• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.286-291
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• 2000

The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and number of subclad cracks have been found during an in-service-inspection. Therefore assessment for subclad cracks should be made for normal operating conditions and faulted conditions such as PTS. Thus, in order to find the optimum fracture assessment procedures for subclad cracks under a pressurized thermal shock condition, in this paper, three different analyses were performed, ASME Sec. XI code analysis, an LEFM(Liner elastic fracture mechanics) analysis and an EPFM(Elastic plastic fracture mechanics) analysis. The stress intensity factor and the Maximum $RT_{NDT}$ were used for characterizing. Analysis based on ASME Sec. XI code does not completely consider the actual stress distribution of the crack surface, so the resulting Maximum allowable $RT_{NDTS}$ can be non-conservative, especially for deep cracks. LEFM analysis, which does not consider elastic-plastic behavior of the clad material, is much more non-conservative than EPFM analysis. Therefore, It is necessary to perform EPFM analysis for the assessment of subclad cracks under PTS.

• 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.

• 대한기계학회논문집A
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• 제24권1호
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• pp.206-214
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tube s. the integrity evaluation must be carried out. and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire: integrity evaluation process. For this reason. an integrity evaluation system, which provides efficient of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). This system does not only provide various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications but also is compatible with other commercial database software. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• 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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• 대한기계학회논문집A
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• 제35권4호
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• pp.439-445
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• 2011

대기오염 등 환경에 대한 관심이 높아지면서 경유차 배출 가스 저감을 위한 최적의 대안으로 최근 세계적으로 천연가스차량의 보급이 크게 확대되고 있는 추세이다. 이러한 추세에도 불구하고 그 안전성에 대해서는 많은 논란을 불러 일으키고 있으며 최근 그 이용에 따른 안전사고 또한 빈번하게 발생하고 있다. 천연가스버스 압력용기 파열 사고는 대중들이 많이 이용한다는 점에서 대형참사로 이어질 수 있는 잠재력을 가지고 있어 그 심각성은 매우 크다고 할 수 있다. 이에 법공학적인 측면에서 좀 더 전문화되고 체계적인 사고조사와 원인 규명을 통해서 사전에 예방대책을 마련함으로써 유사 및 동종재해의 발생을 최소화해야 한다. 본 연구에서는 구조해석을 통한 용기의 설계 검증과 용기의 파손형태 검사 및 재료 물성평가 등을 통해 용기파손에 의한 CNG 차량 폭발 사고에 관한 정확한 사고 원인을 규명함으로써 동일 형태의 차량 안전사고 예방에 기여하고자 한다.

• 비파괴검사학회지
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• 제34권3호
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• pp.254-259
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• 2014

중수로 원자로는 한 개의 원자로용기로 구성된 경수로와는 달리 약 380여개의 연료채널(fuel channel)로 구성되어 있다. 연료채널을 구성하는 압력유지 기기인 압력관(pressure tube)은 지르코니움 합금(Zr-2.5wt% Nb) 재질로서 치수는 내경이 103.4 mm, 두께가 약 4.19 mm, 길이가 6.36 m인 튜브 형태의 관이다. 압력관은 내부에 핵연료 다발과 냉각재가 내장되며 압력관의 기능은 연료를 지지하고 열수송 유체인 중수($D_2O$)를 이송한다. 압력관의 단순한 기하학적인 형상으로 인하여 자동화 비파괴검사가 가능하고 접근성이 우수하다. 연료채널은 경수로형 원전과 동일하게 설치전과 운전중에 원자력안전위원회 법령 요건에 따라 주기적으로 엄격한 비파괴검사를 수행하여 건전성을 확인한다. 연료채널의 주기적 비파괴검사에는 초음파탐상 및 와전류탐상검사 기법을 적용한 체적 비파괴검사 기술이 적용된다. 이중에서 와전류탐상검사 기법은 초음파탐상검사에서 검출된 결함의 확인을 위한 보충검사기술로 적용되고 있지만 표면결함에 대한 검출능이 초음파탐상검사 기법보다 우수한 장점을 가지고 있다. 본 논문에서는 압력관 내부 표면 비파괴검사에 적용되고 있는 와전류탐상검사 기술의 압력관 내면에 발생할 수 있는 결함의 검출 및 깊이 측정 특성에 대한 연구결과를 기술하였다. 즉, 와전류검사 기술은 압력관 내면에 발생할 수 있는 아주 미세한 결함을 매우 우수한 분해능으로 검출할 수 있으므로 초음파탐상검사 결과 확인을 위한 보충기술로서 매우 유용하지만, 결함의 깊이 측정은 오차가 매우 크게 발생하므로 결함 깊이 측정에는 적합하지 않고 오직 표면결함 검출에만 적용하는 것이 바람직하다.

• 기계저널
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• 제18권4호
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• pp.35-40
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• 1978

여기서 상기 ASME Code에 대하여 간단히 설명하기로 한다. ASME Code 는 첫부분에서 ASME Code의 적용을 받아야 하는 압력용기를 정의하고, 압력용기의 건설에 관한 일반원칙을 설명한후 그 다음에는 세개의 Subsection으로 나뉘어져 있다. 즉 Subsection A General Requirements Subsection B Requirements Pertaining to Methods of Fabrication of Pressure Vessels Subsection C Requirements Pertaining to Classes of Material 여기서 Subsection A는 압력용기 재료나 제작방법의 상위와 관계없이 적용하여야 할 일반적인 요구사항을 규정한 것이며, Subsedction B에서는 압력용기의 제작방법을 용접,리벳팅,단조,경납 땜의 4가지로 나누어 각 제작방법에 따른 특수 요구사항을 규정하였고, 마지막으로 Subsection C 는 재료에 따른 특수 요구사항을 규정한 것이다. 이 각 Subsection은 다시 General, Materials, Design, Fabrication, Inspection and Tests, Stamping and Reports, Pressure Relief Devices로 나누어 이에 대한 각각의 요구사항들을 설명하고 있다. 그러나 이 기술기준에서는 제정방향으로, 다음의 목차에서도 알 수 있는 바와 같이 이들의 순서를 바꾸어 총칙,재료,설계,제작, 검사 및 시험, 압력릴리프장치를 배치한 후 이미 KS B 6231에 제정되어 있는 것은 그 규정을 대부분 그 대로 인용하였고, 그렇지않은 것은 우리의 실정을 참작하여 삭제, 보완, 수정하였다. 삭제한 내용 중 대표적인 것으로 공인검사관(Authorized Inspector) 및 Stamping and Reports 에는 9개의 Mandatory Appendix 와 16개의 Nonmandatory Appendix가 있는데, 이둘 중 이 기술기준에서 필요하다고 생각되는 것은 발췌 수록하였다. 단위에 대해서는 국가시책에 따라 메트릭 시스템을 사용하였고 단위의 환산에서 야기되는 소수점등의 처리는 공학적인 판단에 의거하였다.

• 대한기계학회논문집A
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• 제20권4호
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• pp.1233-1240
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• 1996

Conventional measurement methods using ultrasonic wave or x-ray, eddy current for non-destructive testing(NDT) in nuclear power plants and other industrial plants have been utilized as the method of contact with objects to be inspected. For this reason these methods require relatively much time and inspection area is limited by the location of probe or film. But holograpic interferometry which is a non-contact optical measurement method using a coherent light source has an advantage that quantative measurement can be performed at a time. In this paper a new method using realtime holographic interfreometry and image processing for detecting internal flaws of pressure vessels is presented.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.166-171
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• 2004

Since early 1950's fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet bas been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an internet-based cooperative system for integrity evaluation system which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and agent programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet.