• 비파괴검사학회지
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• 제3권2호
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• pp.44-47
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• 1984

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.173-173
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• 1998

• 비파괴검사학회지
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• 제14권1호
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• pp.32-38
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• 1994

현재 국내의 원자력(原子力) 발전소(發電所)는 경수로형 원자로(PWR)가 대부분이며, PWR의 경우 ASME Sec. XI의 요건에 따라 가동전(稼動前) 중(中) 검사(檢査)(preservice and inservice inspection)을 핵연료 교체 및 정기 보수 기간에 수행하고 있다. 현재 83년 및 86년 규격들이 검사에 적용되고 있으며 조만간 89년 규격(規格)이 적용(適用)될 시점이다. 따라서 원전 주요 부품에 대한 ASMB규격의 검사(檢査) 요건(要件)의 변화(變化) 과정(過程)을 살펴 보고 검사 강화 및 완화 내용을 검토하였다. 규격은 71년부터 89년까지를 대상으로 편집년도기준(編輯年度基準)으로 검토하였으며, 부록판의 내용은 제외하였다.

• 비파괴검사학회지
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• 제12권2호
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• pp.14-20
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• 1992

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in accordance with ASME Sec. III in order to predict possibility of fatigue failure due to thermal stratification phenomena in pipes connected to reactor coolant system of nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants.

• 비파괴검사학회지
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• 제5권1호
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• pp.34-47
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• 1985

The developing country, KOREA where does not possess the natural resources for traditional energy such as oil and gas, so. The nuclear energy is the most single reliable source available for closing the energy gap. For these reason, It is inavoidable to construct the nuclear power plant and to develop technology related nuclear energy. The rate of operation in large nuclear power facilities depends upon the performance of work system through design and construction, and also the applied technology. Especially, it is the most important element that safety and reliability in operation of nuclear power plant. In view of this aspects, Nuclear power plant is performed severe examinations during preservice and inservice inspection. This study provide an automation of analysis for volumetric examination which is required to nuclear power plant components. It is composed as follows: I. Introduction II. Inservice Inspection of Nuclear Power Plant ${\ast}$ General Requirement. ${\ast}$ Principle and Methods of Ultrasonic Test. ${\ast}$ Study of Flaw Evaluation and Design of Classifying Formula for Flaws. III. Design of Automation for Flaw Evaluation. IV. An Example V. Conclusion In this theory, It is classifying the flaws, the formula of classifying flaws and the design of automation that is the main important point. As motioned the above, Owing to such as automatic design, more time could be allocated to practical test than that of evaluation of defects, Protecting against subjective bias tester by himself and miscalculation by dint of various process of computation. For the more, adopting this method would be used to more retaining for many test data and comparative evaluating during successive inspection intervals. Inspite of limitation for testing method and required application to test components, it provide useful application to flow evaluation for volumetric examination. Owing to the characteristics of nuclear power plant that is highly skill intensive industry and has huze system, the more notice should be concentrated as follows. Establishing rational operation plan, developing various technology, and making the newly designed system for undeveloped sector.

• 비파괴검사학회지
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• 제18권4호
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• pp.321-329
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• 1998

발전설비의 건전성은 원자력발전소의 안전운영을 위하여 필수적이며 이를 위하여 원자력발전소의 안전 규제법에서는 원전사업자 책임하에 신뢰성있는 가동중검사를 주기적으로 수행하도록 규정하고 있다. 따라서 한전은 국내 유일의 원전사업자로서 가동중검사의 정확도 및 신뢰성을 높이기 위한 기술개발에 최선을 다하고 있으며, 전력연구원은 한전의 기술개발 및 연구업무의 중추적 역할을 담당하는 부서로서 효율적인 원전 가동중검사 수행을 위하여 종합지원체계 (total support system : TSS)를 확립하고, 국내기술자립을 위한 요소기숱개밭은 물론 검사요원의 교육 및 자격인증체계 확립에 만전을 기하고 있다. 이 논문은 이와 같은 원자력발전소의 발전설비 건전성 진단을 위하여 현재 전력연구원에서 수행중인 가동중검사 기술개발업무 전반에 관해 상세하게 설명하기 위한 것이다.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.20-24
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• 2014

Recently, it is getting more and more important ensuring the integrity for the equipment degradation according to the increase of nuclear power plant operating period. In many equipment of the nuclear power plant, snubbers mainly installed in reactor coolant pumps, steam generators and piping protected the equipment and piping from the occurrence of transient dynamic loads such as the earthquake, thermal load during the plant operation. This report describes the function, regulation, inspection requirements and management status of the snubbers installed in domestic nuclear power plants.

• 비파괴검사학회지
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• 제15권2호
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• pp.407-414
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• 1995

원자력발전소의 주요 기기들에 대한 가동중검사는 관련법규에 따라 철저히 수행되고 있다. 그러나 최근 선진국에서는 이에 만족하지 않고 원전 기기의 안전성을 더욱 확고히 하기 위해 기존의 가동중검사 요건을 계속 강화하고 있으며, 원전 관련 당사자들은 강화된 요건들을 충족시키기 위한 노력을 끊임없이 계속하고 있다. 이 글에서는 원전 기기 가동중검사 신규 요건들인 초음파탐상검사 시스템의 기량검증(Performance Demonstration) 요건, 비파괴검사자 및 초음파검사자 자격 인정 요건(ANSI/ASNT CP-189, Appendix VII of ASME Sec. XI), 증기발생기 전열관 와전류검사, 신호평가자 자격인정(Qualified Data Analyst : QDA), 미국규제기관(NRC)에서 발행하고 있는 NRC Bulletin, NRC information 등의 가동중검사 관련 사항들을 살펴보고 선진 외국에서는 이들 요건 및 정보에 대해 어떻게 대처하고 있는가를 알아본다. 또한 국내에서도 이들 신규 요건에 대한 대처 현황과 대처 방안을 모색한다.

• 비파괴검사학회지
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• 제11권1호
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• pp.52-60
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• 1991

The importance and role of preservice and inservice inspection(PSI/ISI) for nuclear power plant components are intimately related to plant design, safety, reliability and operation etc.. The Korea Atomic Energy Research Institute(KAERI) has been performing PSI/ISI in Korea since the PSI of Kori nuclear power plant, unit 1 had been performed in 1977. KAERI has localized PSI/ISI technology and has done much experience in ultrasonic flaw detection, evaluation and its excavation on the weldments of large pressure vessels. The results of flaw estimation using ultrasonic examination are compared with the actual flaw sizes revealed by field excavation. KAERI's experience regarding PSI/ISI was described and some discussions were added.

• 비파괴검사학회지
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• 제7권1호
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• pp.42-50
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• 1987

Most of significant defects in the pressure boundaries of nuclear power plant we re dispositioned to be monitored periodically every inservice inspection. Due to the difficulty of the defect sizing during operation, it is necessary to develope the continuous flaw monitoring techniques. The Tip Diffraction method, specifically speaking, spot seems to be suitable for flaw monitoring. The optimum conditions of selecting the transducer were 3.5 MHz and 45-57 degree according to compatibility with the defect height. The effective calculation of the defect height was to assume the fact that the incident beam is parallel. This method would be supplemented to ASME method about the defect characterization for the surface flaw.