Abstract Purpose : This study sought to find ways to improve awareness and ethics of opticians' code of ethics. Methods : A survey was conducted on 102 Korean opticians to investigate the awareness and performance satisfaction of the optician's code of ethics, as well as differences according to gender, age, education level, work experience, and workplace. Results : This study found no gender-based differences in opticians’ awareness and satisfaction with the code of ethics. However, age and education significantly influenced these factors, with higher levels observed among opticians in their 20s and 30s, particularly those in the early stages of their careers. Furthermore, a positive correlation was noted between education level and both awareness and satisfaction. The study revealed that opticians with volunteer experience exhibited higher awareness and satisfaction with the code of ethics than those without. This suggests that volunteering, which fosters professional identity and social responsibility, influences opticians’ ethical awareness. Additionally, groups with a strong interest in foreign languages and self-development demonstrated higher ethical awareness and satisfaction. Conclusion : This study underscores the need for enhanced ethical awareness among opticians. It advocates for the development of methods, case studies, and guidelines to address ethical issues not covered by the existing code, thereby revitalizing ethical practices in the profession. In addition, it is judged that active promotion and education on the code of ethics and professional ethics are necessary for the continuing education of opticians. Key words : Awareness, Code of ethics, Performance satisfaction

In 2017, the permanent shutdown of Kori Unit 1 was decided, marking the initiation of preparations for the decontamination and decommissioning of Kori Unit 1. The dismantling of radiologically contaminated equipment and concrete structures such as the Reactor Vessel (RV), Reactor Vessel Internals (RVI), and the Bio shield is crucial in the nuclear decommissioning process. These components became radiologically contaminated due to nuclear fission reactions occurring in the reactor during its operational period. The RVI dismantling at Spain’s Jose Cabrera Nuclear Power Plant involved the use of mechanical saws and disk cutters to divide it into approximately 430 pieces, taking 16 months to complete. Germany’s Stade Nuclear Power Plant employed mechanical circular saws to segment their RVI into about 170 pieces, which took 30 months to accomplish. Meanwhile, the RVI at Germany’s Wurgassen Nuclear Power Plant was subdivided into approximately 1,200 pieces using a combination of mechanical saws and abrasive water jets, requiring 61 months for completion. Due to the radioactivity in Kori Unit 1’s Reactor Vessel (RV) and Reactor Vessel Internals (RVI), remote-controlled systems were developed for cutting within the cavity to reduce radiation exposure. Specialized equipment was developed for underwater cutting operations. This paper focuses on modeling related to RVI operations using the MAVRIC code. The upper and lower parts of the RVI are classified as low-level radioactive waste, while the sides of the RVI that come into contact with fuel are classified as intermediate-level radioactive waste. Therefore, the modeling presented in this paper only considers the RVI sides since the upper and lower parts have a minimal impact on radiation exposure. Accurate calculations were performed through geometric modeling and radiation dose modeling. These research findings are anticipated to contribute to enhancing the efficiency and safety of nuclear reactor decommissioning operations

For the release of the nuclear power plant site after the decommissioning, a reliable exposure dose assessment considering the environmental impact of residual radionuclides is essentially required. In this study, the Derived Concentration Guideline Level (DCGL) for the hypothetically contaminated surface soil at the Wolsong nuclear power plant (NPP) unit 1 site was preliminarily calculated by using the RESRAD-OFFSITE computational code and compared with the other case studies. Moreover, radiation exposure dose for local residents and relevant exposure pathways were quantitatively analyzed based on the calculation model established through this work. For the target site modeling, the source term was determined by referring to the previous case studies regarding the nuclear power plant decommissioning, quantification analysis data of pressure tubes of Wolsong NPP unit 1, and radionuclide data estimated by using the MCNP/ORIGEN-2 code. In total, 14 different radioisotopes such as Ag-108m, C-14, Co-60, Cs-134/137, Fe-55, H-3, Nb-93m/94, Ni-63, Sb-125, Sn-121m, Sr-90, and Zr-93 were considered as target radionuclides. In addition, the geological structure model of the Wolsong NPP site was established based on the final safety analysis report of Wolsong NPP unit 1. The distribution coefficients (Kd) were taken from the JAEA-SDB to estimate the migration/retardation behavior of various radionuclides under the groundwater condition of the Wolsong NPP site. In the present work, the DCGL values were calculated according to the site release criterion of 0.1 mSv/yr, which indicates the radiation protection standard for the site release. Moreover, the exposure pathway and sensitivity analyses were conducted to assess the sensitive input parameters remarkably influencing the calculation result. For the evaluation of exposure dose for local residents, a site layout centered around Wolsong NPP unit 4, located in the closest proximity to the residents’ habitation area, was alternatively established and all potential exposure pathways were considered as a comprehensive resident farmer scenario. The results obtained from this study are expected to serve as a preliminary case study for the DCGL values regarding the surface soil at the Wolsong NPP unit 1 site and for evaluating the radiation exposure dose to local residents resulting from the residual radioactivity at the site after the decommissioning.

The HADES (High-level rAdiowaste Disposal Evaluation Simulator) was developed by the Nuclear Fuel Cycle & Nonproliferation (NFC) laboratory at Seoul National University (SNU), based on the MOOSE Framework developed by the Idaho National Laboratory (INL). As an application of the MOOSE Framework, the HADES incorporates not only basic MOOSE functions, such as multi-physics analysis using Finite Element Method (FEM) and various solvers, but also additional functions for estimating the performance assessment of Deep Geological Repositories (DGR). However, since the MOOSE Framework does not have complex mesh generation and data analyzing capabilities, the HADES has been developed to incorporate these missing functions. In this study, although the Gmsh, finite element mesh generation software, and Paraview, finite element analysis software, were used, other applications can be utilized as well. The objectives of HADES are as follows: (i) assessment of the performance of a Spent Nuclear Fuel (SNF) disposal system concerning Thermal-Hydraulic-Mechanical-Chemical (THMC) aspects; (ii) Evaluation of the integrity of the Engineered Barrier System (EBS) of both general and high-efficiency design perspective; (iii) Collaboration with other researchers to evaluate the disposal system using an open-source approach. To achieve these objectives, performance assessments of the various disposal systems and BMTs (BenchMark Test), conducted as part of the DECOVALEX projects, were studied regarding TH behavior. Additionally, integrity assessments of various DGR systems based on thermal criteria were carried out. According to the results, HADES showed very reasonable results, such as evolutions and distributions of temperature and degree of saturation, when compared to validated code such as TOUGH-FLAC, ROCMAS, and OGS (OpenGeoSys). The calculated data are within the range of estimated results from existed code. Furthermore, the first version of the code, which can estimate the TH behavior, has been prepared to share the contents using Git software, a free and open-source distribution system.

Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. Thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. COBRA-SFS is a legacy code based on subchannel analysis code, and its fidelity is verified for evaluating the thermal analysis for licensing a dry cask system. Herein, thermal analysis result based on CFD and COBRA-SFS codes is compared and the Dry Cask Simulator (DCS) is assessed as a benchmark experiment in this study. Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to address the degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT was used to simulate DCS. Simulations were conducted in various decay heat and helium pressure inside the canister. In realistic conditions of decay heat and helium pressure of actual dry cask system, CFD and COBRA-SFS analysis results gave good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also showed good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K. In high decay heat and high pressure condition, however, CFD and COBRA-SFS underestimated peak cladding temperature than experimental results.

본 연구에서는 BTS의 노래가사 언어 코드 비율 변화와 트위터 내 팬 덤 아미(ARMY)의 메타언어적 코멘트를 통해 BTS 현상에서의 언어 이 데올로기를 탐구하였다. 분석대상은 2013년부터 2022년까지 BTS의 앨 범에 포함된 121곡의 노래이다. 연구결과, 영어 코드 선택 비율은 꾸준 히 증가하여 2013년 21.2%에서 2022년에는 57.5%까지 상승하였으며, 반면에 한국어 코드 선택 비율은 2013년 78.8 %에서 점차 하락하여 42.5%로 감소하는 양상을 확인했다. 또한 트위터에 나타난 팬들의 메타 언어 코멘트를 분석한 결과, 언어 민족주의적 순수주의의 태도와 반-언 어 민족주의적 순수주의의 태도가 대립하는 형태를 관찰하였다. 언어 민 족주의적 순수주의의 태도를 가진 팬들은 한국어 사용을 한국적 정체성 의 실현으로 연결 짓고, 영어 코드의 증가를 우려하는 경향을 보였다. 한 편, 반-언어 민족주의적 순수주의의 태도를 가진 팬들은 BTS의 언어 코 드 선택과 정체성의 연결을 지양하며, 언어 코드 선택에 대한 비판에 저항하는 모습을 보였다. 이러한 연구결과를 바탕으로 BTS는 적절한 언 어 코드를 사용하여 글로벌 팬들과 소통함으로써 다양성과 포용성을 장 려하고, 팬덤 "아미"는 언어 선택 행위에 대한 이해와 존중을 확대함으로 써 언어 이데올로기에 대한 인식을 높일 수 있는 방안을 제공하였다는 점에서 의의가 있다.

This paper presents a study on the design and implementation of a secure contactless system leveraging Quick Response (QR) codes as a core component. The main goal of this system is to bridge the gap between strong security and improved user experience within the realm of digital interaction. The system's versatility can be expanded with broad compatibility with a variety of applications. Utility can be expanded to areas such as contactless payments, electronic ticketing, secure identity verification, and convenient access to medical records. The international standardization of QR codes ensures seamless cross-platform compatibility, strengthening their role in the digital ecosystem. We actually create and develop a non-contact security QR code system and check the expandability of the system. This study highlights the pivotal role of QR codes within the realm of secure contactless systems. Through its effective balance of digital security and user convenience, QR codes are emerging as an important element in the continued development of a secure and user-friendly digital environment. The potential for future research lies in exploring more complex use cases and further advancements that improve both security and user-centered design.

In the event of a loss of a SNF (spent nuclear fuel) transport cask during maritime transportation, it is essential to evaluate the critical depth at which the integrity of the cask can be maintained under high water pressure. SNF transport casks are classified as Type B containers and the integrity of of the containment boundary must be maintained up to a depth of 200 meters unless the containment boundary was breached under beyond-design basis accidents. However, if an intact SNF cask is lost at a depth deeper than 200-meter, release of radioactive material may occur due to breach of containment boundary with over-pressure. In this study, we developed a code for the evaluation of the pressure limit of SNF transport cask, which can be evaluated by inputting the main dimensions and loading conditions of cask. The evaluation model was coded as a computer module for ease of use. In the previous study, models with three different fidelities were developed to ensure the reliability of the calculation and maintain sufficient flexibility to deal with various input conditions. Those three models consisted of a high-fidelity model that provided the most realistic response, a low-fidelity model with parameterized simplified geometry, and a mathematical model based on the shell theory. The maximum stress evaluation of the three models confirmed that the mathematical model provides the most conservative results than the other two models. The previous results demonstrate that mathematical models can be used in the code of computer modules. In this study, additional models of transport cask were created using parametric modeling techniques to improve the accuracy of the pressure limit assessment code for different cask and situations. The same boundary conditions and loading conditions were imposed as in the previous simplified model, and the maximum stress results considering the change in the shape of the transport container were derived and compared with the mathematical model. The comparison results showed that the mathematical model had more conservative values than the simplified model even under various input conditions. Accordingly, we applied the mathematical model to develop a transportation container pressure limit evaluation code that can be simulated in various situations such as shape change and various situations.

본 연구는 東晉의 王羲之(303-361)가 쓰고 宋人(米芾(1051-1107)로 추정)이 편찬한『草訣 百韻歌』에서 시작한다. 기존 歌訣 해설과 다르게 본 연구는 전통 이론과 실제 붓을 잡고 쓴 경험을 바탕으로, 宋人이 가결을 통해 후학에게 전달하고자 했던 초서의 요체를 새롭게 밝히고 자 한다. 이로써 전통 이론 중, 초서에게 부여된 ‘積其點畫, 乃成其字’, ‘草以點畫為情性, 使轉 為形質’과 같은 간단명료한 원칙의 의미를 더욱 깊이 있게 이해할 수 있을 것이다. 또한, 초서 문자 구조에 잠재된 변화규칙과 초서의 ‘連綿’, ‘游絲’의 추론과 설명 방식을 고찰할 것이다. 이는 초서 문자의 해독과 초서 문헌의 이해할 수 있게 한다는데 의의가 있다. 본 연구의 방법 및 과정은 크게 4가지로 구분된다. (1) 『草訣百韻歌』 중 간략하게 서술된 초서 변화의 요체를 해석하고, 점을 찍는 것에서부터 선을 연결하고 문자를 구성하는 기본 ‘結字 法’을 상세하게 분석한다. ‘圓筆’, ‘潤筆’, ‘渴筆’, ‘轉筆’ 등의 필법이 포함되며, ‘楷’에서 ‘草’로 전개되는 과정에서 문자의 틀과 필획의 구조 변화도 설명한다. (2) 초서에서 가로획과 세로획의 교차 운용 법칙을 상세하게 분석하고, 나아가 ‘撇捺改筆’, ‘戈法變換’ 등 획과 선의 배치, 한자 구조의 변화 등도 아울러 분석하고 논한다. 이는 초서 쓰기, 읽기, 이해에 이론적이고 과학적 해결법을 제공할 것이다. (3) 초서 章法의 기본을 해석하고 초서 필획 중 虛･實의 분배와 운용 및 ‘計白當黑’, ‘意連形連’과 같은 필획 구성의 요체를 고찰한다. (4) ‘連綿’을 재구조화하고, 초서의 점과 획의 변화 규칙을 활용하여 ‘連綿’, ‘游絲’의 필법 원리를 해석한다. 또한 고문헌의 해석에 응용하여 ‘龍蛇競筆端’에 포함된 초서로 쓰인 한자의 수수께끼를 풀어낼 것이며, 이로써 초서 筆法字學 학습의 중요성과 그 가치를 밝힌다.

This paper summarizes and assesses the international trend, both in doctrine and in legal provisions dealing with conflict of laws in intellectual property field and reviews the relevant provisions in the Part 5th on applicable law to civil relations evolving foreign elements of the Vietnam’s 2015 Civil Code. The author shows that the two new provisions of the Vietnam’s 2015 Civil Code, namely Article 679 and Article 683, has partly caught up with the international trend in recognizing conflicts of laws and providing choice-of-law rules for resolving these conflicts in the intellectual property relations. The shortcoming of the Vietnam’s 2015 Civil Code is the absence of a particular provision dealing with the conflict of laws in case of infringement of intellectual property rights. On that basis, the paper offers comments and suggestions on the need to make the provisions of the Vietnam’s Civil Code more specific in the future.

In case a spent nuclear fuel transport cask is lost in the sea due to an accident during maritime transport, it is necessary to evaluate the critical depth by which the pressure resistance of the cask is maintained. A licensed type B package should maintain the integrity of containment boundary under water up to 200 m of depth. However, if the cask is damaged during accidents of severity excessing those of design basis accidents, or it is submerged in a sea deeper than 200 m, detailed analyses should be performed to evaluated the condition of the cask and possible scenarios for the release of radioactive contents contained in the cask. In this work, models to evaluate pressure resistance of an undamaged cask in the deep sea are developed and coded into a computer module. To ensure the reliability of the models and to maintain enough flexibility to account for a variety of input conditions, models in three different fidelities are utilized. A very sophisticated finite element analysis model is constructed to provide accurate response of containment boundary against external pressure. A simplified finite element model which can be easily generated with parameters derived from the dimensions and material properties of the cask. Lastly, mathematical formulas based on the shell theory are utilized to evaluate the stress and strain of cask body, lid and the bolts. The models in mathematical formula will be coded into computer model once they show good agreement with the other two model with much higher fidelity. The evaluation of the cask was largely divided into the lid, body, and bottom, bolts of the cask. It was confirmed that the internal stress of the cask was increased in accordance with the hydrostatic pressure. In particular, the lid and bottom have a circular plate shape and showed a similar deformation pattern with deflection at the center. The maximum stress occurred where the lid was in the center and the bottom was in contact with the body. Because the body was simplified and evaluated as a cylinder, only simple compression without torsion and bending was observed. The maximum stress occurred in the tangential direction from the inner side of the cylinder. The bolt connecting the lid and the body was subjected to both bending and tension at the same time, and the maximum stress was evaluated considering both tension and bending loads. In general, the results calculated by the formulas were evaluated to have higher maximum stresses than the analysis results of the simplified model. The results of the maximum stress evaluation in this study confirms that the mathematical models provide conservative results than the finite element models and can be used in the computer module.

The measurement activities to evaluate material balance of nuclear material are usually performed by operator. It is because that the IAEA does not have enough manpower to carry out nuclear measurement accountancy of all nuclear materials in the world. Therefore, the IAEA should consider scenarios which facility operator tries to divert nuclear material for misuse by distorting measurement record. It is required to verify the operator’s measurement data whether it is normal or not. IAEA measures inventory items using their own equipment which is independent of facility operator equipment for verification. Since all inventory lists cannot be verified due to limited resources, the number of items to be verified is determined through statistical method which is called as sample size calculation. They measure for the selected items using their own equipment and compares with operator’s record. The IAEA determines sample size by comprehensively considering targeted diverted nuclear material amount and targeted non-detection probability and performance of measurement equipment. In general, the targeted diverted nuclear material amount is considered significant quantity (plutonium: 8 kg, uranium-235: 75 kg). If the targeted non-detection probability or the performance of the verification equipment is low, the sample size increases, and on the contrary, in the case of high non-detection probability or good performance of verification equipment, even a small sample size is satisfied. It cannot be determined from a single sample size calculation because there are so many sample size combinations for each verification equipment and there are many diversion scenarios to be considered. So, IAEA estimates initial sample size based on statistical method to reduce calculation load. And then they calculate non-detection probability for a combination of initial sample size. Through the iteration calculation, the sample size that satisfies the closest to the target value is derived. The sample size calculation code has been developed to review IAEA’s calculation method. The main difference is that IAEA calculates sample size based on approximate equation, while in this study, sample size is calculated by exact equation. The benchmarking study was performed on reference materials. The data obtained by the code show similar results to the reference materials within an acceptable range. The calculation method developed in this study will be applied to support IAEA and domestic inspection activities in uranium fuel fabrication facility.

The role of QR Code robots in smart logistics is great. Cognitive robots, such as logistics robots, were mostly used to adjust routes and search for peripheral sensors, cameras, and recognition signs attached to walls. However, recently, the ease of making QR Codes and the convenience of producing and attaching a lot of information within QR Codes have been raised, and many of these reasons have made QR Codes recognizable as visions and others. In addition, there have been cases in developed countries and Korea that control several of these robots at the same time and operate logistics factories smartly. This representative case is the KIVA robot in Amazon. KIVA robots are only operated inside Amazon, but information about them is not exposed to the outside world, so a variety of similar robots are developed and operated in several places around the world. They are applied in various fields such as education, medical, silver, military, parking, construction, marine, and agriculture, creating a variety of application robots. In this work, we are developing a robot that can recognize its current position, move and control in the directed direction through two-dimensional QR Codes with the same horizontal and vertical sides, and the error is to create a QR Code robot with accuracy to reach within 3mm. This paper focuses on indoor mobile robot position recognition and driving experiment using QR Code during the development of QR Code-aware indoor mobility robots.

현대의 기술은 비약적으로 발전하고 있으나 그러한 기술적 진보가 안전수준 의 향상도에 직접적으로 비례하지는 않는다. 새로운 기술과 그것의 운용은 또 다른 위험요소를 낳기 때문이다. 본 논문은 현대 해사에서 최신기술의 집약체 인 자율운항선박의 안전관리를 현행 ISM Code상 규정에 근거하여 검토한다. 먼저 ISM Code상 핵심주체들의 역할과 책임을 자율운항선박에 적용 가능한지 분석해 보고, 특히 ISM Code에서 정의하는 선장의 책임과 권한이 자율운항선 박에서 어떻게 적용되어야 할 것인지를 집중적으로 살펴본다. 또한 현행 ISM Code상 핵심주체의 역할에 더하여 원격운항센터, 원격운항자, 사이버보안 등 추가적으로 ISM Code에 규정되어야 하는 부분에 대해서도 접근 방향을 제시한 다. 특히 원격운항의 도입에 따른 공백을 채우기 위해 자율운항 단계별로 새로 운 역할의 도입을 제안한다. 이로써 이 논문은 자율운항선박에 대한 효과적인 안전관리를 위한 안전관리체제의 틀을 제시한다.

In order to improve the implementation of safety and health education at the site for industrial accident prevention activities, research was conducted to minimize inconvenience and increase utilization by redesigning and developing existing education methods. To date, occupational safety and health education has been conducted without considering the general work characteristics and functional facilities (mechanical, electrical, instrumentation, chemical) of workers (mechanical: turbine, valve, pump, hydraulic system, electrical: generator, breaker, motor, etc.). In particular, plant facilities were classified as mechanical and electrical facilities to improve the methodology for industrial safety and health education for plant maintenance workers. In addition, the “One Page Education Plan” was announced as a learning case because the spread of COVID-19 infectious diseases made it impossible to reduce or control the number of people in all groups and groups. The improvement of this training method will play a major role in improving the effectiveness of safety education in power plant workplaces.