• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.139.2-139
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• 2001

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.1
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• pp.8-14
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• 2004

Model-based inversion tools for eddy current signals have been developed by combining neural networks and finite element modeling, for quantitative flaw characterization in steam generator tubes. In the present work, interpretation of experimental eddy current signals was carried out in order to validate the developed inversion tools. A database was constructed using the synthetic flaw signals generated by the finite element model. The hybrid neural networks composed of a PNN classifier and BPNN size estimators were trained using the synthetic signals. Experimental eddy current signals were obtained from axisymmetric artificial flaws. Interpretation of flaw signals was conducted by feeding the experimental signals into the neural networks. The interpretation was excellent, which shows that the developed inversion tools would be applicable to the Interpretation of real eddy current signals.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.449-450
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.181.1-181.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.395.1-395
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• 2002

• Proceedings of the KIEE Conference
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• 2009.04b
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• pp.109-111
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• 2009

본 논문에서는 전자기 유한요소 해석을 이용하여 원전 증기발생기 세관에서의 결함 변화에 따른 배열와전류프로브의 와전류탐상 특성을 해석하였다. 프로브의 전자기적 특성을 해석하기 위하여 3차원 전자기유한요소법을 이용하였다. 해석 대상으로 FBH 결함이 있는 세관을 사용하였으며, 결함의 위치는 관의 외부표면에 존재하게 하고 결함의 깊이는 세관 두께의 20%, 40%, 60%, 80%, 100%로 하였다. 결함의 크기를 변화시켰으며, 시험주파수는 100kHz, 300kHz, 400kHz를 사용하였다. 배열와전류프로브의 방향성에 대한 특성을 확인하기 위하여 축방향 및 원주방향 Notch 결함 신호의 차이를 비교하였다. 본 논문을 통하여 결함형상, 깊이 및 크기, 시험주파수의 변화에 따른 탐상신호의 변화를 확인할 수 있었으며, 본 논문의 결과는 배열와전류프로브의 와전류탐상 신호 평가 시 도움이 될 것이다.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.1186-1191
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• 2003

The 40% of wall criterion, which is generally used for the plugging of steam generator tubes, may be applied only to a single crack. In the previous study, a total of 9 failure models were introduced to estimate the local failure of the ligament between cracks and the optimum coalescence model of multiple collinear cracks was determined among these models. It is, however, known that parallel axial cracks are more frequently detected during an in-service inspection than collinear axial cracks. The objective of this study is to determine the plastic collapse model which can be applied to the steam generator tube containing two parallel axial through-wall cracks. Nine previously proposed local failure models were selected as the candidates. Subsequently interaction effects between two adjacent cracks were evaluated to screen them. Plastic collapse tests for the plate with two parallel through-wall cracks and finite element analyses were performed for the determination of the optimum plastic collapse model. By comparing the test results with the prediction results obtained from the candidate models, a plastic zone contact model was selected as an optimum model.

• 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.

• 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.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.41-53
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• 1994

A SGTR accident postulated at Kori unit 1 is simulated with Mihama unit experience, which occurred on February 1991, to evaluate the capability of plant to cope with the transient. The system design and plant conditions of Kori Unit 1 are much similar with those of Mihama Unit 2. Therefore, special concern has been given to evaluate the sequence and the resulting consequence of the postulated SGTR accident at the Kori unit 1 An analysis is peformed as realistically as possible, with following the EOP of Kori unit 1. The result indicates that the leak through tube break terminates within about forty minutes, and the Kori unit 1 may be sufficient to cope with SGTR accident with same type of sequence. However, the reconsideration may be required for the design of Kori unit 1 which disconnects non-safety AC power from off-site power on SI signal generation. It may be pointed out that the content of EOP for SGTR accident is not enough to require operator's proper judgements. An analysis of SGTR accident tested in the LSTF which simulated the SGTR accident at the Mihama Unit 2 is peformed using the RELAP5/MOD3. The results indicates that the code yields in general good agreement with the test, except the break flowrate at the early stage of the event.