• 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

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.118-118
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.292-292
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• 2003

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.11a
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• pp.144-145
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• 2006

• 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.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.05a
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• pp.670-676
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• 2003

To investigate the group plugging effect the fluid-elastic instability analysis has been performed on various column and row number of the KSNP steam generator lutes. This study compares the stability ratio of the plugged tube with that of the intact one. The information on the thermal-hydraulic data of the steam generator have been obtained by using the ATHOS3-MOD1 code with and without the thermal energy transfer at the plugged region. Both of the boundary conditions of hot-leg temperature and feedwater mass flow rate are fixed for this investigation. From the results of this study the stability ratios inside the group plugging zone are decreased slightly. At the outside of group plugging zone, however, most of the stability ratios tend to be increased.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.302-310
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• 2010

In this paper, we studied Bagging neural network for predicting defect size of steam generator(SG) tube in nuclear power plant. Bagging is a method for creating an ensemble of estimator based on bootstrap sampling. For predicting defect size of SG tube, we first generated eddy current testing signals for 4 defect patterns of SG tube with various widths and depths. Then, we constructed single neural network(SNN) and Bagging neural network(BNN) to estimate width and depth of each defect. The estimation performance of SNN and BNN were measured by means of peak error. According to our experiment result, average peak error of SNN and BNN for estimating defect depth were 0.117 and 0.089mm, respectively. Also, in the case of estimating defect width, average peak error of SNN and BNN were 0.494 and 0.306mm, respectively. This shows that the estimation performance of BNN is superior to that of SNN.

• Journal of computational fluids engineering
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• v.7 no.2
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• pp.43-52
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• 2002

This paper presents the numerical methodology of ATHOS3 code for thermal hydraulic analysis of steam generators in nuclear power plant. Topics include porous media approach, governing equations, physical models and correlations for solid-to-fluid interaction and heat transfer, and numerical solution scheme. The ATHOS3 code is applied to the thermal hydraulic analysis of steam generator in the Korea Kori Unit-1 nuclear power plant and the computed results are presented

• Nuclear industry
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• v.24 no.3 s.253
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• pp.16-25
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• 2004

• 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