• 한국가스학회지
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• 제21권2호
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• pp.26-31
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• 2017

이 논문은 디젤 자동차의 터보차저 시스템에 관련된 고장사례를 연구하기 위한 것이목적이다. 첫 번째 사례는 터보차저를 분해하여 확인한 결과 터보 휠의 파손으로 터빈이 손상되어 작동되지 않음으로 배기관 구멍이 막혀 가속이 되지 않는 것을 확인하였다. 터보차저 장착차량의 경우 높은 회전수에서는 엔진을 바로 정지시키게 되면 고온에 의한 터보의 고착현상이 발생될 수 있으므로 공회전 상태를 충분히 유지한 다음 시동을 끄도록 한다. 두 번째 사례는 터보차저를 작동시키는 VGT 액튜에이터의 로드가 고착되어 가속불량 현상이 발생된 것으로 확인되었다. 세 번째 사례는 알터네이터(Alternator) 저항과 인터쿨로 팬 릴레이 커넥터 오조립으로 인해 주행 중 출력부족 현상이 발생한 것을 확인하였다. 따라서, 터보차저 시스템의 철저한 관리를 통해 고장이 발생하지 않도록 하여야 한다.

• 조명전기설비학회논문지
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• 제14권1호
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• pp.67-74
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• 2000

본 논문에서는 GIS내 파티클에 의해 발생할 수 있는 절연사고를 미연에 방지하고 또한 파티클의 상태를 파 악하기 위하여 GIS 41 파티클 상태에 따른 부분방전 신호를 측정하고 분석하였다. 본 논문에서는 GIS내 파티콜 의 상태를 4가지(전극부착시, 외함부착시, 스페이서 부착시, 자유운동시)로 모의하였으며, 이때 발생하는 부분방전 신호의 $\Phi$-Q-N 분포를 형태적으로 분석하였다 또한, $\Phi$-Q 분포에 대하여 통계적 연산자를 도입하여 파티 클 상태에 따른 부분방전 신호를 정량적으로 분석하였다. 그 결과 파티클 상태에 따른 $\Phi$-Q-N 분포의 형태적인 분석과 $\Phi$-Q 분포의 정량적인 분석으로 파티콜의 상태 구별이 가능하였다.

• 소음진동
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• 제5권3호
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• pp.361-371
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• 1995

The vibration of reactor internals should be monitored and diagnosed for the early detection of the failure of reactor pressure vessel. This can be performed by analyzing the time-history signals from the excore neutron flux detertors. The conventional method is an on-demand system which generates power spectra through Fast Fourier Transform(FFT) algorithm. The operator can make his own decision to detect abnormal vibration using these spectra. This post- processing method, however, requires special expertise in the reactor noise analysis and signal processing for random data. It may mislead the operator into erroneous decision-making, if he is a novice in reactor noise analysis. Hence this study is focused on the automated monitoring and diagnosis procedure for the reactor noise analysis, especially on the Fuzzy algorithm to recognize the pattern of the vibration of Core Suport Barrel. The excore neutron signals of Yonggwang Nuclear Power Plant unit 3 is acquired and analyzed using conventional FFT spectra and tested to adopt the Fuzzy method. An Automated Monitoring and Diagnosis System for CSB Vibration using this Fuzzy method is proposed. Furthermore, vibration data for CSB of Youggwang Nnclear Power Plant unit 3 is presented.

• 항공우주기술
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• 제11권2호
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• pp.129-139
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• 2012

천리안위성의 안정적인 장기 일반운영과 예기치 못한 비상상황을 대비하기 위한 관제원 운영능력 유지를 위해 개발된 관제모의훈련시스템의 구조와 특징에 대해 설명한다. 본 논문에서 특히 교관 편의 기능을 강조한 GUI 환경의 고장주입 기능과 훈련 시나리오를 자동으로 동작시키는 스케줄러, 기존 시뮬레이터의 고장발생 기능개선 부분을 중점적으로 기술하여 훈련시스템의 차별적인 장점과 기대효과를 조명한다. 아울러 고장 복구 훈련 예제를 통하여 훈련시스템 동작과정의 이해를 돕는다.

• 대한안전경영과학회지
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• 제22권1호
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• pp.15-21
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• 2020

Industrial Motors diagnostic equipment is highly dependent on the automation system, so if there are defects in the automation equipment, it can only rely on the operator's intuitive judgment.To help with intuitive judgment, Park's Vactor Approach(PVA) represents the current signal as a pattern of circles, so it can tell if a fault occurs when the circle is distorted. However, the failure to judge the degree of distortion of the circle pattern is the basis of the fault, so it will face difficulties. In this paper, in order to compare the faults of PVA, the period of d-axis current of PVA pulsation was mastered, so that two phase differences occurred in the same signal source. Through experiments, it is confirmed that this is a 90 degree cross formation of PVA, which is convenient for judging from the vision that there is no fault, thus helping the operator to make intuitive judgment.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2008년도 하계학술발표대회 논문집
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• pp.30-35
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• 2008

This paper describes a fault detection and diagnosis (FDD) system developed for the heat source apparatus in building air-conditioning system. As HVAC&R systems in building become complex and instrumented with highly automated controllers, the processes and systems get more difficult for the operator to understand and detect the mal-functions. Poorly maintained, degraded, and improperly controlled equipment wastes an estimated 15% to 30% of energy used in commercial building. When operating a complex facility, FDD system is beneficial in equipment management to provide the operator with tools which can help in decision making for recovery from a failure of the system. Automated FDD for HVAC&R system has the potential to reduce energy and maintenance costs and improves comfort and reliability. Over the last decade there has been considerable research for developing FDD system for HVAC&R equipment. However, they are being made too much of a theoretical study, so only a small of FDD methods are deployed in the field. This study deduced an actual defect source for the heat source apparatus and suggested a low price FDD method which is ready to be deployed in the field.

• 한국가스학회지
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• 제22권6호
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• pp.8-15
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• 2018

플랜트 내 위험지역의 안전을 위해 작업자 중심의 안전관리가 필요하다. 최근 5년간 가스 사고의 원인은 시설 노후 및 장비고장 뿐만 아니라, 사용자의 취급부주의나 고의사고, 공급자 취급부주의 등 작업자의 행동에 밀접한 관련이 있다. 이와 같은 사고를 미연에 방지하기 위해서, 플랜트 내 위험지역에 대한 실시간 모니터링이 필요로 하다. 하지만 실시간 모니터링을 위해서 작업(근로)공간에 카메라 설치 시, 인권침해와 같은 문제가 발생한다. 이를 방지하기 위해서 작업자의 신원 노출이 적은 저해상도의 Infrared 카메라를 이용한다. 또한 실시간 모니터링 시, 사람이 아닌 CNN알고리즘을 이용하여 이미지 분석을 통하여 인권침해 문제를 예방한다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.866-872
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• 2024

Design Extension Conditions (DEC)-A assessments of the operating nuclear power plants are generally considered for the purpose of getting additional safety demonstrations of their capability to undergo conditions that are generally more severe than DBAs by features implemented in the design and accident management measures. The pursued methodology is generally based upon Best Estimate approaches aiming at verifying that the safety limits in terms of integrity of the barriers against eventual large or early releases of radioactive material are fulfilled. These aspects are nowadays being experimentally and analytically addressed within the OECD/NEA experimental projects like the ATLAS and PKL series where a set of DEC-A experiments are considered. In this paper, experiments related to SBLOCA at the vessel upper head of the pressurized vessel of ATLAS and PKL are analytically assessed using the CATHARE code. These experiments includes issues related to common cause failure of the safety injection system and operator actions for preventing core excessive overheating. It is shown that, on the one hand, the safety features embedded in the design together with the operator actions are capable to prevent the progression towards a severe accident state and on the other hand, the code prediction capabilities for such scenario are generally good but still to be enhanced.

• Nuclear Engineering and Technology
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• 제30권5호
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• pp.403-415
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• 1998

The CANDU 6 reactor assembly consists of an array of 380 pressure tubes, which are installed horizontally in a large cylindrical vessel, the Calandria, containing the low pressure heavy water moderator. The pressure tube is located inside the calandria tube and the annulus between these tubes, which forms a closed loop with $CO_2$ gas recirculating, is called the Annulus Gas System(AGS). It is designed to give an alarm to the operator even for a small pressure tube leak by a very sensitive dew point meter so that he can take a preventive action for the pressure tube rupture incident. To judge whether the operator action time is enough or not in the design of Wolsong 2,3 & 4, the Leak Before Break(LBB) assessment is required for the analysis of the pressure tube failure accident. In order to provide the required data for the LBB assessment of Wolsong Units 2, 3, 4, a series of leak detection capability tests was performed by injecting controlled rates of heavy water vapour. The data of increased dew point and rates of rise were measured to determine the alarm set point for the dew point rate of rise of Wolsong Unit 2. It was found that the response of the dew point depends on the moisture injection rate, $CO_2$ gas flow rate and the leak location. The test showed that CANDU 6 AGS can detect the very small leaks less than few g/hr and dew point rate of rise alarm can be the most reliable alarm signal to warn the operator. Considering the present results, the first response time of dew point to the AGS $CO_2$ flow rate is approximated.

• 한국안전학회지
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• 제35권5호
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• pp.121-127
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• 2020

Since the Fukushima nuclear accident in 2011, the public were concerned about the safety of Nuclear Power Plants (NPPs) in extreme natural disaster situations, such as earthquakes, flooding, heavy rain and tsunami, have been increasing around the world. Accordingly, the Stress Test was conducted in Europe, Japan, Russia, and other countries by reassessing the safety and response capabilities of NPPs in extreme natural disaster situations that exceed the design basis. The extreme natural disaster can put the NPPs in beyond-design-basis conditions such as the loss of the power system and the ultimate heat sink. The behaviors and capabilities of NPPs with losing their essential safety functions should be measured to find and supplement weak areas in hardware, procedures and coping strategies. The Loss of Ultimate Heat Sink (LUHS) accident assumes impairment of the essential service water system accompanying the failure of the component cooling water system. In such conditions, residual heat removal and cooling of safety-relevant components are not possible for a long period of time. It is therefore very important to establish coping strategies considering all available equipment to mitigate the consequence of the LUHS accident and keep the NPPs safe. In this study, thermal hydraulic behavior of the LUHS event was analyzed using RELAP5/Mod3.3 code. We also performed the sensitivity analysis to identify the effects of the operator recovery actions and operation strategy for charging pumps on the results of the LUHS accident.