• 해양환경안전학회지
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• 제28권1호
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• pp.184-192
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• 2022

밸브의 내부 누설 현상은 밸브의 내부 부품의 손상에 의해 발생하며 배관 시스템의 사고와 운전정지를 일으키는 주요 요인이다. 본 연구는 버터플라이형 밸브의 내부 누설에 따라 배관계에서 발생하는 음향방출 신호를 이용하여 배관 가동 중 실시간 누설 진단의 가능성을 검토하였다. 이를 위해 밸브의 작동 모드별로 측정한 시간영역의 AE 원시신호를 취득하였으며 이로부터 구축한 데이터셋은 데이터 기반의 인공지능 알고리즘에 적용하여 밸브의 내부 누설 유무를 진단하는 모델을 생성하였다. 누설 유무진단을 분류의 문제로 정의하여 SVM 기반의 머신러닝과 CNN 기반의 딥러닝 분류 알고리즘을 적용하였다. 데이터의 특징 추출에 기반한 SVM 분류 모델의 경우, 이진분류 모델에서 구축된 모델에 따라 83~90%의 정확도를 나타냈으며, 다중 클래스인 경우 분류 정확도가 66%로 감소하였다. 반면, CNN 기반의 다중 클래스 분류 모델의 경우 99.85%의 분류 정확도를 얻을 수 있었다. 결론적으로 밸브 내부 누설 진단을 위한 SVM 분류모델은 다중 클래스의 정확도 향상을 위해 적절한 특징 추출이 필요하며, CNN 기반의 분류모델은 프로세서의 성능 저하만 없다면 누설진단과 밸브 개도 분류에 효율적인 접근방법임을 확인하였다.

• 한국기계가공학회지
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• 제20권7호
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• pp.1-7
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• 2021

Restrictions on the emissions of nitrogen oxides, sulfur oxides, carbon dioxide, and particulate matter from marine engines are being tightened. Each of these emissions requires different reduction technologies, which are costly and require many pieces of equipment to meet the requirements. Liquefied natural gas (LNG) fuel has a great advantage in reducing harmful emissions emitted from ships. Therefore, the marine engine application of LNG fuel is significantly increasing in new ship buildings. Accordingly, this study analyzed the internal support structure, insulation type, and fuel supply piping system of a 35 m3 International Maritime Organization C type pressurized storage tank of an LNG-fueled ship. Analysis of the heat transfer characteristics revealed that A304L stainless steel has a lower heat flux than A553 nickel steel, but the effect is not significant. The heat flux of pearlite insulation is much lower than that of vacuum insulation. Moreover, the analysis results of the constraint method of the support ring showed no significant difference. A553 steel containing 9% nickel has a higher strength and lower coefficient of thermal expansion than A304L, making it a suitable material for cryogenic containers.

• 대한기계학회논문집A
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• 제33권2호
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2013년도 추계학술대회 논문집
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• pp.126-131
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• 2013

As the testing instrument for seismic research, the multi-platform shaking table system of SESTEC in the Pusan National University was introduced to suggest the multi-support shaking table testing methods and also to investigate its ability and applicability. 2 spans single-pylon cable-stayed bridge model, 3 spans girder bridge model and nuclear piping system model are presented and the acceleration and displacement table feedbacks of the each tests are compared to verify the simultaneous excitation ability in time domain and frequency domain.

• 한국정밀공학회지
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• 제22권6호
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• pp.83-89
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• 2005

Most domestic fossil power plants have exceeded 100,000 hours of operation with the severe operating condition. Among the critical components of fossil power plant, high temperature steam pipe systems have had a many problems and damage from unstable displacement behavior because of frequent start up and shut down. In order to prevent the serious damage and failure of the critical pipe system in fossil power plants, 3-dimensional displacement measurement system was developed for the on-line monitoring. Displacement measurement system was developed with a use of a LVDT type sensor and two rotary encoder type sensors. This system was installed and operated on the real power plant successfully.

• 한국항해항만학회지
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• 제46권6호
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• pp.562-569
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• 2022

기존 운항선박에 적용되어 있는 알람 모니터링 기술은 온도, 압력 등의 데이터 항목을 AMS(Alarm Monitoring System)으로 관리하고 해당 센싱 데이터가 정상 수준 범위를 초과할 경우만 선원에게 알람을 제공한다. 또한 기존 선박의 정비는 PMS(Planned Maintenance System)를 따른다. 이는 장비로부터 측정된 센싱 데이터가 설정범위 이상으로 측정되어 이에 따른 알람을 통해 정비하거나, 대상 기기의 고장 유무에 관계없이 일정 시간 사용 후 해당 부품을 사전에 교체하는 방식으로 운영되고 있다. 하지만 선박 기관운영의 신뢰성과 운항 안전성을 확보하기 위해서는 실시간 상태 모니터링 데이터 기반의 사전적 진단 및 예측이 가능해야 한다. 그러기 위해서 실선 데이터를 종합적으로측정하여 데이터베이스화 하고 이를 선박의 보조기기와 배관의 상태기반 예지보전을 위한 상태 진단 모니터링 시스템을 구현하고자 한다. 특히 반응형 웹 기반으로 선박의 보조기기와 배관 상태 정보를 관리할 수 있도록 하였으며, 선내 개인용 컴퓨터(Personal Computer, PC)에서 보는 용도뿐만 아니라 스마트폰 등 다양한 모바일 기기의 접근 및 활용이 가능하도록 화면과 해상도에 맞춰 최적화된 상태 관리가 가능하도록 하여 업데이트 비용이 적게 들며, 관리 방법도 쉽다. 본 논문에서는 자율운항선박 핵심 기술인 상태기반정비(Condition Based Management, CBM) 기술력을 확보하기 위해 선박의 보조기기 중 펌프와 청정기, 그리고 배관 중 해수 및 스팀 배관의 상태 진단 모니터링을 통해 이상 현상을 파악하고, 이를 통해 융합 분석할 수 있도록 선박 보조기기 및 배관의 성능 진단 및 고장 예측에 활용하여 예방정비 의사결정을 지원하고자 한다.

• Nuclear Engineering and Technology
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• 제25권2호
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• pp.204-214
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• 1993

본 논문은 원자력발전소의 배관설계에 파단전 누설(leak-before-break : LBB) 개념이 적용됨에 따라 새롭게 해석대상이 된 분기관파단에 의한 노심지지배럴의 쉘응답을 계산한 것이다. 앞으로 직경 10인치 이상의 고에너지 배관에 대해 LBB 개념이 적용될 것으로 예상되는 바, 이 경우 LBB 적용대상에서 제외되는 유일한 1차측 배관인 3인치 가압기 분무관의 파단을 가정하였고 이때 노심 지지배럴에 가해지는 쉘응답을 구하였다. 이들 응답을 직경 10인치 이상인 배관파단시의 응답과 비교한 결과 앞으로 직경 10인치 이상의 배관에 대해 LBB 개념이 적용될 경우 배관파단에 대한 노심지지배럴의 쉘응답은 무시할 수 있음을 보였다.

• Journal of Mechanical Science and Technology
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• 제16권10호
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• pp.1229-1238
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• 2002

In this paper, a method to obtain the sensitivity of eigenvalues and the random responses of the structure with uncertain parameters is proposed. The concept of the proposed method is that the perturbed equation of each uncertain substructure is obtained using the finite element method, and the perturbed equation of the overall structure is obtained using the mode synthesis method. By this way, the reduced order perturbed equation of the uncertain system can be obtained. And the response of the uncertain system is obtained using probability method. As a numerical example, a simple piping system is considered as an example structure. The damping and spring constants of the support are considered as the uncertain parameters. Then the variations of the eigenvalues, the correlation function and the power spectral density function of the responses are calculated. As a result, the proposed method is considered to be useful technique to analyze the sensitivities of eigenvalues and random response against random excitation in terms of the accuracy and the calculation time.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.322-326
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• 2008

In this paper, we have studied the images which have been reconstructed by using combination of images acquired by the variation of operating frequency. When inner images have been reconstructed, they have been superposed by the surface state effect. In this case, the images of the phase object can be enhanced by the contrast of inner images. There is a kind of specimen, one is a reference block having 1/4T, 1/2T, 3/4T side drilled holes as main run piping material of the steam generator in nuclear power plants. It has been shown that the two results of defect shapes have better than before in this processing and phase contrast grow about twice. And we have constructed the acoustic microscope by using a quadrature detector that enables to acquire the amplitude and phase of the reflected signal simultaneously. Further more we have studied the reconstruction method of the amplitude and phase images, the enhancement method of the defect images' contrast.

• Journal of Mechanical Science and Technology
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• 제19권5호
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.