• Nuclear Engineering and Technology
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• 제45권2호
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• pp.265-276
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• 2013

Pressure tubes made of Zr-2.5 wt% Nb alloy are important components consisting reactor coolant pressure boundary of a pressurized heavy water reactor, in which unanticipated through-wall cracks and rupture may occur due to a delayed hydride cracking (DHC). The Canadian Standards Association has provided deterministic and probabilistic structural integrity evaluation procedures to protect pressure tubes against DHC. However, intuitive understanding and subsequent assessment of flaw behaviors are still insufficient due to complex degradation mechanisms and diverse influential parameters of DHC compared with those of stress corrosion cracking and fatigue crack growth phenomena. In the present study, a deterministic flaw assessment program was developed and applied for systematic integrity assessment of the pressure tubes. Based on the examination results dealing with effects of flaw shapes, pressure tube dimensional changes, hydrogen concentrations of pressure tubes and plant operation scenarios, a simple and rough method for effective cooldown operation was proposed to minimize DHC risks. The developed deterministic assessment program for pressure tubes can be used to derive further technical bases for probabilistic damage frequency assessment.

• 한국가스학회지
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• 제20권5호
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• pp.112-117
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• 2016

국내에서는 연료 가스 배관용 탄소강관으로 KS D 3631의 기준을 따르도록 되어있으며, 탄소강관이 지하에 매설되는 경우 부식을 방지하기 위해 강관에 합성수지로 피복을 한다. 피복은 부식 방지 조치를 하는 것으로 압축된 연료가스를 이송하는데 필요한 강도는 원관인 탄소강관이 그 역할을 수행한다. 따라서 탄소강관에 결함이 발생하였을 때, 그 결함에 따른 강도를 평가하는 것이 중요한데 국내에는 잔존강도를 평가하기 위한 방법이 부족하다. 결국 국외 기준인 DNV RP F101 또는 ASME B31G의 기준에 따라 잔존강도를 평가하는데, 그 결과가 국내 배관에 적용 타당성이 있는지 검증이 부족하며, 85% 이상의 결함에 대해서는 예측 할 수 없다. 따라서 본 연구에서는 KS D 3631의 배관을 이용하여 인위적 결함을 가공하여 잔존강도를 평가하기 위한 실험을 수행하였다. 그 결과 폭에 따라서 일정한 잔존강도의 차이를 보이며, 국내 연료 가스용 배관에는 DNV RP F101이 적합하다는 결론을 도출하였다. 본 연구의 결과는 결함 깊이가 큰 경우에 대한 결과와 타당한 국의 기준을 파악하기 위한 결과를 포함하고 있어, 배관의 잔존강도를 이용한 위험성 평가 등에 유용하게 활용될 것으로 기대한다.

• Smart Structures and Systems
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• 제19권4호
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• pp.371-382
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• 2017

Concrete filled steel tubes are extensively applied in engineering structures due to their resistance to high tensile and compressive load and convenience in construction. But one major flaw, their vulnerability to environmental attack, can severely reduce the strength and life of these structures. Degradation due to corrosion of steel confining the concrete is one of the major durability problems faced by civil engineers to maintain these structures. The problem accelerates as inner surface of steel tube is in contact with concrete which serves as electrolyte. If it remains unnoticed, it further accelerates and can be catastrophic. This paper discusses a non-destructive degradation monitoring technique for early detection corrosion in steel tubes in CFST members. Due to corrosion, damage in the form of debonding and pitting occurs in steel sections. Guided ultrasonic waves have been used as a feasible and attractive solution for the detection and monitoring of corrosion damages in CFST sections. Guided waves have been utilized to monitor the effect of notch and debond defects in concrete filled steel tubes simulating pitting and delamination of steel tubes from surrounding concrete caused by corrosion. Pulse transmission has been used to monitor the healthy and simulated damaged specimens. A methodology is developed and successfully applied for the monitoring of concrete filled steel tubular sections undergoing accelerated chloride corrosion. The ultrasonic signals efficiently narrate the state of steel tube undergoing corrosion.

• 비파괴검사학회지
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• 제23권1호
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• pp.38-44
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• 2003

원자력발전소의 안전성등급 기기에 적용되는 비파괴검사는 실제 결함을 실현한 시험편을 사용하여 결함탐지능력을 검증하도록 하는 기량검증이 요구되고 있다. 가동중인 원전에서 발생 가능한 균열으로는 기계적 피로균열, 열 피로균열 및 입계부식균열 등이 있으나 본 연구에서는 기계적 피로균열을 대상으로 하였다. 인장 피로하중을 사용하여 기계적 피로결함을 제조하기 위해서 시험편을 설계하였고 원하는 피로결함 파면의 조도를 얻기 위해서 인가하중의 크기 및 사이클 수를 조절하여 피로결함을 발생시켰다. 발생된 결함에 대한 정확한 크기와 위치에 대한 물리적 정보를 얻은 후에 결함이 설계된 크기와 위치에 존재하도록 기밀용접을 실시하였다. 기밀용접 후 잔여 용접 흠은 가스 텅스텐 아크용접 및 플럭스 코어드 아크용접으로 채워졌다. 최종 완성된 피로결함 시험편을 방사선투과검사 및 초음파탐상검사를 통하여 검사한 결과, 설계된 길이와 깊이로 피로결함이 형성되었음을 확인하였다.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.16-22
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• 2010

Cast austenitic stainless steel(CASS) is used in the primary cooling piping system of nuclear power plant for it's relative low cost, corrosion resistance and easy of welding. However, the coarse-grain structure of cast austenitic stainless steel can strongly affect the inspectability of ultrasonic testing. The major problems encountered during inspection are beam skewing, high attenuation and high background noise of CASS component. So far, the best inspection performance involving CASS components have been achieved using low frequency TRL(Transmitter/Receiver side-by-side L wave) angle beam probe. But TRL technique could not detect shallow defect and it contains an uncertainty for sizing capability. Currently, most of researchers are studying to overcome these challenge issue. In this study, low-frequency phased array TRL technique used to detect and sizing the flaws in CF8A cast austenitic stainless steel.As conclusion, we could detect and size not only axial flaw but also circumferential flaw using low frequency phased array technique.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 2005년도 춘계학술대회 논문집
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• pp.216-227
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• 2005

The head penetrations for control rod drive mechanism and instrumentation systems are installed at the reactor pressure vessel head of PWRs. Primary coolant water and the operating conditions of PWR plants can cause cracking of these nickel-based alloy through a process called primary water stress corrosion cracking (PWSCC). Inspection of the head penetrations to ensure the integrity of the head penetrations has been interested since reactor coolant leakages were found at U. S. reactors in 2000 and 2001. The complex geometry of the head penetrations and the very low echo amplitude from the fine, multiple flaws due to the nature of the see made it difficult to detect and size the flaws using conventional pulse-echo UT methods. Time-of-flight-diffraction technique, which utilizes the time difference between the flaw tips while pulse-echo does the flaw response amplitude from the flaw, has been selected for this inspection for it's best performance of the detection and sizing of the head penetration see flaws. This study defines the limits of the detectable and accurately sizable minimum flaw size which can be detected by the General TOFD and the Delta TOFD techniques for circumferentially and axially oriented flaws respectively. These results assures the reliability of the inspection techniques to detect and accurately size for various kind of flaws, and will also be utilized for the future development and qualifications of the TOFD techniques to enhance the detecting sensitivity and sizing accuracy of the flaws of the reactor head penetrations in nuclear power plants.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2010년도 춘계학술대회 논문집
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• pp.779-780
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• 2010

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2004년도 춘계 학술발표대회 개요집
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• pp.308-310
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• 2004

1980년대 초 미국의 비등형 경수로(Boiling Water Reactor : BWR) 원자력발전소 배관계통의 입계 응력 부식 균열(Inter-Granular Stress Corrosion Crack) 검사 결과 및 미국 EPRI(Electric Power Research Institute)에서 실시한 round robin test 결과에서 기존 초음파 검사 방법의 실효성에 많은 문제점이 제기 되었다. (중략)

• 한국재료학회지
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• 제25권10호
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• pp.516-522
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• 2015

Aging aircraft structures are inevitably exposed to environment for a long time facing many potential problems, including corrosion and wide spread fatigue damage, which in turn cause the degradation of flight safety. In this study, the environmental surface damages on aging aircraft structures induced during service were quantitatively analyzed. Additionally, S-N fatigue tests were performed with center hole specimens extracted from aging aircraft structures. From the results of quantitative analyses of the surface damages and fatigue tests, it is concluded that corrosion pits initiated during service reduce the fatigue life significantly. Finally, using the fracture mechanics and the EIFS (equivalent initial flaw size) concepts, the remaining fatigue life was predicted based on actual fatigue test results.

• 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.