In this paper, we propose a analysis of domestic and foreign standards on protecting human body from EMF(Electromagnetic filed) and effect of the human body through EMF analysis of military electronic equipment in operation. We measured the EMF intensity of the side lobe and back lobe main beam on Radar Equipment at an accessible position. As a result, it was confirmed that the measured EMF intensity of three types of radar equipment did not exceed the standard on domestic protecting human body from EMF.

본 연구는 소재·부품·장비의 한·중 산업간 무역구조 및 무역 경쟁력의 변화를 분 석하기 위하여 수행되었다. 먼저 수출입 추세를 보았을 때 소재·부품·장비의 대중국 무역은 지난 10여 년간 지속적으로 무역수지 흑자를 달성하고 있지만 흑자 규모는 지속적으로 하락하는 추세에 있는 것으로 나타났다. 그리고 GL 지수를 보았을 때 산 업 내 무역의 수준이 지속적으로 증가하고 동종 산업 간의 경쟁 또한 치열해지고 있 는 것으로 파악되었다. 10년간 전체 소재·부품·장비 산업의 TSI 지수는 양(+)으로 나 타나 우리나라 산업의 경쟁력이 우세한 것으로 나타났지만 TSI 지수가 지속적으로 줄어들어 산업의 경쟁력이 점차 감소하고 있는 것으로 파악되었다. 중국의 급격한 성장에 대응하기 위해서는 우리 제품의 자급률 향상 및 경쟁력 강화 등을 위한 노력 을 지속적으로 기울여야 할 것으로 판단이 된다.

In the case of a die-casting process, defects that are difficult to confirm by visual inspection, such as shrinkage bubbles, may occur due to an error in maintaining a vacuum state. Since these casting defects are discovered during post-processing operations such as heat treatment or finishing work, they cannot be taken in advance at the casting time, which can cause a large number of defects. In this study, we propose an approach that can predict the occurrence of casting defects by defect type using machine learning technology based on casting parameter data collected from equipment in the die casting process in real time. Die-casting parameter data can basically be collected through the casting equipment controller. In order to perform classification analysis for predicting defects by defect type, labeling of casting parameters must be performed. In this study, first, the defective data set is separated by performing the primary clustering based on the total defect rate obtained during the post-processing. Second, the secondary cluster analysis is performed using the defect rate by type for the separated defect data set, and the labeling task is performed by defect type using the cluster analysis result. Finally, a classification learning model is created by collecting the entire labeled data set, and a real-time monitoring system for defect prediction using LabView and Python was implemented. When a defect is predicted, notification is performed so that the operator can cope with it, such as displaying on the monitoring screen and alarm notification.

Currently, agriculture in Korea is facing a manpower shortage due to a decrease in the population and an aging workforce. Involving disabled people in agriculture as a solution to the problem of manpower shortage in agriculture can help both farmers and people with disabilities. For this, it is necessary to develop agricultural tools that disabled people can use in agricultural activities. This study tried to design agricultural work equipment for the disabled by applying a design technique using a digital human body model. To this end, in this study, a human body model for the disabled was created using anthropometric data for the disabled. After that, the agricultural work simulation using human body models was used to identify the problems in agricultural work of the disabled, and design supplementation was performed to solve the problems. As a result, design dimensions such as the height of the farm work bench for the disabled, the length of the fruit harvester, the length of the sprayer, and the height of the starter switch were determined. This study has significance in that the digital human body model-based design method was applied to the design of agricultural work equipment for the disabled.

3D imaging equipment is essential for automated robotic operations that cut radiologically contaminated structure and transfer segmented pieces in nuclear facility dismantling site. Automated dismantling operations using programmed robotic arms can make conventional nuclear facility dismantling operations much more efficient and safer, so dismantling technologies using robotic arms are being actively researched. Resolving the position uncertainty of the target structure is very important in automated robot work, and in general industries, the problem of position uncertainty is solved through the method of teaching the robot in the field, but at the nuclear facility dismantling site, the teaching method by workers is impossible due to activated target structures. Therefore, 3D imaging equipment is a key technology for a remote dismantling system using automated robotic arms at nuclear facility dismantling site where teaching methods are impossible. 3D imaging equipment available in radioactive and underwater environments is required to be developed for a remote dismantling system using robotic arms because most commercial 3D scanners are available in air and certain 3D scanners available in radioactive and underwater environments cannot satisfy requirements of the remote dismantling system such as measurement range and radiation resistance performance. The 3D imaging equipment in this study is developed based on an industrial 3D scanner available in air for efficient development. To protect the industrial 3D scanner against water and radiation, a housing is designed by using mirrors, windows and shieldings. To correct measurement errors caused by refraction, refraction model for the developed 3D imaging equipment is defined and parameter studies for uncertain variables are performed. The 3D imaging equipment based on the industrial 3D scanner has been successfully developed to satisfy the requirements of the remote dismantling system. The 3D imaging equipment can survive up to a cumulative dose of 1 kGy and can measure a 3D point cloud in the air and in water with an error of less than 1 mm. To achieve the requirements, a proper industrial 3D scanner is selected, a housing and shielding for water and radiation protection is designed, refraction correction are performed. The developed 3D imaging equipment is expected to contribute to the wider application of automated robotic operations in radioactive or underwater environments.

A large amount of concrete radioactive waste is generated during the decommissioning of nuclear facilities, including nuclear power plants, and it is expected that the radioactive waste management expenses will be huge. In order to reduce the concrete radioactive waste, a decontamination or removal process is required, but working on concrete may present a risk of worker exposure in a high-radioactive space. Therefore, in this study, a remote control integrated decontamination equipment that can reduce concrete radioactive waste and ensure the safety of workers during the concrete decontamination process in a high-radioactive space was developed. The integrated decontamination equipment consists of remote movement, automatic surface contamination measurement, automatic surface decontamination and debris/dust removal systems. The remote movement system generates ‘mapping data’ of topographic features for the work space and ‘location data’ that coordinates the location of the integrated decontamination equipment through LiDAR (Light Detection and Ranging) sensor and SLAM (Simultaneous Localization And Mapping) technique. The user can check the location of the integrated decontamination equipment through ‘location data’ outside the work space, and can move it by remote control through wired/wireless communication. The automatic surface contamination measurement system uses a radiation detector and an automatic measurement algorithm to generate ‘surface measurement data’ based on the measurement distance interval and measurement time set by the user. ‘Surface measurement data’ is combined with ‘location data’ to create a visualized map of radioactive contamination, and users can intuitively realize the location and degree of contamination based on the map. The automatic surface decontamination system uses a laser and an automatic removal algorithm to decontaminate the concrete surface. Concrete debris and dust generated during this process were collected by the debris/dust removal system, minimizing waste generation and radiation exposure due to secondary pollution. The integrated decontamination equipment developed through this study was applied with technologies that reduced radioactive concrete waste and ensured the safety of workers. If technology verification and on-site applicability review are performed using concrete specimens simulating nuclear power plant or similar environments, it is reasoned to contribute to the domestic and overseas decommissioning industry.

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.

IAEA has the right and obligation to verify the states’ commitments for safeguards under the comprehensive safeguards agreement and additional protocols. There are IAEA inspections such as PIV, DIV, RII, SNRI under these agreements. As part of the implementation of this mission, the IAEA inspectors perform the verification for the state’s accounting reports related the nuclear materials such as ICR, PIL, MBR. To do well this verification, the inspectors often use non-destructive analysis, which aims to measure attributes of the items during the inspections. This kind of an activity aims to detect the missing nuclear items or wrong things in the facility using nuclear materials. In general, NDA techniques use the neutron counting and gamma ray spectrometry. Besides, IAEA also performs several verification measures as follows. - C/S (Containment and Surveillance techniques) is to maintain the continuity of the knowledge by giving assurance that its containment remains unimpaired. - Unattended and remote monitoring is to transmit the data from onsite of the facility through the on-line system. - E/S (Environmental Sampling) is to detect the minute traces of nuclear materials by smearing some points in the on-site of the facility. Nowadays, the above mentioned techniques are important ways to increase the effectiveness of the safeguards approaches reducing IAEA actual costs. To strengthen the effectiveness and improve the efficiency of safeguards approaches, IAEA always develops and adopts the techniques and equipment for safeguards. Especially, IAEA seems to be concerned with the improvement and development of the non-destructive techniques and equipment in the fields of nuclear fuel cycle. IAEA develops the new techniques and equipment through the help of MSSP (Member States Support Programs). The IAEA defines the needs of safeguards and coordinates the support programs. After the IAEA tests and evaluates the techniques/equipment developed, IAEA decides whether to use the developed techniques and equipment during the inspection by the procedure of the IAEA quality assurance. This paper aims at studying the current changes of the IAEA equipment such as DCVD, NGSS and HCES.

The demand for chiller equipment that keeps each machine at a constant temperature to maintain the best possible condition is growing rapidly. PID (Proportional Integral Derivation) control is a popular temperature control method. The error, which is the difference between the desired target value and the current system output value, is calculated and used as an input to the system using a proportional, integrator, and differentiator. Through such a closed-loop configuration, a desired final output value of the system can be reached using only the target value and the feedback signal. Furthermore, various temperature control methods have been devised as the control performance of various high-performance equipment becomes important. Therefore, it is necessary to design for accurate data-driven temperature control for chiller equipment. In this research, support vector regression is applied to the classic PID control for accurate temperature control. Simulated annealing is applied to find optimal PID parameters. The results of the proposed control method show fast and effective control performance for chiller equipment.

In this study is a study to measure the section by voltage through the high integration of the circuit of the inspection equipment for the power supply circuit of semiconductor equipment. The experiment was conducted by increasing the -1.5∼4 voltage section by 0.5V. At this time, the tolerance was applied to ±0.1%+5mA. Although the voltage increased through the experiment, the accuracy of the measurement data did not change, and it was confirmed through this experiment that the null hypothesis(H0) was adopted in each section through the hypothesis test.

Spent filters with a high radiation dose rate of 2 mSv·hr−1 or more are not easily managed. So far, the Korean policy for spent filter disposal is to store them temporarily at nuclear power plants until the waste filters can be easily managed. Nuclear power plant decommissioning in Korea is starting with Kori unit 1. Volume reduction of waste generated during decommissioning can reduce the cost and optimize the space usage at disposal site. Therefore, efficient volume reduction is a very important factor during the decommissioning process. A conceptual method, based on the experiences of developing 200 and 800 ton compactors at Orion EnC, has been developed considering worker exposure with the followings a crusher (upgrade of compaction efficiency), an automatic dose measuring system with a NaI(Tl) detector, a shield box, an inner drum to prepare for easy handling of drums and packaging, a 30 ton compactor, and an automatic robot system. This system achieves a volume reduction ratio of up to 85.7%; hence, the system can reduce the disposal cost and waste volume. It can be applied to other types of wastes that are not easily managed due to high dose rates and remote control operation necessity.

대형액화천연가스(Liquefied Natural Gas, LNG)선이 연안 터미널에 정박할 경우 바람과 조류 등의 환경하중에 대응하여 안전을 확보할 수 있는 계류 안전을 위한 케이블 계류력 산정이 필요하다. 이에 기존의 주요 계류역(Mooring Force) 계산방법의 비교 및 분석을 수 행하였다. 비교 및 분석을 통해 석유회사국제해운포럼(Oil Companies International Marine Forum, OCIMF)의 계류설비지침에서 권고하는 계산 방법을 선정하였으며 이를 기반으로 본 논문에서는 실제 대형 LNG선에 적용하여 OCIMF 계류설비지침의 스펙트럼을 이용한 계류줄의 계 류력 계산 사례를 제시하였다. OCIMF 계류설비지침에 따른 스펙트럼으로 계산한 계류력은 환경 외력과 풍동 시험으로 계산한 바람저항 계수 기반 선박 환경 외력과 최대값에서 매우 유사한 결과값을 주는 것을 확인할 수 있었다. OCIMF 계류설비지침에 따른 스펙트럼으로 계산한 계류력에 대한 검증으로 전문 계류력 계산 소프트웨어인 OPTIMOOR 소프트웨어를 사용하여 결과를 비교하였으며 둘의 결과는 매 우 유사한 것을 확인하였다. OPTIMOOR를 사용할 경우에는 각각의 케이블의 인장력을 정밀하게 계산할 수 있어 경제적인 제약이 없을 때 적극적 사용이 추천된다. 결론적으로 OCIMF 계류설비지침에 따른 스펙트럼으로 계산한 계류력이 대형 LNG선의 계류력 계산에 적용함에 문제가 없음을 실제 계산 사례를 통해 검증할 수 있었다.

The background of the development is to contribute to the reduction of radioactive waste, recycling of resources and effectively purifying the air in the workplace. Ultimately, it affects the reduction of internal exposure of workers. According to the standard procedure of KHNP,「Use and Management of Respiratory Protection Equipment」, the expiration date of mask filter is indicated by the manufacturer before opening. It is 1 year from the date of first combination after opening. We have developed an air purifying equipment that can recycle and reuse expired mask filter waste in nuclear power plant. In order to confirm the performance, we observed air pollution level by operation time. The location was measured at 3 locations including the decontamination product warehouse in NPP, and the size of the measurement locations were 53 m3, 150 m3, 180 m3, and 900 m3. As a result of measurement, significant air purification effect was found in 53 m3 and 150 m3. Decontamination effect of 80% was shown after 1 hour of operation, and 20% of decontamination effect was shown gently for 3 hours thereafter. On the other hand, there was no significant decontamination effect in the 180 m3 and 900 m3 spaces. Significant results were derived by statistical methods. Statistical procedure involves the collection of data leading to test of the relationship between two statistical data sets, or a data set and synthetic data drawn from an idealized model. The basic composition and product characteristics was as follows: Blower, filter fixing unit, Air purifier outlet round shape, Differential pressure gauge, inverter (200 V, 3π, 200 W). The developed product weigh is 25 kg. This is lighter than the existing product weighing 100 kg. It is judged that it is suitable for convenient use. Because the area where the major air pollution level occurs is isolated by a room in NPP. This developed product has a greater significance in that it recycles radioactive waste within the radiation management area rather than air purification efficiency.