• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1997년도 한국자동제어학술회의논문집; 한국전력공사 서울연수원; 17-18 Oct. 1997
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• pp.1367-1370
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• 1997

This paper presents a vision algorithm for finding the centers of steam generator tubes using the generalized symmetry transform, which is used for ECT(Eddy Current Test) of steam generator tubes in nuclear power plants. The geometrical properties of the image representing steam generator tubes shows that they have amost circular or somewhat elliptic appearances and each tube has strong symmetry about its center. So we apply the generalized symmetry transform to finding centers of steam geneator tubes. But applying the generalized symmetry transform itself without any modification gives difficulties in obtaining the exact centers of steam generator tubes. But applying the generalized symmetry transform itself without any modification gives difficulties in obtaining the exact centers of tubes due to the shadow effect generated by the local light installed inside steam generator. Therefore we make the generalized symmetry transform modified, which uses a modified phase weight function in getting the symmetry magnitude in order to overcome the misleading effect by the local light. The experimental results indicate that the proposed vision algorithm efficiently recongnizes centers of steam generator tubes.

• International Journal of Safety
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• 제11권1호
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• pp.1-5
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• 2012

It has been reported that the plugged steam generator tube of Three Mile Island Unit 1 in America was damaged by growing flaw and then this steam generator tube destroyed the nearby steam generator tubes of normal state. On this account, stabilizer installation is necessary to prevent secondary damage of the steam generator tubes. The flow-induced vibration is one of the major causes of the fluid-elastic instability. To guarantee the structural integrity of steam generator tubes, the flow-induced vibration caused by the fluid-elastic instability is necessary to be suppressed. In this paper, the effective velocity and the critical velocity are calculated to evaluate the fluid-elastic instability. In addition, stability ratio value of the steam generator tubes is evaluated in order to propose one criterion when to determine stabilizer installation.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.250-255
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• 2007

Steam generator tubes play an important role in safety because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear power plant. For this reason, the integrity of the tubes is essential in minimizing the leakage possibility of radioactive water. The integrity of the tubes is evaluated based on NDE (non-destructive evaluation) inspection results. Especially ECT (eddy current test) method is usually used for detecting the flaws in steam generator tubes. However, detection capacity of the NDE is not perfect and all of the "real flaws" which actually existing in steam generator tunes is not known by NDE results. Therefore reliability of NDE system is one of the essential parts in assessing the integrity of steam generators. In this study POD (probability of detection) of ECT system for ODSCC in steam generator tubes is evaluated using multivariate logistic regression. The cracked tube specimens are made using the withdrawn steam generator tubes. Therefore the cracks are not artificial but real. Using the multivariate logistic regression method, continuous POD surfaces are evaluated from hit (detection) and miss (no detection) binary data obtained from destructive and non-destructive evaluation of the cracked tubes. Length and depth of cracks are considered in multivariate logistic regression and their effects on detection capacity are evaluated.

• 대한전자공학회:학술대회논문집
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• 대한전자공학회 2002년도 하계종합학술대회 논문집(4)
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• pp.263-266
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• 2002

This paper describes extraction procedure for the center coordinates of steam generator tubes of Youngkwang NPP #6, which are arrayed in triangular patterns. Steam generator tube images taken with wide field-of-view lens and low-light lamp mounted on a ccd camera tend to have low contrast, because steam generator is sealed and poorly illuminated. The extraction procedures consists of two steps. The first step is to process the region with superior contrast in entire image of steam generator tubes and to extract the center points. Using the extracted coordinates in the first step and the geometrical array characteristics of tubes lined up in regular triangle forms, the central points of the rest region with low contrast are estimated. The straight lines from center point of a tube to neighbour points in horizontal and 60, 120$^{\circ}$ degree directions are derived. The intersections of straight line In horizontal direction and slant line in regular triangle direction are selected as the center coordinates of steam generator tubes. The Chi-square interpolation method is used to determine the line's coefficients in horizontal and regular triangle direction.

• 대한기계학회논문집A
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• 제30권4호
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• pp.420-427
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• 2006

The structural and leakage integrity of steam generator tubes should be sustained against all postulated loads even if a crack is present. During the past three decades, most of the efforts with respect to integrity evaluation of steam generator tubes have been focused on limit load solutions but, recently, the applicability of elastic-plastic fracture mechanics was examined cautiously due to its effectiveness. The purpose of this paper is to introduce a testing method to estimate fracture resistance characteristics of steam generator tubes with a through-wall crack. Due to limited thickness and diameter, inevitably, the steam generator tubes themselves were tested instead of standard specimen or alternative ones. Also, a series of three dimensional elastic-plastic finite element analyses were carried out to derive closed-form estimation equations with respect to J-integral and crack extension for direct current potential drop method. Since the effectiveness of $J_{IC}$ as well as J-R curves was proven through comparison with those of standard specimens taken from pipes, it is believed that the proposed scheme can be utilized as an efficient tool for integrity evaluation of cracked steam generator tubes.

• Nuclear Engineering and Technology
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• 제32권5호
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• pp.465-476
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• 2000

In a nuclear power plant, steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus very conservative approaches have been taken in the light of steam generator tube integrity According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever causes are. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about twenty years ago when wear and pitting were dominant causes for steam generator tube degradation. And it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram.

• Corrosion Science and Technology
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• 제20권1호
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• pp.22-25
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• 2021

In the past, Alloy 600 nickel-based alloys have been widely used in steam generators. However, most of them have been replaced by thermally treated alloy 690 tubes in recent years because mill annealed alloy 600 materials are known to be susceptible to stress corrosion cracking. Unlike this general perception, some steam generators using mill annealed alloy 600 tubes show excellent performance even though they are designed, manufactured, and operated in the same way. Therefore, various analyses were carried out to determine causes for the degradation of steam generators. Based on the general stress corrosion cracking mechanism, tube material susceptibility, residual stress, and sludge deposits of steam generators were compared to identify factors affecting stress corrosion cracking. It was found that mill annealed alloy 600 steam generator tubes showed higher resistance to stress corrosion cracking when the amount of sludge deposits on tube surface was smaller and residual stress generated during the fabrication was lower.

• 대한기계학회논문집A
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• 제28권6호
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• pp.732-738
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• 2004

One of the main degradation mechanisms in steam generator tubes is stress corrosion cracking induced by residual stress. The resulting damages can cause tube bursting or leakage of the primary water which contains radioactivity. Shot peening technique has been used to prevent stress corrosion crack growth in steam generator tubes. In order to investigate the shot peening effect on stress corrosion cracking stress intensity factors are calculated for the semi-elliptical surface crack which is located in residual stress region. The residual stress distribution in steam generator tubes is obtained from the simple model proposed by Frederick et al.

• 비파괴검사학회지
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• 제32권5호
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• pp.545-550
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• 2012

원자력발전소 증기발생기 세관은 방사성물질이 외부로 누출되지 않도록 압력경계 역할을 하는 주요부품이다. 설비 운전기간이 증가함에 따라 이차측에서 유입된 슬러지가 증기발생기 2차측 유체 흐름을 따라 상부로 이동하면서 유체비등과 난류에 의해 세관 외면에 스케일이 부착되어 세관열화, 유로홈 막힘 및 열전달을 감소시키는 파울링을 유발하는 원인으로 작용한다. 따라서, 원전 운영자는 세관 외면에 쌓인 스케일의 두께를 확인하여 일정시점이 되면 화학세정 등의 정비를 수행한다. 본 논문에서는 보빈코일 와전류신호를 이용하여 세관 외면에 부착된 스케일 두께를 정량적으로 평가하는 기술을 개발하고자 스케일 시험편을 제작하여 스케일 두께와 와전류신호 진폭 간의 상관관계를 분석하였고, 이를 바탕으로 스케일의 두께를 정량적으로 평가하는 기법과 대량의 와전류 데이터를 평가할 수 있는 프로그램을 개발하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.1684-1689
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• 2007

Fretting wear is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Especially, impact fretting wear occurred between steam generator tubes and tube support plates or anti-vibration bar. Various tests have been carried out to investigate the wear mechanisms and to report the wear coefficients. Those are fruitful to get insight for the wear damage of steam generator tubes; however, most wear researches have concentrated on sliding wear of the steam generator tubes, which may not represent the wear loading modes in real plants. In the present work, impact fretting tests of steam generator tube were carried out. A wear progression model for impact-fretting wear has been investigated and proposed. The proposed wear progression model of impact-fretting wear is as follows; oxide film breaking step at the initial stage, and layer formation step, energy accumulation step and finally particle torn out step which is followed by layer formation in the stable impact-fretting progress. The wear coefficient according to the work-rate model has been also compared with one between tube and support.