• Corrosion Science and Technology
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• v.22 no.3
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• pp.214-219
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• 2023

Considering the case in the United States where most nuclear power plants with an initial design life of 40 years continue to operate until 60 or 80 years after undergoing material soundness evaluation, it is time to plan a more robust long-term operation strategy for nuclear power plants in Korea. There are some reports that SRO/LRO might be formed when Alloy 690 is heat treated for 10,000 hours to 100,000 hours at 360 to 450 ℃. The possibility of LRO formation in Alloy 690 steam generator tubings of Kori nuclear power plant unit 1 (Kori-1) was investigated using existing research papers. The mechanism in which SRO/LRO occurred was also surveyed. Alloy 690 was found to be more likely to cause ordering than Alloy 600 in terms of alloy composition. The ordering could be evaluated through changes in material properties. However, it is difficult to evaluate it from a microstructural point of view. The likelihood of LRO in Alloy 690 of the Kori-1 plant operated at 320 ℃ for 19 years seemed to be low in terms of time and exposure temperature.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.95-95
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.94-94
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.211.1-211.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.947-948
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• 2005

• Proceedings of the KSME Conference
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• 2001.06e
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• pp.463-468
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• 2001

This paper presents the methodology for thermal hydraulic analysis of Pressurized Water Reactor (PWR) steam generators. Topics include porous media approach, governing equations, physical models and correlations for solid-to-fluid interaction and heat transfer and numerical solution scheme. Some details about the ATHOS3 code currently used widely for thermal hydraulic analysis of PWR steam generators in the industry are presented. The ATHOS3 code is applied to the thermal hydraulic analysis of steam generator in the Korea YGN 3&4 nuclear power plant and the computed results are presented.

• Journal of Energy Engineering
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• v.6 no.2
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• pp.170-175
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• 1997

An experimental study was carried out with the Mock-up made for the improved water level control of the steam generator in nuclear power plant and for the confirmation of swell/shrink status of the water level by opening and closing the steam dump valve. We can confirmed the possibility of using the Mock-up by introducing the PI controller and the FUZZY controller. Accordingly, we can confirmed that the practical usability of advanced controllers, which will be developed for the improved water level control of the steam generator in nuclear power plant by using the Mock-up.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2007.06a
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• pp.553-554
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• 2007

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.312-318
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• 1996

Flow induced vibration experiments were done to determine the vibration magnitude of tubes in the economizer tube lane region of a steam generator. The objective of this experiment is to demonstrate that the tubes are not susceptible to harmful levels of vibration at 100% of full power flow and to quantify the remaining design margin at 120% and 140% of full power flow.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3892-3901
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• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.