밀은 세계 3대 주요 작물이지만 우리나라는 대부분 수입에 의존하고 있으며, 자급률이 1%도 도달하지 못한 실정이다. 국내 밀 품종은 40여 종 이상으로 지속적으로 개발되고 있으며, 품종마다 용도와 농업적 특성이 달라 목표 형질에 맞는 적절한 품종을 선택하여 재배하는 것이 중요하다. 품종을 정확하고 빠르게 판별하기 위해선 분자 마커 기술이 필요하다. 분자 마커는 생육 환경과 시기에 영향을 받지 않고 다양한 품종을 신속하고 정확하게 구분할 수 있어 유전적 변이성, 농업 형질 등을 분석하기에 유용하다. 본 연구는 기존에 보고된 국내 밀 32품종에 대한 SCAR (Sequence Characterized Amplified Region) 판별 마커를 재검정하여 재현성이 높은 마커를 선별하였으며, 기존에 검정되지 않은 국내 밀 9품종에 추가 적용하여 비교분석하였다. 15개의 마커 세트 중 6개의 마커가 재현성과 정확도가 높은 것으로 확인되었고, 최종적으로 국내 41품종 중 4, 7, 8, 10, 11, 12번 마커를 이용하여 다홍, 금강, 밀성, 조은, 수강, 한백, 조광, 영광을 판별할 수 있었다. 또한, 4번 마커를 통한 증폭산물의 염기서열 분석을 통해 SNP (Single Nucleotide Polymorphism)를 발굴하여 ‘올밀’에 대한 새로운 품종 판별 마커인 SdHRM1과 SdHRM2를 개발하여 HRM (High Resolution Melts) 분석을 시행하여 육안으로 판단하기 어려운 SNP를 정확하게 판별할 수 있었다. 품종간의 SNP를 활용한 품종 마커 기술은 다양한 농업형질에 적용할 수 있으며, 분자 육종 프로그램에 실질적인 도움이 되는데 큰 기여를 할 것으로 사료된다.

With the aging of nuclear power plants (NPPs) in 37 countries around the world, 207 out of 437 NPPs have been permanently shutdown as of August 2022 according to the IAEA. In Korea, the decommissioning of NPPs is emerging as a challenge due to the permanent shutdown of Kori Unit 1 and Wolsong Unit 1. However, there are no cases of decommissioning activities for Heavy Water Reactor (HWR) such as Wolsong Unit 1 although most of the decommissioning technologies for Light Water Reactor (LWR) such as Kori Unit 1 have been developed and there are cases of overseas decommissioning activities. This study shows the development of a decommissioning waste amount/cost/process linkage program for decommissioning Pressurized Heavy Water Reactor (PHWR), i.e. CANDU NPPs. The proposed program is an integrated management program that can derive optimal processes from an economic and safety perspective when decommissioning PHWR based on 3D modeling of the structures and digital mock-up system that links the characteristic data of PHWR, equipment and construction methods. This program can be used to simulate the nuclear decommissioning activities in a virtual space in three dimensions, and to evaluate the decommissioning operation characteristics, waste amount, cost, and exposure dose to worker. In order to verify the results, our methods for calculating optimal decommissioning quantity, which are closely related to radiological impact on workers and cost reduction during decommissioning, were compared with the methods of the foreign specialized institution (NAGRA). The optimal decommissioning quantity can be calculated by classifying the radioactivity level through MCNP modeling of waste, investigating domestic disposal containers, and selecting cutting sizes, so that costs can be reduced according to the final disposal waste reduction. As the target waste to be decommissioning for comparative study with NAGRA, the calandria in PHWR was modeled using MCNP. For packaging waste container, NAGRA selected three (P2A, P3, MOSAIK), and we selected two (P2A, P3) and compared them. It is intended to develop an integrated management program to derive the optimal process for decommissioning PHWR by linking the optimal decommissioning quantity calculation methodology with the detailed studies on exposure dose to worker, decommissioning order, difficulty of work, and cost evaluation. As a result, it is considered that it can be used not only for PHWR but also for other types of NPPs decommissioning in the future to derive optimal results such as worker safety and cost reduction.

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.

Material balance evaluation is an important measure to determine whether or not nuclear material is diverted. A prototype code to evaluate material balance has been developed for uranium fuel fabrication facility. However, it is difficult to analyze the code’s functionality and performance because the utilization of real facility data related to material balance evaluation is very limited. It is also restricted to deliberately implement various abnormal situations based on real facility data, such as nuclear diversion condition. In this study, process flow simulator of uranium fuel fabrication facility has been developed to produce various process data required for material balance evaluation. The process flow simulator was developed on the basis of the Simulink-SimEvents framework of the MathWorks. This framework is suitable for batch-based process modeling like uranium fuel fabrication facility. It dynamically simulates the movement of nuclear material according to the time function and provides process data such as nuclear material amount at inputs, outputs, and inventories required for Material Unaccounted For (MUF) and MUF uncertainty calculation. The process flow simulator code provides these data to the material balance evaluation code. And then the material balance evaluation code calculates MUF and MUF uncertainty to evaluate whether or not nuclear material is diverted. The process flow simulator code can simulate the movement of nuclear material for any abnormal situation which is difficult to implement with real process data. This code is expected to contribute to checking and improving the functionality and performance of the prototype code of material balance evaluation by simulating process data for various operation scenarios.

For national security, approximately 500 facilities have been designated as national security facilities. It is divided into grades A, B, and C depending on the fields such as electrics, energy and public service. In 2016, KINAC developed and suggested the installation standards and evaluation indicators for security equipment for national security facilities and they have been utilized to evaluate and manage the facilities. In order to update them by reflecting the recent security trend of the facilities and new technology, KINAC have investigated and updated the installation standards and evaluation indicators for security equipment. As a part of developing installation standards and evaluation indicators for national security facilities project, we conducted survey to identify the status and capabilities of facilities: how the security system is well designed, which type of security equipment is utilized. According to the results, the passive infrared (PIR) sensor is commonly used for interior intrusion detection (not included in 2016 project). Therefore, we investigated the passive infrared sensor and provided its installation standards and evaluation indicators. PIR sensor detects the energy generated by the objects which is different as their temperatures. PIR sensors do not emit any energy like active infrared sensor, and do not measure the energy. Rather, PIRs measure variation in thermal radiation. PIR sensor detects thermal energy by sensing the change between a heat source and the background temperature. In this paper, the characteristics and features of PIR sensor were discussed. Considering them, the standards and indicators to install/evaluated the PIR sensors were also proposed. The results would support national security facility to design and evaluate their security system.

Nuclear power plants, like other national critical infrastructures, could be under the threat of terrorism or other malicious action. Thus, a nuclear power plant has a robust security system that includes security guards, sensors, barriers, access control systems, lights, and alarm stations with security procedures. However, an effective security system is hard to design because a chain is only as strong as its weakest link, and there could be a vulnerable hole even in the robust security system. Thus, an effective security system requires the evaluation of all possible scenarios. Evaluation software for security system effectiveness assists in systematically assessing all the possible attack scenarios. Many countries developed security effectiveness evaluation software. The first software was developed by the U.S. Sandia National Laboratories in the 1980s. Now there are several commercially available software packages with a function to simulate limited-scope combat between security guards and attacking enemies. However, academic communication is comparatively weak because it may contain sensitive information on the vulnerability of nuclear power plants. We developed original software called Tools for Evaluating Security Systems (TESS) to identify the most vulnerable path to the designated target and model the security systems of all South Korean nuclear power plants. We also used commercial security effectiveness evaluation software, AVERT, to model the same nuclear power plants. TESS was developed to verify the results of commercial security effectiveness evaluation software for the purpose of regulatory use. For the feasibility test, we compared the results of two software with those of force-on-force (FoF) exercises in nuclear power plants. According to the relevant Act, every nuclear power plant site should perform the FoF exercises every year. KINAC was in charge of evaluating the FoF exercise and used several of its results for the study. In the results, even in some differences in detail, the two software and FoF exercises showed qualitative similarity. Conclusively, evaluation software is a useful tool to design and/or assess the security systems of nuclear power plants. We modeled the security systems of all South Korean nuclear power plants, and compared the developed software, a commercial software and FoF exercises. The results showed qualitative similarity. We provided the results of evaluation to nuclear operators for the better security of nuclear power plants.

This study aims to develop the designs and patterns of yoga tops that are better adjusted to suit females in their 30s and 40s. After conducting a comparative analysis of three different popular yoga garments, one yoga top currently on the market was selected. Subsequently, a fit evaluation was conducted on Trail 1-garment α, which was developed body analysis performed based on selected yoga top C, followed by the production of the Trial 2 garment after making adjustments according to the comparative observation results. Based on these results, garment C with the longest top length was evaluated as the best. The results of the evaluation of appearance and fit conducted of Trial 1-garment α compared to those of C showed that Trial 1-garment α was superior in both evaluations. Trial 2-garment β was produced after making improvements on Trial 1-garment α and then placed under identical comparative evaluation condition as Trial 1-garment α. Results showed a significant improvement compared to Trial 1-garment α, and the Trial 2 garment with an additional arm pattern was shown to be superior in shoulder strap width stability, shoulder strap pressure, chest stability, degree of waist pressure, waist comfort, general fitting, and supportiveness.

해양사고의 감소를 위하여 국제해사기구(IMO) 체약국 정부에서 IMO 협약 및 규정에 대한 이행은 필수적이다. IMO 협약이행코 드(III Code)에서는 각 주관청에게 협약이행을 주기적으로 점검 및 평가하여 개선하는 체계를 갖추도록 요구하고 있으며 성과지표를 평가방 법에 포함하도록 요구하고 있음에 따라 회원국은 자국의 협약이행 평가를 위한 성과지표를 개발할 필요가 있다. 이에, 본 논문의 목적은 주관청의 협약이행 수준을 계량화하여 평가할 수 있도록 대표적인 평가측정 기법인 균형성과표(BSC) 및 핵심성과지표(KPI)를 활용한 평가 모델을 개발하여 제시하는 것이다. 논문의 결과로서, III Code 요건에 BSC 개념을 적용한 III-BSC 관점에서 수립된 비전의 달성을 위해 반드 시 확보해야 하는 핵심성공요인(CSF)을 도출하고, 도출된 CSF를 측정할 수 있도록 평가영역별 후보 KPI를 개발하여 4계층의 초기 연구모형 을 설계하였으며, SMART 기법을 활용한 조사설계를 통해 KPI의 타당성을 검증하여 연구모형을 완성하였다. 또한, 개발한 연구모형을 기반 으로 AHP 분석을 통해 각 계층별 요소의 가중치를 도출하여 BSC 기반 IMO 협약이행에 대한 평가모델을 개발하여 제시하였다. 개발된 평 가모델이 정량적으로 IMO 협약이행 수준을 파악하는 도구로 활용될 수 있으며, 이를 통해 IMO 회원국들의 협약이행 수준 개선에 기여할 수 있을 것으로 기대한다.

Prior to the investigations on fuel degradation it is necessary to describe the reference characteristics of the spent fuel. It establishes the initial condition of the reference fuel bundle at the start of dry storage. In a few technology areas, CANDU fuels have not yet developed comprehensive analysis tools anywhere near the levels in the LWR industry. This requires significantly improved computer codes for CANDU fuel design. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, clad stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the in-house code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

In December 26, 2017, the Ministry of Science and ICT’s Cyber Infringement Response Division, the KCC’s Internet Ethics Team, and the National Police Agency’s Cyber Investigation Division announced the “Comprehensive IP Camera Measures” to prevent damage to users by strengthening IP camera security. The background was a countermeasure against the spread of public anxiety such as leakage of privacy and threatening national security as cases of illegal filming and distribution of videos by accessing IP cameras without permission at that time occurred. At that time, the measures consisted of three major strategies: institutionalization of products with security (manufacturing and import stage), rapid action and response to hacking threats in advance, and intelligent advancement of IP cameras (industrial development) and fostering various video and safety industries. This paper deals with the development of installation standards and evaluation indicators for CCTV systems installed in the relevant national security facility. When designing a CCTV system, the resolution and sensitivity of the camera, the angle of view of the lens, and the aperture should be selected in consideration of the length and width of the monitoring area. When it is necessary to determine an alarm based on the recorded image information, the resolution of the recording device should also be considered. If there is a restriction on the surveillance area of the camera due to topographic features, visitors, etc., additional cameras should be considered. Additional surveillance cameras should be considered to effectively monitor sections where fences are bent or sections with severe slopes. When designing a CCTV system, supplementary means that can be used for intrusion monitoring should be devised even if the entire or part of the system causes a failure. KISA’s performance test and certification procedures shall apply mutatis mutandis to intelligent CCTV image analysis test procedures and certification. Intelligent CCTV image analysis evaluates the performance of the equipment by comparing files of ground truth (GT) and system alarm (SA) for verification to calculate the evaluation score.

보행환경은 개인의 영역이자 공공 공간으로서 시민들의 일상생활에 매우 중요한 요소이다. 보행환경의 중요성이 인지되면서 국가적 차원에서도 보행환경 실태조사를 전국 지자체가 5년마다 시행하도록 법으로 규정하는 등 체계적인 실태조사가 필요한 실정이다. 하지만 보행환경에 대한 실태조사는 일부 지역을 대상으로 현장 조사에 의지하는 등 실태조사 방법론에 있어서는 기존의 한계를 벗어나지 못하고 있다. 본 연구는 고해상도 거리 영상과 딥러닝 기술을 활용한 보행환경 평가 지표 개발을 목표로 하였다. 보행환경 평가 지표 개발을 위해 보행환경 평가와 관련된 국내외 문헌 및 딥러닝 기술을 활용한 보행환경 평가 연구를 리뷰를 토대로 보행환경 평가 지표 초안을 개발하고, 도출된 보행환경 평가 지표의 구체적 데이터 구축 가능성을 확인하기 위해 거리 영상의 시멘틱 세그멘테이션(semantic segmentation) 결과 정확도와 영상 외 필요한 자료에 대한 취득 가능성을 검토한 후 최종 보행환경 평가 지표를 제안하였다. 도출된 보행환경 평가 지표는 안전성, 편리성, 쾌적성, 접근성 4개 카테고리에 8개 지표를 활용하는 것을 제안하였다. 본 연구의 결과는 현장 관찰 조사나 설문조사에 기반한 기존 보행환경 연구의 한계점을 탈피하고 고해상도 거리 영상과 딥러닝 기술을 활용한 도시 연구의 지능화 계기를 마련하고 보행환경 평가 업무를 보다 효율적으로 수행할 수 있는 초석이 될 것으로 기대한다.

In this study, a study was conducted to improve the reliability of the valve by developing a valve leakage and reliability measurement system designed to secure the high quality and reliability of the butterfly valve. The system measuring the torque required for valve opening and closing operation, and was configured to operate after multiple opening and closing of the valve to check for leakage of the valve. Finally, a system that can perform efficient work in terms of productivity was developed by enabling leak inspection, torque measurement, and reliability inspection through one integrated system.