Colloid Formation and Migration (CFM) project is being carried out within the Grimsel Test Site (GTS) Phase Ⅵ. Since 2008, the Korea Atomic Energy Research Institute (KAERI) has joined CFM to investigate the behavior of colloid-facilitated radionuclide transport in a generic Underground Research Laboratory (URL). The CFM project includes a long-term in-situ test (LIT) and an in-rock bentonite erosion test (i-BET) to assess the in-situ colloid-facilitated radionuclide transport through the bentonite erosion in the natural flow field. In the LIT experiment, radionuclide-containing compacted bentonite was equipped with a triple-packer system and then positioned at the borehole in the shear zone. It was observed that colloid transport was limited owing to the low swelling pressure and low hydraulic conductivity. Therefore, a postmortem analysis is being conducted to estimate the partial migration and diffusion of radionuclides. The i-BET experiment, that focuses more on bentonite erosion, was newly designed to assess colloid formation in another flow field. The i-BET experiment started with the placement of compacted bentonite rings in the double-packer system, and the hydraulic parameters and bentonite erosion have been monitored since December 2018.

In 1951, Colin Russell Austin and Min Chueh Chang identified “capacitation”, a special process involving ejaculated spermatozoa in the female reproductive tract. Capacitation is a phenomenon that occurs in vivo , but almost all knowledge of capacitation has been obtained from in vitro studies. Therefore, numerous trials have been performed to establish in vitro capacitation methods for various studies on reproduction. Although a series of studies have been conducted to develop an optimal protocol for inducing capacitation, most have focused on identifying the appropriate chemical compounds to induce the capacitation of boar spermatozoa in vitro. Therefore, the purpose of this study was to identify the optimal incubation time for inducing capacitation in vitro. Duroc semen was incubated for various periods (60, 90, and 120 min) to induce capacitation. Sperm function (sperm motility, motion kinematic parameters, and capacitation status) was evaluated. The results showed that total sperm motility, rapid sperm motility, progressive sperm motility, curvilinear velocity, and average path velocity significantly decreased in a time-dependent manner. However, the capacitation status did not show any significant changes. Taken together, these results indicate that an incubation time of more than 60 min suppresses sperm motility and motion kinematic parameters. Therefore, we suggest that 60 min may be the best incubation time to induce capacitation without negative effects on sperm motility and motion kinematics in boar spermatozoa in vitro.

국립원예특작과학원에서 2016년 거베라 신품종 ‘샤이핑크 (Shy Pink)’를 육성하였다. ‘Shy Pink’는 2008년 10월부터 2008년 11월까지 녹색 화심의 분홍색 반겹꽃 ‘프리존’과 갈 색 화심의 살구색 반겹꽃 ‘아틸리아’를 교배하여 획득한 실생 72개체 중에서 2009년부터 2016년까지 생육 및 개화특성검 정을 거쳐 선발한 절화용 대륜 거베라 품종이다. ‘샤이핑크 (Shy Pink)’는 화형이 안정적인 갈색 화심의 연분홍색(RHS R49A) 반겹꽃 대륜계 품종이다. 꽃은 평균 직경이 12.4cm이 고, 내부설상화와 화반의 직경은 각각 7.2cm와 2.3cm이다. 외부설상화의 길이와 폭은 각각 6.0cm와 1.2cm로 꽃잎이 좁은 도란형이다. 꽃대 직경은 상부 6.1mm, 중부 7.0mm 정 도로 강건하며, 꽃대의 길이는 48.9cm이다. 절화수명은 12.1 일로 대조구인 ‘하모니(Harmony)’ 12.0일로 유사하였지만 절 화수명이 길며, 연간 채화량은 52.8본/주로 다수성 품종이다.

This paper explores the L2 motivation of 11 students about to start studying English as junior college English majors. The participants took part in a ten-day L2 motivational program designed to help them with L2 skills and life on campus. To understand how they participated in the program, data gathered from online learning logs were coded using qualitative thematic analysis. Consequently, the participants shifted to embody ‘none to something’ qualities in terms of L2 selves and effort during the program. That is, they established self-images to believe in when using and learning English, which eventually developed into effort-driven attitudes and behavior. The students also set goals and plans according to their L2 selves created or revealed during the program. Although the program was comparatively short with a small number of participants entering a junior college, the findings indicate that an L2 motivational program for new adult students could be conducive to promoting and sustaining their effort, L2 selves, and motivation.

The study aims to investigate Korean English learners' usage of pragmatic discourse markers in their impromptu speech. Korean English Learners' Spoken Corpus, abbreviated as KELSC, is the main data source for analysis(Jung, 2021). To analyze the features of pragmatic discourse markers in KELSC, three focal pragmatic discourse markers(PDM) okay, like, and so are selected and analyzed by using WordSmith 7.0(Scott, M. 2016) and Lancsbox 6.0(Brezina et al., 2021). The findings indicate that three focal PDMs okay, like, so are not appearing more frequently in the higher levels of learners, unlike previous studies. Instead, the intermediate levels of B1 and C1 used two focal PDMs, okay and like the most, respectively. However, the lowest A1 learners rarely used the focal PDMs in their speech. Two PDMs okay and so are used as a co-occurring pattern in level C while so and they have a shared collocate in level A.

Background: The lateral step down (LSD) is a form of stair negotiation used by the elderly because it requires less movement of the lower extremity. Although it is necessary to study the amount of pelvic drop and the strength of a hip abductor during LSD for intervention, limited studies have investigated the relationship between the amount of pelvic drop and strength of a hip abductor during LSD in elderly people. Objects: This study aimed to determine the relationship between the amount of pelvic drop on an unsupported leg and the strength of the hip abductor during LSD in the elderly. Methods: Thirty elderly people (male: 17, female: 13) were recruited. Subjects performed the LSD task, and the evaluator measured and the amount of pelvic drop on an unsupported side. Also, the isometric strength of the hip abductor was measured in a supine position. Results: We found significant relationships between the strength of the hip abductor and the amount of pelvic drop (r = –0.386). The average hip abductor strength normalized by body weight was 1.06 N/kg (max: 1.99, min: 0.52) and the average contralateral pelvic drop (CPD) angle was 4.16° (max: 15.3, min: 0). Conclusion: Our results indicated that the strength of the hip abductor had a moderate correlation with the CPD during a LSD in the elderly. Hip abductor weakness could translate into altered movement of the pelvis.

Investigations and monitoring of environmental radiation are important for preventing expected accidents or for early detection of unexpected accidents, in nuclear facilities and the surrounding. In the event of an environmental radiation accident, it should be possible to identify and analyze the radiation-contaminated area. Therefore, a rapid radiation monitoring system is required for immediate response and necessary measures. In this study, the distribution of radiation mapping is performed on a contaminated area using 2-dimensional or 3-dimensional contour mapping techniques. The entire surrounding area can be understood at a glance by displaying the radiation contour line on the map of the measured area.

When decommissioning a nuclear power plant, the structure must be made to a disposable size. In general, the cutting process is essential when dismantling a nuclear power plant. Mainly, thermal cutting method is used to cutting metal structures. The aerosols generated during thermal cutting have a size distribution of less than 1 μm. The contaminated structures are able to generate radioactive aerosols in the decommissioning. Radioactive aerosols of 1 μm or less are deposited in the respiratory tract by workers’ breathing, causing the possibility of internal exposure. Therefore, workers must be protected from the risk of exposure to radioactive aerosols. Prior knowledge of aerosols generated during metal cutting is important to ensure worker safety. In this study, the physical and chemical properties of the aerosol were evaluated by measuring the number and mass concentrations of aerosols generated when cutting SUS304 and SA508 using the laser cutting method. High-resolution aerosol measuring equipment (HR-ELPI+, DEKATI) was used to measure the concentration of aerosols. The HR-ELPI+ is an impactor-type aerosol measuring equipment that measures the aerosol number concentration distribution in the aerodynamic diameter range of 6 nm to 10 um in real-time. And analyze the mass concentration of the aerosol according to the diameter range through the impactor. ICP-MS was used for elemental mass concentration analysis in the aerosol. Analytical elements were Fe, Cr, Ni and Mn. For the evaluation of physical and chemical properties, the MMAD of each element and CMAD were calculated in the aerosol distribution. Under the same cutting conditions, it was confirmed that the number concentration of aerosols generated from both materials had a uni-modal distribution with a peak around 0.1 um. CMAD was calculated to be 0.072 um for both SUS304 and SA508. The trend of the CMAD calculation results is the same even when the cutting conditions are changed. In the case of MMAD, it was confirmed that SUS304 had an MMAD of around 0.1 μm in size for only Fe, Cr and Mn. And SA508, Fe, Cr, Ni and Mn were all confirmed to have MMAD around 0.1 μm in size. The results of this study show that a lot of aerosols in the range of less than 1 μm, especially around 0.1 μm in size, are generated when metal is cut using laser cutting. Therefore, in order to protect the internal exposure of workers to laser metal cutting when decommissioning NPPs, it is necessary to protect from nano-sized aerosols beyond micron size.

Gyeongju radioactive waste repository has been operated to dispose low and intermediate level radioactive waste in Korea since 2016. Currently, only deep geological disposal facility (1st) is in operation, surface disposal facility (2nd) is scheduled to operate from 2024. As a result, the annual amount of radioactive waste that can be disposed of at deep geological disposal facilities and surface disposal facilities is almost determined. According to this result, it was possible to derive the total annual disposal amount to dispose of all radioactive waste at the Gyeongju repository after landfill disposal facility (3rd) construction. To evaluate it, a predictive model has been designed and radioactive waste generation, storage, and disposal data were input. The predictive model is based on system dynamics, which is useful to analyze the correlation between input variables. As a result of analysis, radioactive waste generation amount and maximum annual radioactive waste disposal were predicted to reach 741,615 drum and 17,030 drum per year respectively. From these results, it seems that the expansion of radioactive waste acceptance system or temporary storage is necessary.

When the radioactive nuclides are leaked from a deep geological repository by groundwater, the migration path of the nuclides is mostly consisted of rock fractures to the surface biosphere. Thus, assessing the safety of the disposed radioactive wastes depends upon understanding of nuclide migration in the fractured rocks. Fractures in rocks tend to dominate the hydrological characteristics of the dissolved nuclides. To study migration of nuclides in the rock fracture, a granite block of 1 m scale was quarried from the Hwangdeung site. The block has a single natural fracture. The six faces of the rock including fracture gaps were sealed with silicone adhesives to prevent leaking or diffusion of the water. Usually flow in fractured rock is unevenly distributed and most of the water flow occures over a small portion of the fracture zone, that is so called channeling flow. It is caused by uneven distribution of apertures in a fracture field. Flow rate is proportional to the cubic of the aperture. Thus, figuring out aperture distribution in a fracture field is the most important step on the study of the migration of nuclides in the fractured region. The ideal way to figure out the aperture distribution in a fractured rock is to use a non-destructive tool such as X-ray tomagraphe. However, it has a limitation of scale, that is, less than about 30 cm. It is not easy to give a good resolution for this quarried rock of 100×60×60 cm scale. It gives complex and vague images of the fracture. The optimum way to get an aperture distribution in a fractured rock is to drill some boreholes to the fracture and to carry out hydraulic tests. The more number of boreholes gives the more accurate information, but the more disturbance to the fracture field. Thus, it is necessary to optimize between aperture information and disturbing fracture field by selecting a suitable number of boreholes. We drilled nine boreholes from the upper surface of the rock mass just to the fracture without penetrating the fracture. And we carried out dipole tests for the matrix set of 9 boreholes. From each dipole test, an effective average aperture was calculated with the data of flow rate and hydraulic head. Then aperture distribution in the fracture field is calculated with a modified Krigging method. As a result, the aperture is distributed in the range of about 0.03~0.16 mm.

Bentonite, which mostly consists of montmorillonite, is considered as a suitable buffer material for disposal of high-level radioactive wastes in deep geological repository due to its high swelling capacity, low permeability, and strong retention capacity of radionuclide migration. Alkaline and saline solutions originated from degradation of cementitious material and seawater intrusion, respectively, may cause the changes in mineralogical and chemical properties of montmorillonite with various processes such as cation exchange within the interlayer, dissolution of montmorillonite, and precipitation of second minerals. In this study, montmorillonite alteration under alkaline and saline environments and its influences on retention of cesium and iodide by bentonite buffer were investigated. The reactions of bentonite (Bentonil-WRK) with alkaline solutions (0.1 M KOH and NaOH) and simulated saline solution were performed for 7 days in batch experiments at 25°C. After the experiments, reacted bentonite samples were characterized by X-ray diffraction (XRD), Fourier Transform Infrared (FTIR) spectroscopy, Short Wavelength Infrared (SWIR) spectrometry. The concentrations of cesium and iodide dissolved in the solutions were analyzed using an inductively coupled plasma mass spectrometer (ICP–MS). The XRD patterns showed significant decrease in the interlayer space of montmorillonite after the reaction with alkaline solution due to cation exchange and change in hydration status at the interlayer. The retention of cesium and iodide in alkaline and saline solutions were affected by montmorillonite alteration and ion competition. Therefore, the montmorillonite alteration affecting the nuclide retention capacity and long-term stability of bentonite buffer should be considered in the safety assessment of long-term geological disposal performance.

Bentonite has been considered as a buffer material in a deep geological repository for high-level radioactive waste (HLW). Bentonite may come into contacted with various chemical solutions during the long-term storage. In particular, solutions containing K+ can affect stability of bentonite (e.g., illitization). The bentonite can be gradually saturated with the inflow of groundwater, and the temperature can rise simultaneously due to the decay of HLW. This study aimed to evaluate the bentonite stability in contacted with very highly concentrated K+ solutions with different pHs at 150°C. Batch reaction tests using KJ-II bentonite were performed for 30–150 days in teflon-stainless steel reactors. De-ionized (DI) water (pH = 6.0) and 1 M KCl (pH = 6.0), and 1 M KOH (pH = 12.5) solutions were used as reaction solutions. After completing batch reaction tests, the reacted samples were analyzed using various analytical techniques. For DI water, chemical, mineralogical, and physicochemical properties of reacted samples were similar to those of unreacted samples. For 1 M KCl solutions, cation exchage for Ca by K and slight changes in mineralogical properties of reacted samples were observed, but there are no significant changes in the physicochemical properties. In contrast, for 1 M KOH solutions, changes in chemical, mineralogical, and physicochemical properties of reacted samples were observed. Results of X-ray diffraction (XRD) analysis indicated dissolution of montmorillonite and formation of zeolite minerals, which were confirmed by thermogravimetricdifferential thermal analysis (TGA-DTA) and fourier transform infrared (FTIR) analysis. These results suggest that highly concentrated K+ (1 M) solution combined with high pH (12.5) and high temperate (150°C) may affect bentonite alteration. These prelimiary experiments were intended to qualitatively evaluate the mechanism and influncing factors of the buffer material alteration under extreme experimental conditions, and it is revealed that the conditions do not reflect the actual repository environment.

Safe storage of spent nuclear fuel in deep underground repositories needs an understanding of the long-term alteration (corrosion) of metal canisters and buffer materials. We conducted a small-scale laboratory alteration tests on some metal (Cu and Fe) chips by embedding them into the compacted bentonite blocks, which were placed in anaerobic water for 1 year. Some additives like lactate, sulfate, and bacteria were separately loaded into the water to promote biochemical reactions. The bentonite blocks immersed in the water were finally dismantled after 1 year, and they showed that their alteration was insignificant. However, the Cu chip exhibited some microscopic etch pits on its surface, wherein sulfur component was slightly detected. Overall, the Fe chip was more corroded than the Cu chip under the same condition. The secondary phase of the Fe chip was locally found as carbonate materials, such as siderite (FeCO3) and calcite ((Ca, Fe)CO3). These secondary products could imply that the local carbonate production around the Fe chip may be initiated by an evolution (alteration) of bentonite and a diffusive provision of biogenic CO2 gas. These laboratory scale test results suggest that the long-term alteration (corrosion) of metal canister/bentonite blocks in the engineered barrier could be possible and may be promoted by microbial activities.

In this study, we evaluated fracture filling minerals and aperture distribution along the fracture surfaces under the controlled conditions. The fractured granite block which has a single natural fracture of 1 m scale was sampled in a domestic quarry (Iksan), which groundwater had been flowed through. This rock has an interconnected porosity of 0.36 with the specific gravity of 2.57. The experimental setup with the granite block with dimensions of 100×60×60 (cm). The fracture is sealed with rock silicone rubbers when it intersects the outer surfaces of the block and the outer surfaces are coated with the silicone to prevent loss of water by evaporation. Nine boreholes were drilled of orthogonal direction at the fracture surface. A flow of de-ionized water through the fracture between pairs of boreholes was initiated and the pressure required to maintain a steady flow was measured. In additions, fracture filling minerals were sampled and examined by mineralogical and chemical analyses. There are phyllosilicate minerals such as illite, kaolinite, and chlorite including calcite, which are fracture filling minerals. The illite and kaolinite usually coexist in the fracture, where their content ratio is different according to which mineral is predominant. For the evaluation of fracture, surface was divided into an imaginary matrix of 20×20 sub-squares as schematically. The calculated results are expressed as a two dimensional contour and a three dimensional surface plot for the aperture distribution in the fracture. The aperture value is distributed between 0.075 and 0.114 mm and the mean aperture value is 0.082 mm. The fracture volume is about 49 ml. These results will be very useful for the evaluation of environmental factor affecting the nuclides migration and retardation.

The deep geological disposal system is aimed to permanently isolate the high-level radioactive waste from the biosphere through a multi-barrier system composed of engineered and natural barriers. The buffer material used for the engineered barrier should have the performance to prevent and retard the migration of radionuclides to the outside of the deep disposal facility when radionuclides are released from a disposal canister by infiltration of groundwater after a long period of time. When the hydraulic conductivity of compacted bentonite is sufficiently low, the migration of radionuclides released into the surrounding rock can be significantly reduced since they are sorbed to bentonite during the diffusion process. Therefore, an investigation on diffusion of radionuclides in compacted bentonite is a fundamental task to obtain essential data for the safety assessment of the deep geological disposal system. The migration of radionuclides by diffusion can be evaluated by diffusion coefficient. In order to obtain the apparent diffusion coefficients of Sr, Sm, and Eu in compacted Ca-bentonite (1.6 g/cm3) a through-diffusion experiment have been carrying out. A cylindrical apparatus consists of a source cell with an appropriate concentration of radionuclides and diffusion cell filled with radionuclide free solution where the concentration is gradually increased by diffusion of radionuclides. The compacted bentonite was installed between the both cells. The sample used for the experiment is a Ca-type bentonite named Bentonil-WRK, and the diffusion experiment was performed under an oxidizing condition using a synthetic groundwater simulating KURT groundwater composition. The diffusion experiment will be terminated when an increasing rate of concentration of nuclides in the diffusion cell becomes constant over time. The concentration change with regard to the geochemical characteristic of the nuclide may appear to be apparently slow.In this study, the experimental results of the through-diffusion test of Sr, Sm, and Eu in the initial stage (~4 months) were presented. Through the results of the initial stage, the period of the through-diffusion experiment can be rearranged and also it is expected that the initial results provide the qualitative and quantitative diffusion properties of each nuclide.

The purpose of this study is to provide lessons learned in the experience of improvement work of fuel handling equipment at operating nuclear power plants. The upgrade of fuel handling equipment for safety enhancement and performance improvement has been going on for 15 years since the early 2000’s. The main goal is to increase fuel loading/unloading capability of the equipment from about 2.5 fuel assemblies per hour to more than six (6). It is achieved with sequential operations of three (3) fuel handling equipment, which consists of the refueling machine, the fuel transfer system and the spent fuel handling machine. The scope of the upgrade for fuel handling equipment is summarized as follows. The PC data control system based on PLC for interlocks and high speed motor drive system is introduced for better operating efficiency. The motors and drives for bridge, trolley, and hoist are replaced with AC servomotors and drivers, respectively. The fuel transfer system has an auto-initiation feature operating from the refueling machine or the spent fuel handling machine. The winch motor and drive for the carriage of fuel transfer system is also replaced with AC servomotors and drivers. And some of HPU (hydraulic power units) equipment for each building (reactor containment building and fuel handling building) are replaced to improve their function. The considerations for improvement of fuel handling equipment are as belows. 1) Fuel handling should be consistent with the instructions provided by the fuel designer and/or manufacturer, which are for Standard type fuel and Westinghouse type fuel, used in domestic nuclear power plants. Each fuel has unique design characteristics, which are PLC setpoints for overload and underload, slow speed zones for a bridge, trolley and hoist, allowable acceleration/deceleration value in handling, hoist elevation and manual speed in off-index operation at reactor. 2) The interlock system should be designed in accordance with design criteria specified by the utilities of nuclear power plant. 3) Performance should be improved according to the operating characteristics of the fuel handling equipment. High-speed and optimization of FTS upender and carriage are essential for operating performance so that its modification should be considered first. And the low speed and range in the operation mechanism of the hoist should be designed to comply with guidelines. 4) The accident analysis through self-diagnosis function and operation history in modification at domestic operating nuclear power plants would be good lessons learned. It is advisable to utilize such various information as it can help to improve reliability of nuclear fuel handling operation and power plant operation rate.