• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.4 s.247
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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.

• Proceedings of the KSME Conference
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• 2007.05a
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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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.29 no.8 s.239
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• pp.1070-1077
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• 2005

The $40\%$ of wall thickness criterion which has been used as a plugging rule is applicable only to a single cracked steam generator tubes. In the previous studies performed by authors, several failure prediction models were introduced to estimate the plastic collapse pressures of steam generator tubes containing collinear or parallel two adjacent axial through-wall cracks. The objective of this study is to examine the failure prediction models and propose optimum ones for non-aligned two axial through-wall cracks in steam generator tubes. In order to determine the optimum ones, a series of plastic collapse tests and finite element analyses were carried out for steam generator tubes with two machined non-aligned axial through-wall cracks. Thereby, either the plastic zone contact model or COD based model was selected as the optimum one according to axial distance between two clacks. Finally, the optimum failure prediction model was used to demonstrate the conservatism of flaw characterization rules for various multiple flaws according to ASME code.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.12 s.255
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• pp.1618-1625
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• 2006

During the last two decades, several guidelines have been developed and used for assessing the integrity of a defective steam generator (SG) tube that is generally caused by stress corrosion cracking or wall-thinning phenomenon. However, as some of SG tubes are also failed due to fretting and so on, alternative failure estimation schemes are required for relevant defects. In this paper, parametric three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of SG tubes with different defect configurations; elliptical wear, tapered and flat wear type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of SG tube. After investigating the effect of key parameters such as defect depth, defect length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wear region. Comparison of failure pressures predicted by the proposed estimation scheme with corresponding burst test data showed a good agreement.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.32 no.10
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• pp.817-822
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• 2008

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 progress model for impact-fretting wear has been investigated and proposed. The proposed wear progress 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.

• Journal of the Korean Society for Precision Engineering
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• v.22 no.10 s.175
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• pp.137-143
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• 2005

The integrity of steam generator tubes in nuclear power plant should be maintained sufficiently during operation. For sake of this, complicated assessment procedures are required such as fracture mechanics analysis, etc. The integrity assessment of tubes has been performed by using conventional deterministic approaches while there are many uncertainties to carry out a rational evaluation. In this respect, probabilistic integrity assessment is considered as an alternative method for integrity assessment. The objectives of this study are to develop an integrity assessment system based on probabilistic fracture mechanics and to predict the failure probability of steam generator tubes containing an axial through-wall crack. The developed integrity assessment system consists of three evaluation modules, which apply first order reliability method, second order reliability method and Monte Carlo simulation method, respectively. The system has been applied to predict failure probability of steam generator tubes and the estimation results showed a promising applicability of the probabilistic integrity assessment system.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.801-805
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• 1998

1000 MWe 국내 표준형 원전을 대상으로 노심이 손상되는 각종 중대사고 시나리오에 대하여 방사선원항 특성을 평가하기 위하여, 2단계 확률론적 안전성 평가 방법론에 따라 방사선원 방출군을 정의하고 원전 중대사고 발생시 격납건물 손상을 가정하여 각 방출군별로 격납건물 외부로 방출되는 방사능 방출율을 정량화하였다. 도출된 19개의 그룹중에서 방출률이 작거나 발생빈도가 낮은 7개를 제외하고 12가지 대표 사고경위에 대하여 계산을 수행하였으며, 분석결과는 격납건물 내에서 감쇄효과가 작은 증기발생기 세관 파단사고, 격납건물 격리 실패사고 및 조기 격납건물 파손사고 둥이 상대적으로 큰 방사능 방출량을 보여주었다

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.8
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• pp.951-956
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• 2011

Inconel alloys are generally used as steam generator tubes in nuclear power plants. These alloys are highnickel chromium alloys that exhibit excellent resistance to aqueous corrosion. In this paper, the effects of elevated temperatures such as an operating temperature of $320^{\circ}C$ on the fretting fatigue behavior of inconel 600 and 690. We observed that the plain and fretting fatigue limits at $320^{\circ}C$ were slightly lower than those at room temperature. The frictional forces varied depending on the number of load cycles. After each test, we studied the fretting fatigue mechanisms via SEM observations. These results can be used for structural integrity evaluations at elevated temperatures and for studying fretting damage in steam generator systems.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.11
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• pp.1741-1751
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• 2004

It is commonly requested that the steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criterion is known to be too conservative for some locations and types of defects and its application is limited to a single crack in spite of the fact that the occurrence of multiple through-wall cracks is more common in general. The objective of this research is to propose the optimum failure prediction models for two adjacent through-wall cracks in steam generator tubes. The conservatism of the present plugging criteria was reviewed using the existing failure prediction models for a single crack, and six new failure prediction models for multiple through-wall cracks have been introduced. Then, in order to determine the optimum ones among these new local or global failure prediction models, a series of plastic collapse tests and corresponding finite element analyses for two adjacent through-wall cracks in thin plate were carried out. Thereby, the reaction force model, plastic zone contact model and COD (Crack-Opening Displacement) base model were selected as the optimum ones for assessment of steam generator tubes with multiple through-wall cracks. The selected optimum failure prediction models, finally, were used to estimate the coalescence pressure of two adjacent through-wall cracks in steam generator tubes.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.25-31
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• 2009

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which comprises of the pressure boundary of primary system. The integrity of SG tube has been confirmed by the eddy current test every outage. In Korea, Bobbin probe and MRPC probe have been generally used for the eddy current test. Meanwhile the usage of Array probe has gradually increased in U.S., Japan and other countries. In this study, we investigated the defect detection capability of the Array probe through its preliminary application to SG tube inspection. The Array probe has the equivalent capability in the defect detection and sizing as the conventional methods. Thus it is desirable that the Array probe is generally applied to SG tube inspection in the domestic NPPs.