• 전기학회논문지
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• 제65권12호
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• pp.2030-2036
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• 2016

This paper presents a hardware development procedure of the ASB(Active Seat Belt) control system to comply with ISO 26262. The ASIL(Automotive Safety Integrity Level) of an ASB system is determined through the HARA(Hazard Analysis and Risk Assessment) and the safety mechanism is applied to meet the reqired ASIL. The hardware architecture of the controller consists of a microcontroller, H-bridge circuits, passive components, and current sensors which are used for the input comparison. The required ASIL for the control systems is shown to be satisfied with the safety mechanism by calculation of the SPFM(Single Point Fault Metric) and the LFM(Latent Fault Metric) for the design circuits.

• 비파괴검사학회지
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• 제20권4호
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• pp.329-341
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• 2000

• 한국정보전자통신기술학회논문지
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• 제14권6호
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• pp.452-460
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• 2021

전 세계적으로 전기차 시장이 확대됨에 따라 성능 및 안전성의 문제를 보완한 친환경적인 전기차가 계속 출시되고 시장이 더욱 커지고 있다. 하지만 전기차의 경우 충전의 불편함, 감전과 같은 안전 문제, 여러 전장부품들의 연동으로 인한 EMI(Electromagnetic interference) 문제는 전기차에서 해결해야 하는 문제이다. 무선전력전송 기술을 이용하면 전기차 충전에 대한 불편함 해소와 고전류, 고전압을 직접 다루지 않아 안전성의 문제를 해결할 수 있으나 EMI 저감을 위한 설계가 이루어지지 않는다면 오작동을 일으켜 더 큰 문제를 일으킬 수 있다. 본 논문은 전기차 무선전력전송 핵심 전장 부품인 무선충전컨트롤모듈에서 발생할 수 있는 EMI를 저감시키기 위한 전원무결성과 신호무결성을 갖는 전기차 무선전력전송 무선충전컨트롤모듈 EMI 저감 설계하였다. 전원부분에서 발생할 수 있는 공진, 임피던스 등의 문제와 신호 부분에서 발생할 수 있는 고속통신간의 신호왜곡의 문제를 시뮬레이션을 통해 EMI 저감 설계하였다.따라서 전원무결성과 신호무결성을 갖는 EMI 저감 설계를 통해 전기차 무선전력전송 무선충전컨트롤모듈 800 MHz ~ 1 GHz 대역과 1.5 GHz에서 각각 10 dBu V/m, 15 dBu V/m이 저감되는 것을 확인하였다.

• 한국항해항만학회지
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• 제37권1호
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• pp.41-48
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• 2013

개선된 정확도 성능을 확보하기 위하여 보강 시스템을 이용한 많은 연구가 진행되고 있다. 네트워크 RTK는 다중 기준국의 반송파 측정치 보정정보를 이용하여 시공간 오차를 보강한 측위성능을 얻기 위한 기법으로 현재에도 꾸준히 연구되고 있다. 그러나 성능개선을 목적으로 한 알고리즘 개선안에 대한 연구는 지속적으로 연구되었지만, 무결성 확보를 위한 연구는 아직 미비하다. 본 논문에서는 네트워크 RTK에서의 무결성 확보를 위한 기초연구로 위성이상이 발생한 경우에 이상을 검출하고 이상 위성을 식별할 수 있는 알고리즘을 제안하였다. 그리고 GSS7700 모델의 상용 시뮬레이터를 사용하여 오차 시나리오가 인가된 위성 신호를 생성하고, DL-V3 모델의 이중주파수용 상용 수신기를 사용하여 수신한 데이터를 사용하여 제안한 알고리즘의 이상 검출 성능을 검증하였다.

• 한국항해항만학회지
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• 제39권6호
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• pp.479-484
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• 2015

네트워크 RTK는 네트워크 내 다중 기준국의 반송파 측정치 보정정보를 활용하는 정밀 측위 기법으로 성능 개선을 목적으로 꾸준히 연구가 진행되어 왔다. 최근까지는 주로 측지 측량 분야에서 사용하였기 때문에 정확도 개선을 위한 연구 위주로 진행되었으며, 무결성 확보를 위한 연구는 아직 미비하다. 본 논문에서는 네트워크 RTK에서의 무결성 확보를 위한 기초연구로 네트워크 RTK 환경에 적합한 감시 시스템을 설계하였다. 이를 위하여 네트워크 RTK에서의 무결성 결함 조건을 도출하고, 각 결함 조건 별로 활용할 수 있는 이상 검출 및 식별 기법을 소개하였으며, 이를 기반으로 네트워크 RTK를 서비스하는 중앙처리국에서 활용할 수 있는 감시 시스템을 설계하였다.

• 한국안전학회지
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• 제15권1호
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• pp.28-35
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• 2000

As huge energy transfer systems like a nuclear power plant, steam power plant and petrochemical plant are operated for a long time, mechanical properties are changed by degradation. The life time of the systems can be affected by the mechanical properties. So determination of the integrity of the metallic structure is required either to en4sure that failure will not occur during the service life of the components or to evaluate the lifetime extension of the structure. An automated ball indentation(ABI) method was developed as a non-destructive technique for evaluating the integrity of such metallic components. In this paper, we would like to present the aging evaluation technique by the ABI method. The five classes of the thermally aged CF8M specimens were prepared using an artificially accelerated aging method. After holding 100, 300, 900, 1800 and 3600 hours at 43$0^{\circ}C$, the specimens were cooled down using water to room temperature respectively. The tensile test, hardness test, charpy impact test, ABI test were performed. The results of the fracture test were compared with those of ABI test and the new evaluation technique of the integrity of metallic structures was developed.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2001년도 춘계학술발표대회 개요집
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• pp.228-231
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• 2001

Elastic-plastic fracture toughness $J_{1c}$ can be used as an effective design criterion in elastic plastic fracture mechanics. Most of these systems are operated at high temperature and $J_{1c}$ values are affected by temperature. therefore, the $J_{1c}$ valuse at high temperature must be determined for use of integrity evaluation and designing of such systems. Elastic-plastic fracture toughness $J_{1c}$ tests were performed on SA516 carbon steel plate and test results were analyzed according to ASTM E 813-8, ASTM 1813-89. Safety and integrity are required for reactor pressure vessels vecause pthey are operated in high temperature. there are single specimen method, which used as evaluation of safety and integrity for reactor pressure vessels. In this study, elastic-plastic fracture toughness$(J_{1c})$ and $J-\Delta{a}$ of SA 516/70 steel used as reactor pressure vessel steel are measured and evaluated at room Temperature, $150^{\circ}C$, $250^{\circ}C$ and $370^{\circ}C$ according to unloading compliance method.

• 한국해양공학회지
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• 제15권2호
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• pp.99-104
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• 2001

Elastic-plastic fracture toughness $J_{lc}$ can be used as an effective design criterion in elastic plastic fracture mechanics. Most of these systems are$J_{lc}$ $J_{lc}$ value at high temperature must be determined for use of integrity evaluation and designing of such systems. Elastic-plastic fracture toughness $J_{lc}$ tests were performed on SA516/70 carbon steel plate and test results were analyzed according to ASTM E 813-87, ASTM E 813-89 and ASTM E 1152-87.safety and integrity are required for reactor pressure vessels because, they are operated in high temperature. There are single specimen method, which used as evaluation of safety and integrity for reactor pressure vessels. In this study, elastic-plastic fracture toughness($J_{lc}$) and J-$\Delta$a of SA 516/70 steel used as reactor pressure vessel steel are measured and evaluated at room temperature, 150$^{\circ}C $, 250$^{\circ}C $ and 370$^{\circ}C $ according to unloading compliance method.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.466-473
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• 2021

Fire protection is one of important issues to ensure safety and reduce risks of nuclear power plants (NPPs). While robust programs to shut down commercial reactors in any fires have been successfully maintained, the concept and associated regulatory requirements are constantly changing or strengthening by lessons learned from operating experiences and information all over the world. As part of this context, it is necessary not only to establish specific fire hazard assessment methods reflecting the characteristics of research reactors and educational reactors but also to make decisions based on advancement encompassing numerical analyses and experiments. The objectives of this study are to address fire simulation in the control room of an educational reactor and to discuss integrity of digital console in charge of main operation as well as analysis results through comparison. Three electrical fire scenarios were postulated and twenty-four thermal analyses were carried out taking into account two turbulence models, two cable materials and two ventilation conditions. Twelve supplementary thermal analyses and six subsequent structural analyses were also conducted for further examination on the temperature and heat flux of cable and von Mises stress of digital console, respectively. As consequences, effects of each parameter were quantified in detail and future applicability was briefly discussed. On the whole, higher profiles were obtained when Deardorff turbulence model was employed or polyvinyl chloride material and larger ventilation condition were considered. All the maximum values considered in this study met the allowable criteria so that safety action seems available by sustained integrity of the cable linked to digital console within operators' reaction time of 300 s.

• 대한기계학회논문집A
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• 제27권6호
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• pp.930-937
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• 2003

The extension of design life among power plants is increasingly becoming a world-wide trend. Kori #1 unit in Korea is operating two cycle. It has two man-ways for tube inspection in a steam generator which is one of the important components in a nuclear power plant. Especially, stud bolts fur man-way cover have damaged by disassembly and assembly several times and degradation for bolt materials for long term operation. It should be evaluated and compared by ASME Code criteria for integrity evaluation. Integrity evaluation criteria which has been made by the manufacturer is not applied on the stud bolts of nuclear pressure vessels directly because it is controlled by the yield stress of ASME Code. It can apply evaluation criteria through FEM analysis to damaged female threads and to evaluated safety fer helical-coil method which is used according to Code Case-N-496-1. From analysis results, we found .that it is the same results between stress intensity which got from FEM analysis on damaged female threads over 10％ by manufacture integrity criteria and 2／3 yield strength criteria on ASME Code. It was also confirmed that the helical-coil repair method would be safe.