• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.219.2-219
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• 1999

• 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.

• 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.21 no.5
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• pp.493-500
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• 2001

The detection of axial cracks using conventional MFL pig is a significant challenge in the gas pipeline inspection. In this study, a technique using interaction of circumferentially induced torrents with axial stress corrosion crack is presented. The feasibility of this technique is investigated using finite element modeling. Finite element analysis of such interaction is a difficult problem in terms of both computation time and memory requirements. The challenges arise due to the nonlinearity of material properties, the small sire of tight cracks relative to that of the magnetizer, and also time stepping involved in modeling velocity effects. This paper presents an approach based on perturbation methods. The overall analysis procedure is divided into 4 simple steps that can be performed sequentially. Modeling results show that this technique can effectively detect colonies of SCC as well as single SCC.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.1
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• pp.47-54
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• 2013

This study investigates the crack growth behavior due to primary water stress corrosion cracking (PWSCC) in the dissimilar metal butt weld of a reactor piping using Alloy 82/182. First, detailed finite element stress analyses were performed to predict the stress distribution of the dissimilar metal butt weld in which the hydrostatic and the normal operating loads as well as the weld residual stresses were considered to evaluate the stress redistribution due to mechanical loadings. Based on the stress distributions along the wall thickness of the dissimilar metal butt weld, the crack growth behavior of the postulated axial and circumferential cracks were predicted, from which the crack growth diagram due to PWSCC was proposed. The present results can be applied to predict the crack growth rate in the dissimilar metal butt weld of reactor piping due to PWSCC.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.4
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• pp.374-382
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• 2009

Steam generator(SG) tube, as a barrier isolating primary to the secondary coolant system of nuclear power plants(NPP), must maintain the structural integrity far the public safety and its efficient power generation capacity. And SG tubes bearing defects must be timely detected and taken repair measures if needed. For the accomplishment of these objectives, SG tubes have been periodically examined by eddy current testing(ECT) on the basis of administrative notices and intensified SG management program(SGMP). Stress corrosion cracking(SCC) on the SG tubes is not easily detected and even missed since it has lower signal amplitude and other disturbing factors against its detection. However once SCC is developed, that can cause detrimental affects to the SG tubes due to its rapid propagation rate. Accordingly SCC is categorized as prime damage mechanism challenging the soundness of the SG tubes. In this study, reproduced EDM notch specimens are examined for the detectability and quantitative characterization of the axial ODSCC by +PT MRPC probe, containing pancake, +PT and shielded pancake coils apart in a single plane around the circumference. The results of this study are assumed to be applicable fur providing key information of engineering evaluation of SCC and improvement of confidence level of ECT on SG tubes.

• Journal of Korean Society of Steel Construction
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• v.12 no.3 s.46
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• pp.269-279
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• 2000

It is much expected that steel bridges, which have been damaged by increase of vehicle load and corrosion, need repair or strengthening. In this paper, the stress generated by repair welding under loading are analyzed by three dimensional elasto-plastic analyses. The longer and deeper repair weld line bocemes, the larger the magnitude of transient stress becomes. The magnitude of transient stress generated by repair welding under loading $({\sigma}_y/3,\;{\sigma}_a)$ is similar to summation of stresses generated by repair welding and loading. The longer repair weld line ratio(1/b) becomes, the larger the magnitude of transient stress generated by repair welding under loading bocomes. And, the longer repair weld line ratio(1/b) becomes, the larger the magnitude of in-plane displacement generated by repair welding under loading$({\sigma}_y/3,\;{\sigma}_a)$.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.83-88
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• 2010

Domestic nuclear steam generators with Alloy 600 HTMA tubes have experienced axial cracking at eggcrate tube support plates(TSPs). The axial stress corrosion cracks were observed at the crevice between outside of tubes and eggcrate TSPs. The root cause of axial cracking was investigated by thermal hydraulic analysis and sludge distribution diagnosis. It is suggested that deposition of sludge at eggcrate TSPs could increase the outside surface temperature of tube and promote the enrichment of impurities at crevice, and thus accelerate cracking. Additionally strategy for reducing the sludge ingress to steam generators is discussed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.10
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• pp.979-984
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• 2015

In the present paper, we evaluate the effects of the safe-end length and thickness of the similar metal weld (SMW) of safety injection nozzles on stress distributions at the dissimilar metal weld (DMW). For this evaluation, we carry out detailed 2-D axisymmetric finite element analyses by considering four different values of the safe-end length and four different values of the thickness of SMW. Based on the results obtained, we found that the SMW thickness affects the axial stresses at the center of the DMW for the shorter safe-end length; on the other hand, it does not affect the hoop stresses. In terms of the safe-end length, the values of the axial and hoop stresses at the inner surface of the DMW center increase as the safe-end length increases. In particular, for the cases considered in the present study, the stress distributions at the DMW center can be categorized according to certain values of safe-end length.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.10
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• pp.1337-1345
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• 2011

In pressurized water reactors (PWRs), the reactor pressure vessel (RPV) upper head contains numerous control rod drive mechanism (CRDM) nozzles. In the last 10 years, the incidences of cracking in alloy 600 CRDM nozzles and their associated welds has increased significantly. Several axial and circumferential cracks have been found in CRDM nozzles in European PWRs and U.S. nuclear power plants. These cracks are caused by primary water stress corrosion cracking (PWSCC) and have been shown to be driven by welding residual stresses and operational stresses in the weld region. Therefore, detailed finite-element (FE) simulations for the Korea Nuclear Reactor Pressure Vessel have been conducted in order to predict the magnitudes of the weld residual stresses in the tube materials. In particular, the weld residual stress results are compared in terms for nozzle location, geometry factor$r_o$/t, geometry of fillet, and adjacent nozzle.