• 한국정보통신설비학회:학술대회논문집
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• 한국정보통신설비학회 2009년도 정보통신설비 학술대회
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• pp.54-58
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• 2009

In this paper, the ECT(eddy current testing) signal analysis of eddy current array probe for inspection of SG(steam generator) tube in NPP(nuclear power plant) using electromagnetic FEM(finite element method) was performed. To obtain the electromagnetic characteristics of probes, the governing equation was derived from Maxwell's equation, and the problem was solved by using the 3-dimensional FEM. The types of defects were FBH(flat bottomed hole) and OD groove, Spiral groove, natural defects(pitting, SCC, multiple SCC, wear). The depth of FBH defects were 20%, 40%, 60%, 80%, 100 of SG tube thickness, and it was assumed that the defects were located on the tube outside. And the operation frequency of 100kHz, 300kHz and 400kHz were used. Material of specimen was Inconel 600 which is usually used for SG tubes in NPP. The signal difference could be observed according to the variation of size and depth on FBH defects and operation frequencies. The results in this paper can be helpful when the ECT signals from EC array probe are evaluated and analyzed.

• 한국정보통신설비학회:학술대회논문집
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• 한국정보통신설비학회 2006년도 하계학술대회
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• pp.241-245
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• 2006

In the examination of Steam Generator (SG) tube in Nuclear Power Plant (NPP) Eddy Current Testing (ECT) probes play an Important role in detecting the defects. Bobbin probe and Rotating Pancake Coil (RPC) probe is usually used for the inspection of SG tube. Bobbin probe is good at high speed inspection, but ability of detection of circumferential defect is very weak. On the contrary RPC probe, which moves for inspection in the direction of axial and circumferential simultaneously, has very slow inspection speed, but it was excellent detection capability fur small cracks, which is hardly detected by bobbin probe. Many examinations of SG tube examination of NPP are achieved during short period. Therefore, solution about this must develop probe of new form for examination performance and examination time shortening of other probe. In this paper, analyzed technological present condition of Bob-bin probe and RPC probe been using in Nondestructive Testing (NDT) for SG tube defect detection and Appeared about background theory and view of developed probe newly.

• 화약ㆍ발파
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• 제17권4호
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• pp.18-31
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• 1999

The explosive forming has been used for many year to expand tubes into tubesheets. this process has demonstrated ability to direct carefully the energy of an explosive to expand tubes into tubesheet holes without damaging the tubesheet and without causing the excessive cold work at the tube I.D. that is normally associated with mechanical expansion. The success of explosive tube expansion provided the background for the development of the explosive tube plug. The main results are as follows : (1) The optimum explosives and explosive qualities are PETN, RDX, HMS and about 18~31gr/ft of explosive plugging in nuclear steam generator. (2) Explosive plugging's thickness is 0.9~1.8mm. If groove of 0.4 mm formed in plug outside, For the hydraulic leakage is go up, explosive plugging of formed groove are applicate tube and tubrplate. (3) Sheath is designed on the polyethylene of low density, In thermal impact test of the $430^\circ{C}$, hydraulic leakage is $300kg/cm^2$. (4) About 10~60mm oxide inclusions are existed on the space of explosive plug and tube protect to the leakage.

• 대한기계학회논문집A
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• 제30권10호
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• pp.1255-1260
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• 2006

To maintain the structural integrity of steam generator tubes, 40% of wall thickness plugging criterion has been developed. The approach is for the steam generator tube with single crack, so that the interaction effect of multiple cracks can not be considered. Although, recently, several approaches have been proposed to assess the integrity of steam generator tube with two identical cracks whilst actual multiple cracks reveal more complex shape. In this paper, the coalescence pressure of steam generator tube containing multiple cracks of different length is evaluated based on the detailed 3-dimensional (3-D) elastic-plastic finite element (FE) analyses. In terms of the crack shape, two collinear axial through-wall cracks with different length were considered. Furthermore, the resulting FE coalescence pressures are compared with FE coalescence pressures and experimental results for two identical collinear axial through-wall cracks to quantify the effect of crack length ratio on failure behavior of steam generator tube with multiple cracks. Finally, based on 3-D FE results, the coalescence evaluation diagrams were proposed.

• 대한기계학회논문집A
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• 제28권11호
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• pp.1813-1821
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• 2004

Until now, the 40% of wall thickness criterion, which is generally used for the plugging of steam generator tubes, has been applied only to a single cracked geometry. In the previous study by the authors, a total number of 9 local failure prediction models were introduced to estimate the coalescence load of two collinear through-wall cracks and, then, the reaction force model and plastic zone contact model were selected as the optimum ones. The objective of this study is to estimate the coalescence load of two collinear through-wall cracks in steam generator tube by using the optimum local failure prediction models. In order to investigate the applicability of the optimum local failure prediction models, a series of plastic collapse tests and corresponding finite element analyses for two collinear through-wall cracks in steam generator tube were carried out. Thereby, the applicability of the optimum local failure prediction models was verified and, finally, a coalescence evaluation diagram which can be used to determine whether the adjacent cracks detected by NDE coalesce or not has been developed.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1996년도 한국자동제어학술회의논문집(국내학술편); 포항공과대학교, 포항; 24-26 Oct. 1996
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• pp.101-104
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• 1996

In nuclear power plants, workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from the radiation exposure. In this paper, it is proposed a new approach to the development of the automatic vision system to examine and repair the steam generator tubes at remote distance. Digital signal processors are used in implementing real time recognition and examination of steam generator holes in the proposed vision system. Performance of proposed digital vision system is illustrated by experiment for similar steam generator model.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 1996년도 춘계학술대회 논문집
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• pp.200-204
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• 1996

It is a great number of problem for the man to perform maintenance and repairing work owing to radioactive effusion for a nuclear fuel and the pollution of an related equipment in nuclear power plants. Therefore, the vision processing system presented in this research requires to maintain the good performance under the radioactive circumstances and to safety the real time processing system presented in this research requires to maintain the good performance under the radioactive circumstances and to safety the real time processing. The proposed vision scheme adapts the gradient and Laplacian operator to perform the high speed processing in an edge detection and the centroid formula at each direction to obtain the center position of a holes using DSPs

• Nuclear Engineering and Technology
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• 제27권3호
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• pp.327-335
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• 1995

원자로의 가동중지 중이나. 재장전시 중기 발생기의 세관검사 및 보수작업을 병행하면 원전의 운전정지보수기 간을 현저하게 단축할 수 있다. 이때 원자로가 설치되어 있는 수조의 냉가수가 중기발생기내로 유입되는 것을 막는 장비로써 노즐댐이 있다. 노즐댐의 설치는 고방사선환경과 제한된 공간에서 작업을 해야 하는 특수성 때문에 작업자들이 기피하는 현상을 보인다. 현재 쓰이고 있는 무거운 노즐댐은 노즐댐설치 및 제거작업에 장애가 되는 가장 큰 요인이다. 본 논문에서는 노즐댐의 재질선정과 구조설계를 병행하여 현재 쓰이고 있는 노즐댐보다 가벼우면서도 굽힘강성 대 무게비와 비 강도가 증가된 노즐댐을 설계하였으며, 탄소섬유강화 복합재료로 경량노즐댐을 제작 완료하였다.

• Tribology and Lubricants
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• 제21권1호
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• pp.33-38
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• 2005

Fretting is the oscillatory motion with very small amplitudes, which usually occurs between two solid surfaces in contact. Fretting wear is the removal of material from contacting surfaces through fretting action. Fretting wear of steam generator tubes in nuclear power plant becomes a serious problem in recent years. The materials for the tubes usually are Inconel 690 (I-690) and Inconel 600 (I-600). In this paper, fretting wear tests for I-690 and I-600 were performed under various applied loads in water at room temperature. Results showed that the fretting wear loss of I-690 and I-600 tubes was largely influenced by stick-slip. The fretting wear mechanisms were the abrasive wear in slip regime and the delamination wear in stick regime. Also, I-690 had somewhat better wear resistance than I-600.

• Tribology and Lubricants
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• 제24권6호
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• pp.362-367
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• 2008

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. The tube and support materials were Inconel 690 and STS 409. The wear tests were conducted in various environments, which are in water without flow, in flowing water and in flowing water with air. The results showed that the flow of water influenced on the wear-life of tube. The wear-life of tube decreased in water flow as compared with wear-life in stationary water.