y-88 and Cs-137. Since the radioactivity analysis sample will be expected to be the waste subject to selfdisposal or less during the radioactive waste classification, the minimum detectable activity standard was set based on the detection of less than the permissible activity for self-disposal for each nuclide. The measurement methods were measured by classifying it into seven categories: 1000 seconds, 3600 seconds, 10000 seconds, 30000 seconds, 80000 seconds, 100000 seconds, and 150000 seconds. The radioactivity from this measurement are Co-57 2.89 Bq·g−1, Co-58 0.19 Bq·g−1, Y-88 0.20 Bq·g−1, Cs-137 0.15 Bq·g−1, the measurement results under all conditions were similar. On the other hand, the minimum detectable activity showed values above the allowable activity for self-disposal in not but Co-58 at 1000 and 3600 seconds. Only after taking the measurement time of 10000 seconds, the result was derived Co-57 0.0095 Bq·g−1, Co-58 0.0068 Bq·g−1, Y-88 0.0052 Bq·g−1, Cs-137 0.0062 Bq·g−1, which was confirmed to less than the allowable activity for self-disposal by nuclide. Reasonably determining the measurement time in gamma radionuclide analysis is a very important issue in terms of economy of time and accuracy of measurement. Although this study cannot be said to be able to determine a reasonable measurement time for all gamma radionuclide analysis, it is hoped that research on various samples will be made to contribute to the efficient measurement of gamma radioactivity.

The type of accidents associated with the operation of a melting facility for radioactive metal waste is assumed to only marginally differ from those associated with similar activities in the conventional metal casting industry or the current waste melting facility. However, the radiological consequences from a mishap or a technical failure differ widely. Three critical and at the same time possible accidents were identified: (1) activity release due to vapor explosion, (2) activity release due to ladle breakthrough, (3) consequences of failure in the hot-cell or furnace chamber not possible to remedy using remote equipment.

Laser scabbling experiments were conducted with the aim of developing concrete decontamination technology. Laser scabbling system contains a 6 kW fiber laser (IPG YLS-6000, λ=1,070 nm) and optical head, which are connected with process fiber (core dia.: 600 μm, length: 20 m). Optical head consists of two lenses (f = 160 mm and 100 mm) to collimate and focus laser beam. The focused laser beam is passed through the small diameter of nozzle (throat dia.: 3 mm) to prevent the laser-produced debris into head. And then, the focused beam is directed toward concrete block as continuously diverging. The diverged laser beam was incident on the high-strength concrete with 300 mm (length) × 300 mm (height) × 80 mm (width) to induce explosive spalling on the concrete surface. The optical head was moved by X-Y-Z manipulate coupled with a computerized numerical control while scabbling. We have observed not only active spalling on the concrete surface but energetic scattering of laserproduced debris when scabbling on high-strength concretes. It indicates the need for a device capable of collecting the laser-produced debris. We newly designed and manufactured dust collector coupled with cylindrical tube to prevent scattering of laser-produced debris into ambient environment. The collecting system was evaluated by estimating the collecting efficiency for laser-produced debris while scabbling. The collecting efficiency was calculated on the basis of the information on the mass loss of concrete block after laser scabbling and the mass of collected debris in a container. The collecting efficiency was found to be over 85%.

Source localization technique using acoustic emission (AE) has been widely used to track the accurate location of the damaged structure. The principle of localization is based on signal velocity and the time difference of arrival (TDOF) obtained from different signals for the specific source. However, signal velocity changes depending on the frequency domain of signals. In addition, the TDOF is dependent on the signal threshold which affects the prediction accuracy. In this study, a convolutional neural network (CNN)-based approach is used to overcome the existing problem. The concrete block corresponding to 1.3×1.3×1.3 m size is prepared according to the mixing ratio of Wolseong low-to-intermediate level radioactive waste disposal concrete materials. The source is excited using an impact hammer, and signals were acquired through eight AE sensors attached to the concrete block and a multi-channel AE measurement system. The different signals for a specific source are time-synchronized to obtain TDOF information and are transformed into a time-frequency domain using continuous wavelet transform (CWT) for consideration of various frequencies. The developed CNN model is compared with the conventional TDOF-based method using the testing dataset. The result suggests that the CNN-based method can contribute to the improvement of localization performance.

Low and intermediate radioactive wastes in South Korea have been disposed in Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC), Gyeongju. This repository structure is planned to be operated few hundred years while toxicity of the waste is sufficiently decayed. The structural integrity of the repository is required to protect the waste in safe. The integrity of the structure is commonly estimated using acoustic emission (AE) method. The integrity of the structure using AE is obtained by following process: 1) Estimation of maximum acoustic crack energy of the structure, 2) Acoustic signal measurement and filtering, and 3) Measurement of simultaneous acoustic cracking energy. The damage of the structure can be obtained from cumulative cracking energy from the structure divided by the predicted maximum cracking energy of the structure. Estimation of maximum cracking energy is gained by the specimens whose components are identical to the repository structure. The cracking energy of the different specimens are obtained during uniaxial compressive test and volume of the specimen is calculated. Then, the fractal coefficient for the structure is obtained and the maximum crack energy of the target structure can be calculated. The specimens whose diameters vary from 50 mm to 150 mm and heights are twice of the diameter are made with same recipe of WLDC silo concrete. The uniaxial compression test is conducted with loading rate of 0.1 mm·min−1. The fractal coefficient is obtained by least square method from the volume-cumulative energy relationship.

Low to intermediate radioactive waste disposal concrete structures are subjected to coupled hydromechanical conditions and the identification of structural damage is crucial to ensure safe long-term disposal. Different damage models for concrete and the surrounding rock can affect the damage characteristics of radioactive waste disposal structures. In this study, the effects of different rock damage models are applied to the hydro-mechanical-damage coupled structural analysis of the Wolseong Low and Intermediate Level Radioactive Waste Disposal Center silo. A two-dimensional model of the disposal silo was modeled using the finite element analysis software COMSOL and the Mazars’ damage model was applied to the silo concrete. The Mazars’ model parameters were obtained from uniaxial compression and tensile tests on cylindrical concrete specimens after 28 days of water curing and further 32 days of wet curing at 75°C). The COMSOL embedded Richards equation module was used to simulate hydraulic analysis. Structural loading due to waste disposal was applied at the bottom of the silo structure and the damage evolution characteristics were investigated. The non-linear mechanical rock behavior obtained from laboratory tests (Hoek-Brown criterion, resonant column test, Mazar’s damage model) and field tests (Goodman Jack) were input to assess the effects of different rock damage models. The results highlight the importance of structural damage consideration when assessing the long-term stability and safety of underground radioactive waste disposal structures under coupled hydro-mechanical conditions.

Expansive clays (for examples, bentonites) are favored as buffer and backfill materials because of their low hydraulic conductivity, high swelling potential, and good mechanical properties, and are installed in highly compacted blocks in repositories. Compacted expansive clays have a dual-structure system: macrostructural system which is a complex of clay aggregates with the inter-aggregate pores (macropores) which can be filled by either liquids or gases; microstructural system with the intraaggregate pores between or within clay particles (micropores) which is usually considered to be saturated by liquid. Understanding the dual-strucure system of expansive clays is essential for characterizing and modeling multiphysics (stress-strain, swelling pressure, etc.) in buffers and backfills. Existing multiphysics studies of expansive clays, as in non-expansive soils, were mostly conducted with a single structure approach based on the behavior of macropores, and there have been limitations in the comprehensive interpretation and modeling of experimental results. However, with the recent development of measurement techniques, a lot of available information on the pore structure of compacted expansive clays has been reported, and with the results, a dual-structure approach considering both microstructural and macrostructural systems has been increasingly applied to improve the modeling of multiphysics of expansive clays. This study reviewed the dual-structure system of compacted expansive clays, analyzed previous studies on its evolution according to hydromechanical loading (loading-unloading and wetting-drying paths), and based on these, intended to provide technical knowledge and information needed for multiphysics research of expansive clays-based buffer and backfill for the KRS repository.

High level nuclear waste (HLW) is surely disposed in repository in safe by being separated from human life zone. Deep geological disposal method is one of the most potent disposal method. Deep geological repository is exposed to high pressure and groundwater saturation due to its depth over 500 m. And it is also exposed to high temperature and radiation by spent fuels. Thus, HLW repository suffers extremely complex thermo-hydro-mechanical-radioactive condition. Long-term integrity of repository should be verified because the expected lifetime of the repository is over 10,000 years. However, the integrity of monitoring sensors are not reach the endurance lifetime of the repository with present technology. And the disposal condition, thermo-hydro-mechanical-radioactive, should shorten the estimated lifetime of the monitoring sensors. Therefore, it is necessary to improve the long-term integrity of the monitoring sensors. Although long-term tests are required to identify the prolonged durability of monitoring sensors, accelerated tests can help curtail test period. Accelerated tests is classified into accelerated stress test and accelerated degradation test and their methodology and theories are investigated. Their tests are design and proceed by following process: 1) identify failure modes, 2) select accelerated stress parameter, 3) Determine stress level, 4) Determine testing time and number of specimens, 5) Define measurement paremeter and failure criteria, 6) Suggest measurement method and measurement duration. Literature reviews were conducted to identify the influence of the disposal conditions such as thermo-hydro-mechnical-radioactive on integrity of material and monitoring sensors. The investigated data reported in this paper will be utilized to verify the improvement of integrity of monitoring sensors.

The backfill close the deep geological disposal system by filling the disposal tunnel and the connecting tunnel after the installation of buffer in the disposal hole. SKB and Posiva have established and designed the safety function of the backfill for the common goal of the deep geological disposal system. The safety function of backfill material has been set hydraulic conductivity of less than 10−10 m·s−1, a swelling pressure of 0.2 MPa, a compressive modulus of 10 MPa or a buffer density of 1,950 kg·m−3 or more, and freezing resistance. For the selection of the optimum backfill material, SKB and Posiva developed the concept of the backfill and evaluated the candidate that satisfies the requirements in four steps. In the first step, the performance and function that the backfill material should have were conceptualized. For the second step, laboratory tests and in-depth analysis of the candidate material properties were conducted. At this step, the focus has been on testing with the concept of the block method, using key candidate materials. In step 3, laboratory and large-scale experiments were performed to test engineering feasibility. In addition, design specifications for backfill materials were set based on site conditions, installation methods, and short- and long-term functions of materials. In Korea, it is only now in the step of selecting the concepts of the safety function. Therefore, it is necessary to benchmark the development process based on the previous studies of SKB and Posiva. In this study, candidate materials, experimental methods, and results were analyzed. As a result, the research steps and conditions for the selection of the optimum backfill material were reviewed. Using this study, the research steps of domestic backfill was suggested to develop within a short time for the Korean deep geological disposal system.

Deep geological disposal (DGD) of spent nuclear fuels (SNF) at 500 m–1 km depth has been the mainly researched as SNF disposal method, but with the recent drilling technology development, interest in deep borehole disposal (DBD) at 5 km depth is increasing. In DBD, up to 40SNF canisters are disposed of in a borehole with a diameter of about 50 cm, and SNF is disposed of at a depth of 2–5 km underground. DBD has the advantage of minimizing the disposal area and safely isolating highlevel waste from the ecosystem. Recently, due to an increasing necessity of developing an efficient alternative disposal system compared to DGD domestically, technological development for DBD has begun. In this paper, the research status of canister operation technology and plans for DBD demonstration tests, which subjects are being studied in the project of developing a safety-enhancing high-efficiency disposal system, are introduced. The canister operation technology for DBD can be divided into connection device development and operation technology. The developing connection device, emplacing and retrieving canisters in borehole, adopted the concept of a wedge thus making replacement equipment at the surface unnecessary. The new connection device has the advantage of being well applied with emplacement facilities only by simple mechanical operation. The technology of operating a connection device in DBD can be divided into drill pipe, coiled tubing, free-drop, and wireline. The drill pipe is a proven method in the oil industry, but requiring huge surface equipment. The coiled tubing method uses a flexible tube and shares disadvantages as the drill pipe. The free-drop is a convenient method of dropping canister into a borehole, but has a weakness in irretrievability in an accident. Finally, the wireline method can be operational on a small scale using hydraulic cranes, but the number of operated canisters at once is limited. The test facility through which the connection device is to be tested consists of dummy canister, borehole, lifting part, monitoring part, and connecting device. The canister weight is determined according to the emplacement operation unit. The lifting part will be composed following wireline consisting of a crane, a wire and a winding system. The monitoring part will consist of an external monitoring system for hoists and trolleys, and an internal monitoring system for the connection device’s location, pressure, and speed. In this project, a demonstration test will be conducted in a borehole with 1km depth, 10 cm diameter provided by KAERI to verify operation in the actual drilling environment after design improvement of the connecting device. If a problem is found through the demonstration test, the problem will be improved, and an improved connection device will be tested to an extended borehole with a 2 km depth, 40 cm diameter.

Deep geologic disposal of high-level nuclear wastes (HLW) requires intensive monitoring instrumentations to ensure long-term security. Acoustic emission (AE) method is considered as an effective method to monitor the mechanical degradation of natural rock and man-made concrete structures. The objectives of this study are (a) to identify the AE characteristics emitted from concretes as concrete materials under different types of loading, (b) to suggest AE parametric criteria to determine loading types and estimate the failure stage, and finally (c) to examine the feasibility of using AE method for real-time monitoring of geologic disposal system of HLW. This study performs a series of the mechanical experiments on concrete samples simultaneously with AE monitoring, including the uniaxial compression test (UCT), Brazilian tensile test (BTT) and punch through shear test (PTST). These mechanical tests are chosen to explore the effect of loading types on the resulting AE characteristics. This study selects important AE parameters which includes the AE count, average frequency (AF) and RA value in the time domain, and the peak frequency (PF) and centroid frequency in the frequency domain. The result reveals that the cumulative AE counts, the maximum RA value and the moving average PF show their potentials as indicators to damage progress for a certain loading type. The observed trends in the cumulative AE counts and the maximum RA value show three unique stages with an increase in applied stress: the steady state stage (or crack initiation stage; < 70% of yield stress), the transition stage (or damage progression stage; 70–90% of yield stress) and the rising stage (or failure stage; > 90% of yield stress). In addition, the moving average PF of PTST in the early damage stage appears to be particularly lower than that of UCT and BTT. The loading in BTT renders distinctive responses in the slope of the maximum RA–cumulative AE count (or tan ). The slope value shows less than 0.25 when the stress is close to 30% of BTT, 60% of UCT and 75% of PTST and mostly after 90% of yield stress, the slope mostly decreases than 0.25 in all tests. This study advances our understanding on AE responses of concrete materials with well-controlled laboratoryscale experimental AE data, and provides insights into further development of AE-base real-time diagnostic monitoring of structures made of rocks and concretes.

A methodology is under development to restore and predict the long-term evolution of the natural barrier comprise the site of radioactive waste disposal for surface geological outcrop, tunnel face and drill core. 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. In this study, we investigated proper age dating methods to construct lithological model of the disposal site with regard to the long-term safety. Also, preliminary age dating locations were selected using the lithological distribution results by depth through geochemical and micro-structural analysis for the deep drill cores excavated around KURT. In the study area, the dikes presumed the Cretaceous were intruded by Jurassic granites. As for the granotoids, U-Pb age dating for zircon, which is resistant to deformation or metamorphism and has loss, is often used. In the case of the dikes, K-Ar and 40Ar/39Ar age dating for the argon captured in the rocks after magmatism is often used. Through U-Pb zircon ages of KURT site granotoids, we expect to solve the clustering problem (granite and granodiorite), which is different from precious chemical analysis (XRF) results and TAS-diagrams. 40Ar/39Ar age dating to be used for the dikes is suitable for the perspective of lithological model of the disposal site. Because, it can compensate for accuracy problems such as sample heterogeneity in K-Ar age dating and is used for volcanic rocks. In the further study, we plan to determine the appropriate sampling locations by the selected age dating methods from the perspective of disposal in this study.

In South Korea, the master plan for high-level radioactive waste management, announced in 2016, suggested the construction and operation of intermediate storage facilities on a permanent disposal site and specified the adoption of dry storage in consideration of the ease of operation and expansion. As of 2021, the government is again reviewing its overarching policy on the back-end fuel cycles, including intermediate storage and permanent disposal. In the case of dry storage facilities, safety evaluation is being conducted using a combination of deterministic and probabilistic approaches, similar to that of nuclear power plants. The two methods are complementary, of which Probabilistic Safety Assessment (PSA) has the advantage of being able to identify key scenarios affecting safety, but its use in storage facilities has not been highlighted so far. However, depending on the spent fuel management phases such as loading, transportation, and storage, it may be not enough to capture effective and efficient safety evaluation only deterministically, and probabilistic methods may contribute to the evaluation of long-term operation or external events such as an earthquake. There have already been cases where PSA has been performed on a part of the nuclear fuel cycle through previous studies. This paper created the safety assessment model based on open sources such as the released EPRI reports, by targeting arbitrary intermediate storage facilities. The model considered the scenarios for loading, transportation, and storage, with human error respectively. It will be able to be modified and improved to fit domestic and specific intermediate storage facilities in the future.