• 대한전자공학회:학술대회논문집
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• 대한전자공학회 2002년도 하계종합학술대회 논문집(5)
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• pp.153-156
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• 2002

The remotely operated underwater robotic vehicle system has been required to inspect some objects such as baffle former bolts and remove foreign objects in reactor vessel of nuclear power plant. In this paper, we have designed the remotely operated underwater robotic vehicle system that includes a long reach arm that is composed of 4 joints to remove foreign objects in a narrow space, a camera for visual test, instrument sensors for vehicle positioning, 4 thrusters for underwater navigation of vehicle, and supervisory control system implemented with industrial PC that includes robot simulator that has the functions of real time visualization, robot work planning and etc.

• 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.

• 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.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 추계학술대회 논문집
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.

• 방사성폐기물학회지
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• 제18권3호
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• pp.347-354
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• 2020

Site characterization for decommissioning Kori Unit 1 is ongoing in South Korea after 40 years of successful operation. Kori Unit 1's containment building is assumed to be mostly radioactively contaminated, and therefore radiation exposure management and detailed contamination investigation are required for decommissioning and dismantling it safely. In this study, site-specific Derived Concentration Guideline Levels (DCGLs) were derived using the residual radioactivity risk evaluation tool, RESRAD-BUILD code. A conceptual model of containment building for Kori Unit 1 was set up and limited occupational worker building inspection scenario was applied. Depending on the source location, the maximum contribution source and exposure pathway of each radionuclide were analyzed. The contribution of radionuclides to dose and exposure pathways, by source location, is expected to serve as basic data in the assessment criteria of survey areas and classification of impact areas during further decommissioning and decontamination of sites.

• 한국정보과학회:학술대회논문집
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• 한국정보과학회 2001년도 가을 학술발표논문집 Vol.28 No.2 (2)
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• pp.238-240
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• 2001

원자력발전소에서는 기기, 배관 및 각종 지지구조물 등 설비에 대하여 시간의 경과에 따른 취약화 정도를 측정하기 위하여 대략 15개월을 주기로 호기별 비파괴검사로 감시 및 평가하는 가동중검사를 실시한다. 증기발생기, 주복수기와 같은 세관으로 구성된 설비는 와전류탐상검사를 수행하여 신호데이터를 취득하고 건전성 여부를 평가한 다음 그 결과를 Optical Disk에 신호데이터와 함께 저장한다. 본 논문에서는 저장된 방대한 양의 검사 결과 파일을 추출하여 데이터베이스로 구축하고, 행열 수량, 모양, 방향 및 열번호 부여방법이 상이한 다양한 배열 형태의 세관 Map을 편집하여 사용자 요구에 따라 검사정보를 색상 Tube로 Mapping 처리하여 세관의 상태, 검사이력, 결함성장률 및 변화추이 분석을 시각적으로 파악할 수 프로그램 구현 사례를 소개한다.

• Journal of Electrical Engineering and information Science
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• 제2권4호
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• pp.59-65
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• 1997

In this paper, a general fuzzy syntactic method for recognition of flaw patterns and for the measurement of flaw characteristic parameters for a non-destructive inspections signal, called eddy-current, is presented. Solutions are given to the subtasks of primitive pattern selection, signal to symbol transformation, pattern grammar formulation, and event-synchronous flaw pattern extraction based on the grammars. Fuzzy attribute grammars are used as the model for the pattern grammar because of their descriptive power in the face of uncertain constraints caused by nose or distortion in the signal waveform, due to their ability to handle syntactic as well as semantic information. This approach has been implemented and the performance of eh resultant system has been evaluated using a library of law patterns obtained from steam generator tubes in nuclear power plants by an eddy current-based non-destructive inspection method.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 1996년도 특별강연 및 추계학술발표 개요집
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• pp.19-34
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• 1996

A review was made of fatigue failures of nuclear power plant components in Japan, which were experienced in service and during periodical inspection. No case has been recently reported of a service fatigue failure of a reactor pressure vessel itself, excluding nozzle corner cracks, that occurred many years ago. But, service fatigue failures have been occasionally experienced in piping systems, pumps, and valves, on which fatigue design seems to have been inadequately applied. The causes of fatigue failures can be divided into two categories: mechanical-vibration-induced fatigue and thermal-fluctuation-induced fatigue. Vibration-induced fatigue failure occurs more frequently than is generally thought. The lesson gleaned from the present survey is a recognition that a service fatigue failure may occur due to any one or a combination of the following factors: (1) lack of communication between designers and fabrication engineers, (2) lack of knowledge about a possibility of fatigue failure and poor consideration about the effects of residual stresses, (3) lack of consideration on possible vibration in the design and fabrication stages, and (4) lack of fusion or poor penetration in a welded joint.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1998년도 가을 학술발표대회 논문집(III)
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• pp.649-655
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• 1998

In this study, the personal-computer program was developed to predict prestressing losses containment structures of Nuclear Power Plants by concrete creep and shrinkage. This program is constituted of three parts, which are pre-processor, calculation module and post-processor. Input data for his program are : material properties of concrete, rebar, liner and duct, test results of concrete creep and shrinkage, relative humidity, dimension of containment structures, and the number of prestressing tendon related on containment structures. To obtain better results, this program was made to reflect the prestressing losses due to influence that occurred after prestressing each tendon, thus it can predict prestressing losses and allowable prestressing forces of each tendon. As a case study, this program was applied to containment structures of Youngwang 3 & 4 NPP's and analytical result was compared with test results in In-service Inspection of containment structures. From this comparison, it was proved that this program could well predict prestressing losses by concrete creep and shrinkage.

• Corrosion Science and Technology
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• 제15권6호
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• pp.320-325
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• 2016

To perform ultrasonic testing (UT) thickness measurement of the secondary side piping installed in nuclear power plants, the insulation for preventing heat loss should be removed. The type of insulation can be divided into fixed and removable insulation. Fixed and removable insulation have their own strengths and weaknesses. Removable insulation has been installed in the components susceptible to wall thinning caused by FAC and erosion from Shin-Kori unit 1, which commenced its commercial operation in 2011. In this paper, the number of repeated inspections of components and the number of replacements of fixed insulation were estimated and a more economical way was identified based on the manufacturing and installation costs for fixed and removable insulation.