• 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.

• Smart Structures and Systems
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• 제24권3호
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• pp.403-414
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• 2019

Offshore drilling has become a key process for obtaining oil. Offshore platforms have many applications, including oil exploration and production, navigation, ship loading and unloading, and bridge and causeway support. However, vibration problems caused by severe environmental loads, such as ice, wave, wind, and seismic loads, threaten the functionality of platform facilities and the comfort of workers. These concerns may result in piping failures, unsatisfactory equipment reliability, and safety concerns. Therefore, the vibration control of offshore platforms is essential for assuring structural safety, equipment functionality, and human comfort. In this study, an optimal multiple tuned mass damper (MTMD) system was proposed to mitigate the excessive vibration of a three-dimensional offshore platform under ice and earthquake loadings. The MTMD system was designed to control the first few dominant coupled modes. The optimal placement and system parameters of the MTMD are determined based on controlled modal properties. Numerical simulation results show that the proposed MTMD system can effectively reduce the displacement and acceleration responses of the offshore platform, thus improving safety and serviceability. Moreover, this study proposes an optimal design procedure for the MTMD system to determine the optimal location, moving direction, and system parameters of each unit of the tuned mass damper.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.688-706
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• 2024

Austenitic stainless steels (ASS) are extensively employed in various sectors such as nuclear, power, petrochemical, oil and gas because of their excellent structural strength and resistance to corrosion. SS304 and SS316 are the predominant choices for piping, pressure vessels, heat exchangers, nuclear reactor core components and support structures, but they are susceptible to stress corrosion cracking (SCC) in chloride-rich environments. Over the course of several decades, extensive research efforts have been directed towards evaluating SCC using diverse methodologies and models, albeit some uncertainties persist regarding the precise progression of cracks. This review paper focuses on the application of Acoustic Emission Technique (AET) for assessing SCC damage mechanism by monitoring the dynamic acoustic emissions or inelastic stress waves generated during the initiation and propagation of cracks. AET serves as a valuable non-destructive technique (NDT) for in-service evaluation of the structural integrity within operational conditions and early detection of critical flaws. By leveraging the time domain and time-frequency domain techniques, various Acoustic Emission (AE) parameters can be characterized and correlated with the multi-stage crack damage phenomena. Further theories of the SCC mechanisms are elucidated, with a focus on both the dissolution-based and cleavage-based damage models. Through the comprehensive insights provided here, this review stands to contribute to an enhanced understanding of SCC damage in stainless steels and the potential AET application in nuclear industry.

• Journal of Advanced Marine Engineering and Technology
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• 제39권4호
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• pp.425-431
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• 2015

선박의 LNG(liquefied natural gas) 연료 공급 시스템에서 천연가스는 $50^{\circ}C$의 온도와 35MPa의 압력을 가진 가스 상태로 기화기에서 엔진으로 이송되므로, 이러한 열 하중을 고려한 구조 안전성 평가가 반드시 필요하다. 본 연구에서는 먼저 재료에 미치는 열의 영향을 분석하기 위하여 이중 배관의 재료인 슈퍼 듀플렉스 스테인리스강의 어닐링 시간을 고려한 일축 인장 실험을 수행하였다. 또한 구조 안전성 평가를 위해, 현재 널리 사용되는 고정식 지지대를 가지는 고온-고압 이중 배관에 대한 열-구조 해석을 수행하였다. 지지대와 내관 사이의 응력 집중을 최소화하기 위하여, 내관을 따라 미끄러질 수 있는 슬라이딩 지지대의 새 형상들을 제안하였고, 열-구조 해석 결과를 통해 최적의 지지대를 제안하였다. 마지막으로 제안된 지지대를 사용한 전체 이중배관에 대한 해석을 통해 안전성을 평가하였다. 본 연구의 결과는 차후 LNG 연료 공급 시스템의 고온-고압 이중 배관 설계 시 참고자료로서 활용될 수 있으며, 이중 배관의 슬라이딩 지지대를 설계함에 있어서 그 활용가치가 있다고 판단된다.

• 비파괴검사학회지
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• 제33권1호
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• pp.63-68
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• 2013

Some essential components in nuclear power plants are periodically inspected using non-destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in-service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non-magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%.

• Nuclear Engineering and Technology
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• 제39권3호
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• 한국지진공학회논문집
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• 제9권6호
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• pp.13-19
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• 2005

원자력발전소의 기기, 배관 시스템과 같은 부속구조물의 동적 응답을 먼기 위해 사용되는 층응답스펙트럼은 일반적으로 주구조물과 부속구조물의 동적 상호작용이 반영되지 않고 만들어진다. 본 연구에서는 기기와 구조물의 동적 상호작용이 고려된 해석을 통해 층응답스펙트럼을 생성시키는 해석법을 기술하였다. 이 방법은 기기를 모사하는 단자유도계와 기기가 놓여있는 구조물의 임피던스로 분할되는 부분구조 해석법을 적용하여 기기의 응답을 구한다. 단자유도계의 진동수, 감쇠비 및 질량 특성을 변화시키면서 최대 동적 응답을 계산함으로써 일련의 층응답스펙트럼을 작성한다. 전형적인 원자력발전소의 원자로 구조물에서 본 방법을 고려한 층응답스펙트럼과 기기를 포함한 전체 해석으로부터 작성된 층응답스펙트럼과 비교함으로써 본 연구의 타당성을 확인하였다. 기기-구조물 상호작용 효과를 확인하기 위하여 구조물 질량의 1% 이내인 기기에 대하여 기술된 방법과 기존 방법을 각각 적용하여 최대 응답값을 비교하였다. 그 결과 지배 진동수 부근에서 기기-구조물 상호작용을 고려한 응답이 그렇지 않은 경우인 기존방법의 응답에 비하여 저감되는 현상을 보였다.