• 한국압력기기공학회 논문집
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• 제16권2호
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• pp.10-17
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• 2020

Securing the integrity of steam generator tubes is an essential requirement for safe operation of nuclear power plants. Therefore, tubes that do not satisfy integrity requirements are no longer usable and must be repaired according to the related requirements. In general, the repair criterion is that the damage depth is more than 40% of the tube wall thickness. However, the plugging limit can be changed and be applied, provided a technical proof is given that integrity can be secured against specific degradation at a specific plants and that approval can be obtained from a regulatory agency. A typical example is alternative repair criteria for defects within the tube sheet or tube support plates. In this paper, a background of establishing the plugging limit for steam generator tubes and changes in maintenance criteria are reviewed as examples.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.28-33
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• 2010

As a part of license renewal for the continued operation of Wolsong unit 1, the periodic safety review report was submitted near the end of design lifetime, 2012, and now is under reviewing. Major components of primary system such as pressure tubes, feeder pipes and so on are being replaced and many components of secondary system are also being repaired. So the license renewal of Wolsong unit 1 is expected to be acquired without significant issues. But on the other hand, steam generators of Wolsong unit 1 had the good performance and therefore the replacement and repair for the steam generator are not needed. Recently it is reported that some cracks were detected in a few of european steam generator with Alloy 800 tubes and the cause of cracks was the outer diameter stress corrosion cracking(ODSCC) due to the concentration of chemical impurities on the outer surface of tube. Accordingly the overall review on this issue was performed. The long-term operation is likely to results to form the concentration mechanism for the tube corrosion as the sludge build-up in the secondary side of steam generator and the crack in the crevice between tube and tube-sheet and expansion transitions is apt to be occurred. In this paper, the history of steam generator inspection and operation of Wolsong unit 1 are reviewed and the reliability of steam generator tube is evaluated and the steam generator aging management program for Wolsong unit 1 is introduced.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2013년도 춘계학술대회 논문집
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• pp.826-827
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• 2013

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.64-70
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• 2019

In recent years, attention has been paid to the integrity of steam generator (SG) tubes due to severe accident and beyond design basis accident conditions. In these transient conditions, steam generator tubes may be damaged by high temperature and pressure, which might result in a risk of fission products being released to the environment due to the failure. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690 using tube specimens. Based on manufacturer's creep data and creep test results performed in this study, creep life prediction was carried out using the Larson-Miller (LM) Parameter, Orr-Sherby-Dorn (OSD) parameter, Manson-Haford (MH) parameter, and Wilshire's approach. And a hyperbolic sine (sinh) function to determine master curves in LM, OSD and MH parameter methods was used for improving the creep life estimation of Alloy 690 material.

• 한국압력기기공학회 논문집
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• 제5권1호
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• pp.1-8
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• 2009

본 논문에서는 국제적으로 개발 중인 초고온가스로와 연계하여 대량의 수소를 생산하기 위한 방안의 일환으로 국내에서 개발 중인 고온증기전기분해 시스템에 사용될 열교환기 재료의 고온증기 부식실험에 대해 소개하였다. 이를 위해 관련된 국내외 연구현황을 조사분석한 결과를 요약하여 소개하였으며 마지막으로 현재 수행중인 고온증기부식 연구의 실험조건 및 계획을 제시하였다. 실험 및 연구결과는 초고온가스로와 연계된 고온전기분해를 이용한 수소생산시스템의 개발에 활용될 예정이다.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.141-148
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• 2016

This paper presents an overview of worldwide electric power development and National $700^{\circ}C$ Hyper Supercritical coal-fired power generation(HSC) focus on materials and manufacturing process. To Increase the efficiency of electric power generation, It is necessary to increase steam temperature and pressure. In that case, New material and manufacturing process shall be developed for boiler and turbine component in high temperature and pressure operating condition. Therefore, Much Efforts in worldwide are progressing to develop materials and manufacturing technology and to build and operate an HSC.

• 한국소음진동공학회논문집
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• 제20권9호
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• pp.828-833
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• 2010

Steam leakage is one of the major issues for the structural fracture of pipes of nuclear power plants. Therefore a method to inspect a large area of piping systems quickly and accurately is needed. In this paper, we proposed the method for the detecting steam leakage by using image signal processing. Our basic idea come from heat shimmer which shine with a soft light that looks as if it shakes slightly. To test the performance of this technique, experiments have been performed for simple heat source and steam generator. Results show that the proposed technique is quite powerful in the steam leak detection.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.20-24
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• 2014

Recently, it is getting more and more important ensuring the integrity for the equipment degradation according to the increase of nuclear power plant operating period. In many equipment of the nuclear power plant, snubbers mainly installed in reactor coolant pumps, steam generators and piping protected the equipment and piping from the occurrence of transient dynamic loads such as the earthquake, thermal load during the plant operation. This report describes the function, regulation, inspection requirements and management status of the snubbers installed in domestic nuclear power plants.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2012년도 추계학술대회 논문집
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• pp.470-475
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• 2012

The downstream piping system of the water supply system in a supercritical power plant is affected by the fluctuation pressure waves induced by accessing to the acoustic modes of the piping systems with the rotation and impeller passing pulsations of the feed water pump. There are the phenomena amplified at the same time in the range of 415 ~ 455Hz, 830 ~ 910Hz, 1245 ~ 1365Hz and 1660 ~ 1820Hz on the spectrums of the vibration, the sound pressure, and the pressure fluctuation waves.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.1-11
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• 2015

Feedwater Heaters are important components in a nuclear power plant. As the age of heater increases, the maintenance cost required for continuous operation also increases. Most heaters have the carbon steel shells, tube support plates and flow baffles. The carbon steel is susceptible to flow-accelerated corrosion. This is especially true if the flow has a two-phase mixture of steam and condensate. The wall thinning around the wet steam entrance area of the shell is inevitable during some long term operation. The structural integrity of the feedwater heater shell affects the safe operation of the nuclear power plant. Therefore, it is needed for the thinned shell to be repaired. The maintenance method for preventing failure of the shell should be determined by investigating various factors including the stress distribution of thinned area. The stress analysis of the shell including the steam entrance region is studied in this paper. The results of thinned shell is compared with that of intact shell.