• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.465-476
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• 2000

In a nuclear power plant, steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus very conservative approaches have been taken in the light of steam generator tube integrity According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever causes are. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about twenty years ago when wear and pitting were dominant causes for steam generator tube degradation. And it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram.

• Journal of Magnetics
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• v.15 no.2
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• pp.70-73
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• 2010

A magnetic phase is partly produced in a steam generator tube due to stress and heat, because steam generator tubes are exposed to high temperature, high pressure and radioactivity conditions. This adversely affects the safety of steam generator tubes. However, it is difficult to detect it using conventional eddy current methods. Therefore, a new type of probe is needed to separate the signals from the defects and magnetic phases. In this study, a new U-type yoke, which contained two types of coils, a magnetizing coil and detecting coil, was designed. In addition, the signal induced by the magnetic phase and defect in an Inconel 600 plate were simulated.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2002.11b
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• pp.927-932
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• 2002

The stability ratio and vibrational amplitude of each tube inside a steam generator have different values. We estimate the characteristics of flow-induced vibration for CE type steam generator with various column and row number of the tube. To obtain the thermal-hydraulic data and stability ratio we use the ATHOS3-MODI and PIAT-FEI/TE code respectively. It turns out that the steam generator has a bounded central zone with the distributed values of the stability ratio and the vibrational amplitude, and those values across the zone boundary become decreased.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2002.11a
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• pp.390.2-390
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• 2002

The stability ratio and vibrational amplitude of each tube inside a steam generator have different values. We estimate the characteristics of flow-induced vibration for CE type steam generator with various column and row number of the tube. To obtain the thermal-hydraulic data and stability ratio we use the ATHOS3 and PIAT-FEI/TE code respectively. It turns out that the steam generator has a bounded central zone which the distributed values of the stability ratio and the vibrational amplitude, and those values across the zone boundary become decreased.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.11a
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• pp.84-91
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• 2003

Flow-Induced Vibration of steam generator tubes may result in fretting wear damage at the tube-to-support locations. KSNP(Korean Standard Nuclear Power plant) steam generators experienced fretting wear in the upper part of U-bend above the central cavity region of steam generators. This region has conditions susceptible to the flow-induced vibration, such as high flow velocity, high void fraction, and longer unsupported span. To improve its performance, APR1400 steam generator is designed with additional supports in this region to reduce unsupported span and to reduce peak velocity in the central cavity region. In this paper, we examined its performance improvement using ATHOS code. The thermal-hydraulic condition in the region of secondary side of APR1400 steam generator is obtained using the ATHOS3 code. The effective mass for modal analysis is calculated using the void fraction, enthalpy, and operating pressure information from ATHOS3 code result. With the effective mass distribution along the tube, natural frequency and mode shape is obtained using ANSYS code. Finally, stability ratios and real mean squared displacements for selected tubes of the APR1400 steam generator are computed. From these results, the current design of the APR1400 steam generator are examined.

• Nuclear Engineering and Technology
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• v.29 no.1
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• pp.15-24
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• 1997

The design and evaluation of a digital U-tube steam generator level controller of nuclear power plants, which uses model-based compensators to offset the inverse response behavior of water level, is described. Included is a review of steam generator level dynamics, a simulation model that replicates the effects of feedwater and steam flowrate as well as temperature on steam generator level, the design of both the compensators and the overall controller, and the results of simulation studies in which the performances of this model-based controller and existing analog ones were compared. The proposed digital steam generator level controller is stable and its use significantly improves the controllability of steam generator level.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2002.11a
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• pp.390.1-390
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• 2002

In this study, we investigate the plugging effect on the CE type steam generator tube. The natural frequency and mode shape will be changed due to decrease of the effective mass distribution along the tube. We compared the variation of stability ratio for plugged tube with that fur unplugged one. The natural frequency increased because of removing the cooling water inside the steam generator tube, but the stability ratio decreased inversely because of changing the vibrational mode shape. We also investigated the turbulent excitation effect.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.29-33
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• 2008

As the measures of nuclear power plant utilities and manufacturers to reduce the defects of tube expansion region during manufacturing steam generators, many types of NDEs(Non-Destructive Examinations) are conducted to inspect the expansion region. The expansion region of tube is subject to degrade because of stress concentration induced by tube expansion, sludge pile and high temperature. So the inspections for tube expansion region have been reinforced. Liquid penetrant test, helium leak test, Bobbin profile test and hydraulic test are performed to confirm the integrity of tube expanded by hydraulic expansion method. Liquid penetrant test and helium leak test are used to inspect seal weld region on tubesheet end part. Bobbin Profile test is used to inspect fully the expanded region of steam generator tube. Hydraulic test finally verifies the integrity of seal weld region on tubesheet end part.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.828-833
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• 2001

In this study, an analysis tool to assess the susceptibility of steam generator tubes due to the flow-induced vibration was developed. The fluid-elastic instability analysis of the U-tube bundle for CE-type steam generator was accomplished. The effective mass distribution along the U-tube was obtained to calculate the natural frequency and dynamic mode shape. Finally, stability ratios for selected tubes are obtained.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2001.11b
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• pp.1400-1405
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• 2001

In this study, an analysis program to assess the susceptibility of steam generator tubes due to the flow-induced vibration was developed. Analysis of fluid-elastic instability and random turbulence excitation for the U-tube bundle in CE-type steam generator was accomplished. The effective mass distribution along the U-tube was obtained to calculate the natural frequency and dynamic mode shape. Finally, stability ratios and rms vibration amplitude for selected tubes are obtained.