• Journal of the Korean Society for Nondestructive Testing
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• v.34 no.5
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• pp.356-361
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• 2014

The eddy current testing performance directly affects the results of a steam generator tube integrity assessment because the integrity assessment of defected tubes is conducted based on eddy current testing results. This means that it may not be possible to accurately discriminate between adjacent flaws. This paper presents an investigation on the resolution of rotating pancake coils with multiple cracks and the effects on the structural integrity assessment of steam generator tubes.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.176-181
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• 1996

원자로 용기 헤드부위의 관통관은 재질이 Inconel-600이며, 현재 세계각국에서도 원자로 헤드 관통관의 균열이 일부 발견되어 우리나라에서도 관심이 되고 있다. 국내 원전 헤드관 통관 수량도 고리 1,2호기의 경우 40개, 고리3,4호기(영광1,2) 61개, 울진 57개로서 관통관의 균열결함이 존재할 수 있다. 만약 균열이 성장하여 파손 되었을 시 원자로 냉각재 누설등 발전소 안전에 큰영향을 미치므로 균열의 원인으로 알려진 용접부위 잔류응력 및 발전소 정상운전 상태에서의 응력을 해석하였다.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 2002.11a
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• pp.13-18
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• 2002

증기 발생기 전열관의 파열 사고는 지난 20년 동안 2년마다 1개씩의 비율로 발생되어왔고 최근 몇 년간은 매년 발생되고 있는 추세이다(3). 전열관의 파열 사고는 응력부식균열, 피로 그리고 마멸 등의 원인에 의해서 발생되고 있는 것으로 알려져 있다. 초기 발전소에서 균열의 발생 및 성장은 축 방향 균열에 국한하여 관심을 가졌었으나 최근 원주 방향 균열에 의한 사고가 발생되면서 원주 방향 균열에 대해 관심을 가지게 되었다.(중략)

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.2
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• pp.98-103
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• 2009

The nuclear power plants have been safely operated by plugging the steam generator tubes which have the crack indications. Tube rupture events can occur if analysts fail to detect crack indications during in-service inspection. There are various types of crack indication in steam generator tubes and they have been detected by the eddy current test. The integrity assessment should be performed using the crack-sizing results from eddy current data when the crack indication is detected. However, it is not easy to evaluate the crack-depth precisely and consistently due to the complexity of the methods. The current crack-sizing methods were reviewed in this paper and the suitable ones were selected through the laboratory tests. The retired steam generators of Kori Unit 1 were used for this study. The round robin tests by the domestic qualified analysts were carried out and the statistical models were introduced to establish the appropriate depth-sizing techniques. It is expected that the proposed techniques in this study can be utilized in the Steam Generator Management Program.

• Journal of Energy Engineering
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• v.23 no.2
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• pp.149-157
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• 2014

The steam generator tubes of nuclear power plants have various types of corrosion failures during the plant operation. The stress corrosion cracking which occurs on the outer surface of tube is called the secondary side stress corrosion cracking and mainly occurs in the expansion-transition area of tube. The causes are the concentration of impurities by the sludge pile-up related to the geometry of its region and the residual stress by tube expansion in the process of steam generator manufacturing. Especially the directionality and sizes of residual stresses are differed according to the tube expansion methods and the direction and the frequency of tube cracks depend on their characteristics. In bases on the plant experiences, it is notified that circumferential cracks of tubes expanded with explosive expansion method are dominantly occurred compared to those of tubes done with hydraulic expansion one. Therefore in this study, according to tube expansion methods frequencies and sizes of tube cracks with specific direction are compared by means of accelerated immersion test and also the crack morphology and the specific chemicals from water-chemistry environment are observed through the fracture surface examination.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.765-770
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• 1995

지체균열전파(DHC)는 중수로 압력관의 수명에 근 영향을 미치는 중요한 현상 중의 하나이다. 본 연구에서는 열처리를 통하여 압력관 재료인 Zr-2,5Nb의 기계적 성질, 집합조직을 변화시켜 각 인자들이 DHC에 미치는 영향을 조사하였다. 그 결과 지체균열전파속도(DHCV)는 항복강도와 경도와 비례한다는 것과 유사한 미세구조와 집합조직을 갖는 Zr-2.5Nb의 경우 항복강도와 Puls의 모델을 이용하여 지체균열전파속도(DHCV)를 예측할 수 있었다. 그리고 secondary cracking이 annealing한 시편들에서는 관찰되었으나 $\beta$열처리 후 급냉한 시편에서는 관찰되지 않았다. 이것의 수소화물 형상의 차이에 의한 것으로 생각된다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11b
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• pp.481-486
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• 1996

원자력 발전소의 증기 발생기 전열관으로 사용되는 합금 600MA는 미세 구조에 관계없이 가성 용액에서 입계 균열이 발생한다. 본 연구에서는 합금 600 2종과 합금 690 2종의 이음매없는 관 및 진공 용해한 합금 690M 2종의 MA 및 TT재에 대한 부식 조건의 변화에 응력 부식 균열 특성을 일정연신율법(CERT) 및 C-ring법으로 평가하였다. 가성 응력 부식 균열 저항성에 미치는 TT처리의 효과는 용액 조건에 관계없이 TT처리를 행하게 되면 응력 부식 균열에 대한 저항성이 증가하는 것으로 나타났으며, 분극 저항성과는 직접적인 관계가 나타나지 않고 다른 미세 조직 등에 의한 영향을 더 크게 받고 있는 것으로 판단된다. 가성 용액에서의 응력 부식 균열 저항성에 미치는 SO$_4$$^{=}$ 이온의 첨가 효과는 TT처리의 유무에 관계없이 응력 부식 균열 저항성을 크게 감소시키고 있다. 한편 합금 690의 가성 응력부식 균열 저항성에 미치는 Mo의 효과는 Mo이 첨가될수록 응력 부식 균열 저항성이 증가하는 것으로 나타났다.

• Journal of Energy Engineering
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• v.21 no.4
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• pp.362-372
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• 2012

The steam generator tubes of nuclear power plants are pressure boundaries, and if tubes are leaked, the coolant with the radioactive materials was flowed out from the primary system to the secondary system and polluted the plant and the air. Recently most crack defects of tubes are stress corrosion cracks and these defects are located in expansion transition area, sludge pile-up region, and U-bend area. The most effective one of crack initiation factors in expansion transition area and U-bend area is the residual stress. According to the experiences of Korea standard nuclear plants(Optimized Power Reactor-1000), they had the stress corrosion cracks at the tube expansion transition area in early operating stage and especially lots of circumferential cracks were occurred. Therefore in this study, the distributions and conditions of residual stresses by tube expansion methods were compared and the dominant reason of a specific direction was examined.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.5
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• pp.501-509
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• 2001

For the enhancement of ECT reliability on the primary water stress corrosion cracks of nuclear steam generator tubes, of which the occurrence is on the increase, it is important to comprehend the signal characteristics on crack morphology and to select an appropriate probe type. In this paper, the sizing accuracy and the detectability for the inner wall axial cracks of tubes were quantitatively evaluated using the following specimens: the electric discharge machined notches and the corrosion cracks which were developed on the operating steam generator tubes. The difference of eddy current signal characteristics between pancake and axial coil were also Investigated. The results obtained from this study provide a useful information for more precise evaluation on the inner wall axial tracks oi stram generator tubes.

• Journal of Energy Engineering
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• v.22 no.2
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• pp.148-155
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• 2013

The steam generators of OPR-1000 plants have Alloy 600 and Alloy 690 as the tube material and its tube expansion method is the explosive expansion method. According to the experience of these plants, circumferential cracks were largely occurred in steam generator tubes expanded by the explosive expansion method and their locations were the outer surface of tube expansion transition region surrounding with piled-up sludge. But even though tubes have the same conditions, tubes with the hydraulic expansion method shows the prevail trend of axial cracks compared to circumferential cracks. Therefore in this study, in order to identify the difference of such phenomena as above, configurations of tube and tubesheet were modeled and at operating conditions, stress values applied in the tube expansion transition area in accordance with tube expansion methods were calculated by using computational program and the direction and the predominance of cracks were evaluated.