Synthesis of extremely competent materials is of great interest in addressing the energy storage concerns. Manganese oxide nanowires ( MnO2 NWs) are prepared in situ with multiwall carbon nanotubes (MWCNT) and graphene oxide (GO) using a simple and effective hydrothermal method. Powder XRD, Raman and XPS analysis are utilized to examine the structural characteristics and chemical state of composites. The initial specific discharge capacity of pure MnO2 NWs, MnO2 NWs/ MWCNT and MnO2 NWs/rGO composites are 1225, 1589 and 1685 mAh/g, respectively. The MnO2 NWs/MWCNT and MnO2 NWs/rGO composites showed stable behavior with a specific capacity of 957 and 1108 mAh/g, respectively, after 60 cycles. Moreover, MnO2 NWs/rGO composite sustained a specific capacity of 784 mAh/g, even after 250 cycles at a current density of 1 A/g showing outstanding cycling stability.
해사영어는 선박 운항, 해양 안전, 선내 의사소통 및 선외 교신을 위해 설계된 특수한 영어 언어체계이다. 국제해사기구 STCW(선원의 훈련, 자격증명 및 당직근무의 기준에 관한 국제협약)에 따르면 국제항해에 종사하는 항해사가 되기 위해서는 SMCP를 포함한 해사영어 대한 충분한 이해가 수반되어야 한다. 본 연구는 음성인식, 번역, 단어 기입 등 유형의 해사영어시험을 통하여 학생들 의 해사영어 활용 능력을 측정하고 플랫폼 사용에 따른 시험 점수 향상 정도, 나아가 초임항해사로 나가기 위하여 요구되는 해사영어 시험 플랫폼 활용 시간 등을 조사하고자 하였다. 실험은 먼저 초기 시험을 통해 학생들의 일반영어능력과 SMCP 활용 능력에 대한 연 관성을 조사한 후, 중간 시험 및 최종 시험을 통해 플랫폼 활용에 대한 점수 향상 정도, 응시시간 변화 등 요인을 측정하였다. 초기 시 험을 통해 개인 요인(예: 토익 점수, 본인 스스로에 대한 영어능력 평가)에 따른 그룹 간 해사영어시험 점수에 유의한 차이가 있음을 확인하였으며, 중간시험 및 최종시험을 통해 플랫폼 활용이 유의한 시험점수 향상으로 이어졌음을 확인하였다. 해당 연구는 해사 교육 분야에 다양하게 적용할 수 있는 학습 플랫폼 활용 효능을 조사하였으며 향후 해사영어 교육 외 그 범위를 넓혀 활용될 수 있을 것으 로 사료된다.
Abstract Purpose : The aim of this study was to identify knowledge and lifestyle habits for glaucoma prevention and to present basic data for future glaucoma prevention education. Methods : From April to September 2023, 447 people in their 20s to 60s were selected and an online survey was conducted. The questionnaire consisted of a total of 30 questions, including 6 questions about the subject's general characteristics, 14 questions about knowledge and awareness of glaucoma, and 10 questions about lifestyle habits related to intraocular pressure. Results : 89.7% responded that they had heard of glaucoma, and 31.1% of respondents knew the difference between glaucoma and cataracts. 34.5% responded that early detection is difficult, 33.3% that complete treatment is difficult, 23.5% that even normal intraocular pressure can cause glaucoma, and 36.0% that the incidence of glaucoma is increasing day by day. The level of awareness and knowledge was higher with age, in women than men, in those with a family history, in professional occupations, with metabolic diseases, and with high myopia. Lifestyle habits related to intraocular pressure were higher with age, among women, when smartphone use was less than 3 hours, and among housewives and professionals. The relationship between glaucoma knowledge and awareness and intraocular pressure-related lifestyle habits was statistically significant (r=0.248, p<0.001). Conclusion : Although the level of basic knowledge about glaucoma was high, the level of specific knowledge was low compared to the increase in the incidence of glaucoma over the past 10 years. Key words : Glaucoma, Healthy Life Practices, Intraocular pressure
Background: Estrus in cows can be detected through vaginal electrical resistance or conductivity. However, there are no studies measuring vaginal electrical resistance in Hanwoo cows. This study aims to measure the vaginal electrical resistance value in Hanwoo cows and compare it with estrus and ovulation. Methods: Vaginal electrical resistance values of 73 Hanwoo cows were measured before and after estrus at the Gyeongsangbuk-do Livestock Research Institute. Measurements were taken on days -6, -3, -2, -1, 0, 1, 2, 3, and 6 of artificial insemination. Large follicles and ovulation were confirmed using transvaginal ultrasonography. Results: The vaginal electrical resistance averaged 225.6 ± 6.3 Ω days before the artificial insemination date, decreasing until the day of artificial insemination. The average vaginal electrical resistance was 163.7 ± 4.6 Ω on the date of artificial insemination, and 188.8 ± 4.3 Ω one day after artificial insemination, when large follicles were observed. In addition, on the 6th day after artificial insemination, the vaginal electrical resistance averaged 231.4 ± 5.5, which was similar to the 6th day before artificial insemination (222.5 ± 6.3). Transvaginal ultrasonography showed that most of the cows ovulated one day after artificial insemination. Conclusions: The accuracy of estrus is high if the vaginal electrical resistance is measured for cows with confirmed estrus, making is a potentially useful for determining the timing of artificial insemination.
This study presents a rapid and sequential radiochemical separation method for Pu and Am isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin and TRU resin. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the sample was allowed to evaporate to dryness after filtering the leaching solution with 0.45 micron filter. The Pu isotopes were separated in HNO3 medium with anion exchange resin. For leaching solution passed through anion exchange column, the Am isotopes were separated with TRU resin. The purified Pu and Am isotopes were measured by alpha spectrometer, respectively, after micro-precipitation of neodymium. The sequential radiochemical separation of Pu and Am isotopes in radioactive waste samples using anion exchange resin and TRU resin was validated with ICP-MS system.
In this study, we introduce the validation of the analysis guidelines through preliminary experiments of the draft analysis guidelines before analyzing waste materials (non-combustible). This validation data was applied the accuracy and efficiency of the separation and analysis for the waste such as steel generated from NPP. Steel (non-flammable) was leached the mixed acid and the leaching solution was separated by using the separation guidelines. Steel was corroded with radioactive RM (Co-60, Cs-137) and mixed acid. After drying, the corroded steel was measured the initial radioactivity by a HPGe detector (10,000 seconds). The sample was inserted in a beaker and leached with mixed acid (10 M HNO3 + 4 M HCl) for 2 hours. In this solution, it added 2 ml of H2O2 to increase the leaching effect. The ultrasonic device was adjusted so that the temperature does not exceed 60°C. After elution, the surface of the sample was washed with pure water. The weight of the sample was measured accurately, and recorded the weight loss rate after infiltration. The leaching sample was measured radioactivity by a HPGe detector (10,000 seconds). It was calculated the recovery rate based on the difference in total radioactivity before and after leaching. Before the test, radioactive RM (Co-60, Cs-137) was radioactive deposited by corrosion, but Cs- 137 was not detected in the initial gamma measurement and only Co-60 nuclides were deposited. The recovery rate test results were confirmed to be about 100%.
The nuclear licensee must ensure that the nuclear or radiological emergency preparedness and response organization is explicitly defined and staffed with adequate numbers of competent and assessed personnel for their roles. This paper describes the responsibilities of medical and support personnel for the medical action of casualties in the event of a radiological emergency at the KAERI. Currently, there is one medical personnel (nurse) in KAERI, and a total of eight medical support personnel are designated for medical response in the event of a radiological emergency. These medical support personnel are designated as one or two of the on-site response personnel at each nuclear facility, operating as a dedicated team of A, B (4 people each). In the event of a radiological emergency, not all medical support personnel are mobilized, but members of the dedicated medical team, which includes the medical support personnel of the nuclear facility where the accident has occurred, are summoned. Medical and support personnel will first gather in the onsite operational support center (OSC)/technical support center (TSC) to prepare and stand by for the medical response to injured when a radiological emergency is declared. They should take radiation protective measures, such as wearing radiation protective clothing and dosimeters, before entering the onsite of a radiological emergency, because injuries sustained during a radiological emergency may be associated with radioactive contamination. In the event of an injury, direct medical treatment such as checking the patient’s vitals, first aid, and decontamination will be carried out by medical personnel, while support personnel are mainly responsible for contacting the transfer hospital, reporting the patient’s condition, accompanying the ambulance, filling out the emergency medical treatment record, and supporting medical personnel. In order to respond appropriately to the occurrence of injuries, we regularly conduct emergency medical supplies education and medical training for medical support personnel to strengthen their capabilities.
When the parent radionuclide decays, the progeny radionuclide is produced. Accordingly, the dose contribution of the progeny radionuclide should be considered when assessing dose. For this purpose, European Commission (EC) and International Atomic Energy Agency (IAEA) provide weighting factors for dose coefficient. However, these weighting factors have a limitation that does not reflect the latest nuclide data. Therefore, in this study, we analyzed the EC and IAEA methodology for derivation of weighting factor and used the latest nuclide data from ICRP 107 to derive weighting factors for dose coefficient. Weighting factor calculation is carried out through 1) selection of nuclide, 2) setting of evaluation period, and 3) derivation based on ICRP 107 radionuclide data. Firstly, in order to derive the weighting factor, we need to select the radionuclides whose dose contribution should be considered. If the half-life of progeny radionuclides sufficiently short compared to the parent radionuclide to achieve radioactive equilibrium, or if the dose coefficient is greater of similar to that of the parent radionuclide and cannot be ignored, the dose contribution of the progeny radionuclide should be considered. In order not to underestimate the dose contribution of progeny radionuclides, the weighting factors for the progeny nuclides are taken as the maximum activity ratio that the respective progeny radionuclides will reach during a time span of 100 years. Finally, the weighting factor can be derived by considering the radioactivity ratio and branch fraction. In order to calculate the weighting factor, decay data such as the half-life of the radionuclide, decay chain, and branch fraction are required. In this study, radionuclide data from ICRP 107 was used. As a result of the evaluation, for most radionuclides, the weighting factors were derived similarly to the existing EC and IAEA weighting factors. However, for some nuclides, the weighting factors were significantly different from EC and IAEA. This is judged to be a difference in the half-life and branch fraction of the radionuclide. For example, in the case of 95Zr, the weighting factor for 95mNb showed a 35.8% difference between this study and previous study. For ICRP 38, when 95Zr decays, the branch fraction for 95mNb is 6.98×10-3. In contrast, for ICRP 107, the branch fraction is 1.08×10-2, a difference of 54.7%. Therefore, the weighting factor for the dose coefficient based on ICRP 107 data may differ from existing studies depending on the half-life and decay information of the nuclide. This suggests the need for a weighting factor based on the latest nuclide data. The results of this study can be used as a basis for the consideration of dose contributions for progeny radionuclides in various dose assessments.
The operation of nuclear facilities involves the potential for on-site contamination of soil, primarily resulting from pipe leaks and other operational incidents. Globally, decommissioning process for commercial nuclear power plants have revealed huge-amounts of soil waste contaminated with Cs-137, Sr-90, Co-60, and H-3. For example, Connecticut Yankee in the United States produced approximately 52,800 ton of contaminated soil waste, constituting 10% of the total waste generated during its decommissioning. Environmental remediation costs associated with nuclear decommissioning in the US averaged $60 million per unit, representing a significant 10% of the whole decommissioning expenses. Consequently, this study undertook a preliminary investigation to identify important factors for establishing a site remediation strategy based on radionuclide- and site-specific media- characteristics, focusing the efficiency enhancement for the environmental remediation. The factors considered for this investigation were categorized into physical/environmental, socioeconomic, technical, and management aspects. Physical/environmental factors contained the site characteristics, contamination levels, and environmental sensitivity, while socio-economic factors included the social concerns and economic costs. Technical and management factors included subcategories such as technical considerations, policy aspects, and management factors. Especially, technical factors were further subdivided to consider the site reuse potential, secondary waste generation by site remediation, remediation efficiency, and remediation time. Additionally, our study focused the key factors that facilitate the systematic planning for the site remediation, considering the distribution coefficient (Kd) and hydrogeological characteristics associated with each radionuclide in specific site conditions. Therefore, key factors in this study focus the geochemical characteristics of site media including the particle size distribution, chemical composition, organic and inorganic constituents, and soil moisture content. Moreover, the adsorption properties of site media were examined concerning the distribution coefficient (Kd) of radionuclides and their migration characteristics. Furthermore, this study supported the development of a conceptual framework, containing the remediation strategies that incorporate the mobility of radionuclides, according to the site-specific media. This conceptual framework would necessitate the spatial analysis techniques involving the whole contamination surveys and radionuclide mobility modeling data. By integrating these key factors, the study provides the selection and simulation of optimal remediation methods, ultimately offering the estimated amounts of radioactive waste and its disposal costs. Therefore, these key factors offer foundational insights for designing the site remediation strategies according the sitespecific information such as the distribution coefficient (Kd) and hydrogeological characteristics.
KEPCO KPS is the contractor for the full system decontamination (FSD) of Kori Unit 1 and under preparation such as modification, lay out for equipment installation, setting up tie-in/out point for chemical injection and way to pressurize the system, of its successful performance. In this research, KPS introduced how KPS has designed and prepared for the FSD project and how will the chemical decontamination process be implemented. As described in the previous research, chemical decontamination process is planned to be conducted for three cycles and each cycle is consisted of oxidation, reduction, decomposition, and purification. Oxidation and reduction process were conducted at 90°C. Chemical decomposition and purification process were conducted at 40°C due to the damage of IX by the heat. If the decontamination result does not meet the target DF and the dose rate, additional cycle can be conducted. Expected volume of process water for FSD is 200 m3. Three systems have been designated as decontamination targets: reactor coolant system (RCS), residual heat removal system (RHRS), chemical volume control system (CVCS). For the steady flow rate, existed plant equipment such as reactor coolant pump (RCP) will be operated and modifications on some components will be conducted. Due to the limited space for installation, decontamination equipment and other resources are distributed to three different places. KPS designed the layout of equipment installed inside the containment vessel. The layout contains the information of shielding for highly radiated equipment such as IX and filter skid.
The radiological characterization of SSCs (Structure, Systems and Components) plays one of the most important role for the decommissioning of KORI Unit-1 during the preparation periods. Generally, a regulatory body and laws relating to the decommissioning focus on the separation and appropriate disposal or storage of radiological waste including ILW (intermediate level waste), LLW (low level waste), VLLW (very low level waste) and CW (clearance waste), aligned with their contamination characteristics. The result of the preliminary radiological characterization of KORI Unit-1 indicated that, apart from neutron activated the RV (reactor vessel), RVI (reactor vessel internals), and BS (biological shielding concrete), the majorities of contamination were sorted to be less than LLW. Radiological contamination can be evaluated into two methods. Due to the difficulties of directly measuring contamination on the interior surfaces of the pipe, called CRUD, the assessment was implemented by modeling method, that is measuring contamination on the exterior surfaces of the pipes and calculating relative factors such as thickness and size. This indirect method may be affected by the surrounding radiation distribution, and only a few gamma nuclides can be measured. Therefore, it has limitation in terms of providing detailed nuclide information. Especially, α and β nuclides can only be estimated roughly by scaling factors, comparing their relative ratios with the existing gamma results. To overcome the limitation of indirect measurement, a destructive sampling method has been employed to assess the contamination of the systems and component. Samples are physically taken some parts of the systems or components and subsequently analyzed in the laboratory to evaluate detailed nuclides and total contamination. For the characterization of KORI Unit-1, we conducted the radiation measurement on the exterior surfaces of components using portable instruments (Eberline E-600 SPA3, Thermo G20-10, Thermo G10, Thermo FH40TG) at BR (boron recycle system) and SP (containment spray system) in primary system. Based on these results, the ProUCL program was employed to determine the destructive sample collection quantities based on statistical approach. The total of 5 and 8 destructive sample quantities were decided by program and successfully collected from the BR and SP systems, respectively. Samples were moved to laboratory and analyzed for the detail nuclide characteristics. The outcomes of this study are expected to serve as valuable information for estimating the types and quantities of radiological waste generated by decommissioning of KORI Unit-1.
The treatment process for Spent Filter(SF) of Kori-1 was developed that includes the following : 1) Taking out by robot system 2) Screening by ISOCS 3) Collection of representative samples using a sampling machine 4) Compression 5) Immobilization 6) Packaging and nuclide analysis and 7) Delivery/disposal. Although the robot system, ISOCS, sampling machine and immobilization facility are essentially required for building the above processing but decision to build the compression system and nuclide analysis system must be made after reviewing the need and cost benefit for their construction. In addition, for effcient SF treatment, it is necessary to determine the nuclide concentration range of the SF to which immobilization will be applied. In this study, a cost benefit analysis was performed on existing and alternative methods for processes related to compression treatment, nuclide analysis and immobilization methods, which are greatly affected by economics and efficiency according to the design. First, although the disposal cost is reduced with reducing the number of packaging drums by compressed and packaged but the expected benefits not be equal to or greater than the cost invested in building a compression system. As a result, non-compressed treatment of SF is expected to be economical because the construction cost of compression system is more expensive than the benefits of reducing disposal costs by compression. Second, a cost benefit analysis of direct and indirect nuclide analysis methods was performed. For indirect analysis, scaling factors should be developed and the drum scanner suitable for the analysis for DAW should be improved. As a result, direct analysis applied grouping options is expected to be more economical than indirect analysis requiring the cost for developing scaling factors and improving the scanner. Third, it is timeconsuming and inefficient to distinguish and collect filters that are subject to be immobilized according to the waste acceptance criteria among the disorderly stored SFs in the filter rooms. If the benefits of immobilization of the SFs selectively are not greater than the benefits of immobilization of all SFs, it can be economical to immobilize all SFs regardless of the nuclide concentration of them. As a result, it is more economical to immobilize all SFs with various nuclide concentrations than to selectively immobilize them. The conclusion of this study is that it is not only cost-effective but also disposal-effective to design the treatment process of SF to adopt non-compressed processing, direct analysis and immobilization of all SFs.
Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF) for fuel examinations. The facility has pools and hot cells for handling and examining fuel assemblies and rods. Among the hot cells, the second cell is for measuring rod internal pressure (RIP) and then cutting the rod to make samples for destructive tests. Currently, the cutting machine is broken, so it has to be replaced. Because the existing cutting machine consists of many parts and its size was quite a bit large to handle and treat for the radioactive waste disposal, the disassembly work has been performed to make it smaller using manipulators. The drawings of the cutting machine were reviewed and the disassembly tools were developed considering workability when the work performed at the hot cell using the manipulators. The large parts such as motor, mirror and cable, etc., were able to be disassembled and the machine size became so smaller that it could be easily handled for the disposal.
Hydrogen isotope separation involves the separation of hydrogen, deuterium, tritium, and their isotopologues. It is an essential technology for removing radioactive tritium contamination and for obtaining valuable hydrogen isotope resources. Among various hydrogen isotope separation technologies, water electrolysis technology exhibits a high separation factor. Consequently, the electrolysis of tritiated water is of paramount importance as a tritium enrichment method for treating tritium-contaminated water and for analyzing tritium in environmental samples. More recently, hydroelectrolysis technology, which utilizes proton exchange membranes (PEM) to reduce water inventory, has gained favor over traditional alkaline hydroelectrolysis. Nevertheless, it is crucial to decrease the hydrogen permeability of the PEM in order to mitigate the explosion risk associated with tritium hydrogen electrolysis devices. Additionally, efforts are needed to enhance the hydrogen isotope selectivity of the PEM and optimize the manufacturing process of the membrane-electrode assembly (MEA), thereby improving both hydrogen isotope separation performance and water electrolysis efficiency. In this presentation, we will delve into two key aspects. Firstly, we’ll explore the reduction of hydrogen permeability and the enhancement of the hydrogen isotope separation factor in PEM through the incorporation of 2D nanomaterial additives. Secondly, we’ll examine the influence of various MEAs preparation methods on electrolysis and isotope separation performances. Lastly, we will discuss the effectiveness of the developed system in separating deuterium and tritium.