• 한국압력기기공학회 논문집
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• 제11권2호
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• pp.20-24
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• 2015

During refueling Outage, In-service inspections(ISIs) for the Nuclear Power Plant components are mandatory requirement in accordance with ASME Code Sec. XI. Especially, in current ultrasonic testing is one of the most important NDT techniques that are used for volumetric examination methods for nuclear power plant components, and accurate sizing of flaw indication by UT is essential to assure the integrity of the components. However, ASME code specifies minimum requirement for vessel examination procedure, and so far many different flaw sizing approaches have been tried to apply. Through the Round Robin Test(RRT), the accuracy of ultrasonic flaw sizing using DAC techniques was measured with the mock-ups simulating typical pressure vessel welds. These mock-ups contain artificially introduced flaws of known size and location. This paper shows experimental comparison data on the accuracy of techniques using such as 6dB drop, 50%DAC, 20%DAC and 20%DAC with beam spread correction, and also shows that diverse DAC techniques can be effectively applied to the assessment of the flaw sizing for pressure vessel welds in the stage of welding and fabrication.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.291-295
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• 1996

In CANDU reactor, each pressure tubes contain twelve fuel bundles and provide the inlet and outlet for the primary coolant. If a leak develops in the pressure tube, it is detected by Annulus Gas System which contains circulating dry $CO_2$ gas. Since the leaks caused by the flaws are resulted in pressure tube break, establishment of flaw assessment method is very significant in view of the fracture mechanics. In this paper, various criteria for assessing the flaws are presented to prevent the tube rupture and ensure the integrity of reactor operating.

• Journal of Welding and Joining
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• 제27권2호
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• pp.51-56
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• 2009

Film based radiography imaging technique has been applied to the non-destructive test in medical, aircraft, and power industries contributing to the development of the industries. However, the complex process for imaging and analysis has increasingly demanded the reformation of the radiography test. A digital radiography imaging technologies has been com out from the demand. This study was mainly focused on the assessment on the accuracy for the each image from digital radiography test and film radiography test was proven to crate a better image in sensitivity than film radiography test. In the IQI(Image quality indicator) transmission test, one or two more line can be seen in digital image than in film image. When applying to the boiler tube weld, film image is detectable to the 1.0mm depth flaw; and digital image to the 0.5mm depth flaw. As a result of this study, digital radiography technology is determined to enhance the image quality, compared to film radiography technologies

• 대한기계학회논문집A
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• 제26권11호
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• pp.2305-2311
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• 2002

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and heavy water coolant. Pressure tubes are installed horizontally inside the reactor and only selected samples are periodically examined during in-service inspection. In this respect, a probabilistic safety assessment method is more appropriate fur the assessment of overall pressure tube safety. The failure behavior of CANDU pressure tubes, however, is governed by delayed hydride cracking which is the major difference from pipings and reactor pressure vessels. Since the delayed hydride cracking has more widely distributed governing parameters, it is impossible to apply a general PFM methodology directly. In this paper, a PFM methodology for the safety assessment of CANDU pressure tubes is introduced by applying Monte Carlo simulation in determining failure probability Initial hydrogen concentration, flaw shape and depth, axial and radial crack growth rate and fracture toughness were considered as probabilistic variables. Parametric study has been done under the base of pressure tube dimension and hydride precipitation temperature in calculating failure probability. Unstable fracture and plastic collapse are used for the failure assessment. The estimated failure probability showed about three-order difference with changing dimensions of pressure tube.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.1211-1215
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• 2003

Pressure tubes are major component of nuclear reactor, but only selected samples are periodically examined due to numerous numbers of tubes. Current in-service inspection result show there is high probability of flaw existence at un-inspected pressure tube. Probabilistic analysis is applied in this study for the integrity assessment of un-inspected pressure tube. But all the current integrity evaluations procedures are based on conventional deterministic approaches. So many integrity evaluation parameters are not directly apply to probabilistic analysis. As a result of this study failure assessment diagram are proposed based on test data.

• 대한기계학회논문집A
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• 제28권3호
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• pp.289-295
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• 2004

Pressure tubes are major component of nuclear reactor, but only selected samples are periodically examined due to numerous numbers of tubes. Current in-service inspection result show there is high probability of flaw existence at uninspected pressure tube. Probabilistic analysis is applied in this study for the integrity assessment of uninspected pressure tube. All the current integrity evaluations procedures are based on conventional deterministic approaches. So it is expected that the results obtained are too conservative to perform a rational evaluation of lifetime. More realistic failure criteria, based on FAD are also proposed for the probabilistic analysis. As a result of this study failure probabilities for various conditions are calculated, and examined application of FAD and LBB concept.

• 에너지공학
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• 제20권2호
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• pp.90-95
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• 2011

이 연구는 압축천연가스자동차용 Type III 용기 복합재에 발생한 스크래치, 컷, 가우지 등의 손상을 평가하여 계속 사용여부를 판정하기 위해 수행하였다. 결함내구성 시험결과, 복합재의 손상등급에 따라 ISO 19078에서 정한 최소요구 반복회수(11,250회)는 모두 만족하였고 용기의 반복성능은 손상등급 1, 2에서는 20,000회를 상회하다가 손상등급 3에서는 18,000~21,000회의 범위를 나타내었다. 실험용기 12개 중 8개가 Type III 용기복합재 결함의 영향을 받아 파괴되었고 나머지는 결함과 관계없이 파괴된 것으로 판단된다. 전산모사에 의한 구조해석결과는 $1.25\;mm{\times}200\;mm$의 결함모델에 사용압이 가해졌을 때 발생한 응력은 79.5 MPa이고 무결함 용기의 라이너 평탄부에 발생한 응력은 66.6 MPa로서 약 19.37%의 차이가 발생하였다. 이것은 라이너의 피로수명에 영향을 미칠 수 있는 차이로 볼 수 있고 이 결함은 용기의 반복성능에 영향을 미치는 임계값으로 판단할 수 있었다.

• 한국가스학회:학술대회논문집
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• 한국가스학회 1999년도 추계학술발표회 논문집
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• pp.151-158
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• 1999

• 대한기계학회논문집A
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• 제24권5호
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• pp.1331-1342
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• 2000

In power generation systems a variety of structural components typically operate at high temperature and pressure. Therefore a life assessment methodology accounting for gradual creep fracture is increasingly needed for these components. The most critical defects in such structure are generally found in the form of semi-elliptical surface cracks in the welded tubular joints. Therefore the analysis of a semi-elliptical surface crack in a plate or a shell is an important problem in engineering fracture mechanics. On this background, via shell/line-spring finite element analyses of such surface cracks in the welded T and L joints under various loadings, we investigate J-integral along the crack front We first develop T and L joints auto mesh generation program providing ABAQUS input file composed of shell/line-spring finite elements. We then further develop a T and L joints life assessment program based on the experimental creep crack growth law and auto mesh generation program in a graphical user interface format Finally the remaining life of T and L joints for various analytical parameters are assessed using the developed life assessment program.

• 대한기계학회논문집A
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• 제25권2호
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• pp.275-282
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• 2001

The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement. Number of subclad cracks may be found during an in-service-inspection due to the presence of cladding. It is specified, in ASME Sec. XI, that a subclad crack is characterized as a surface crack when the thickness of the clad is less than 40% of the crack depth. This condition is provided to keep the crack integrity evaluation conservative. In order to refine the fracture assessment procedures for such subclad cracks under a pressurized thermal shock condition, three dimensional finite element analyses are applied for various subclad cracks existing under cladding. A total of 36 crack geometries are analyzed, and the results are compared with those for surface cracks. The resulting stress intensity factors for subclad cracks are 6 to 44% less than those for surface cracks. It is proven that the flaw characterization condition as specified in ASME Sec. XI can be overly conservative for some subclad cracks.