• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2012년도 춘계학술대회 논문집
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• pp.1061-1062
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• 2012

• 대한기계학회:학술대회논문집
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• 대한기계학회 2005년도 창립60주년 기념 추계 학술대회
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• pp.79.2-79.2
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• 2005

• 한국분무공학회지
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• 제15권2호
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• pp.83-93
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• 2010

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle inside feedwater heater installed downstream of the turbine extraction stream line. At that point, the extract steam from the turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows to reverse direction after impinging the impingement baffle, the shell wall of feedwater heaters may be affected by flow-accelerated corrosion. In this paper, to compare degree of shell wall thinning mitigation rate to squared type with mitigation rate of other type baffle plate, four different types of impingement baffle plate-squared, curved, mitigating type and multi-hole type-applied inside the shell. With these comparison data, this paper describes operation of experiments and numerical analysis which is composed similar condition with real feed water heater. And flow visualization is operated for verification of experiments and numerical analysis. In conclusion, this study shows that mitigating type and multi-hole type baffle plate are more effective than other baffle plate about prevention of pressure concentration and pressure value decrease.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2013년도 춘계학술대회 논문집
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• pp.1197-1198
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• 2013

• 설비공학논문집
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• 제23권7호
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• pp.483-491
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• 2011

When pipe components made of carbon steel in nuclear, fossil, and industry are exposed to flowing fluid, wall thinning caused by FAC(flow accelerated corrosion) can be generated and eventually ruptured at the portion of pressure boundary. A study to identify the locations generating local wall thinning and to disclose turbulence coefficient related to the local wall thinning was performed. Experiment and numerical analyses for tee of down scaled piping components were performed and the results were compared. In particular, flow visualization experiment which was used alkali metallic salt was performed to find actual location of local wall thinning inside tee components. To disclose the relationship between turbulence coefficients and local wall thinning, numerical analyses were performed for tee components. The turbulence coefficients based on the numerical analyses were compared with the local wall thinning based on the measured data. From the comparison of the results, the vertical flow velocity component(Vr) flowing to the wall after separating in the wall due to the geometrical configuration and colliding with the wall directly at an angle of some degree was analogous to the configuration of local wall thinning.

• 대한기계학회논문집A
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• 제33권10호
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• pp.991-1004
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• 2009

Operating experience of the various components in the nuclear power plants has shown that a variety of degradation mechanisms can occur during operation. Therefore, the accurate lifetime evaluation and systematic management are very important for the safe as well as the economical operation of the nuclear power plants. In this paper, the characteristics of a total of 17 degradation mechanisms were reviewed and the plausible degradation mechanisms such as stress corrosion cracking, fatigue, irradiation embrittlement, and so on, were identified. Also, the lifetime evaluation technologies which have been developed for the application to the domestic nuclear power plants are described. In addition, a total of 48 aging management programs which have been established for the safe operation of the various components are explained.

• 한국압력기기공학회 논문집
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• 제12권2호
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• pp.40-49
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• 2016

Flow accelerated corrosion (FAC) of the carbon steel piping has been a significant problem in nuclear power plants. FAC occurs under certain hydrodynamic, environmental, and material conditions, and extensive research into the factors of FAC has been conducted. The basic process of FAC is now relatively well understood; however, a full mechanistic model has not yet been established. Recently, the Korea Atomic Energy Research Institute (KAERI) has built a large experiment loop system for FAC. To produce significant experimental results using this system, the factors affecting on FAC should be analyzed quantitatively, and a model needs to be developed. In this work, a statistical modeling methodology to develop an empirical model is described in detail, and a preliminary model is suggested. Firstly, FAC data were collected from the research literature in Japan and the results of domestic experiments. The flow rate, water temperature, pH at room temperature, and the Cr content are selected as major factors, and nonlinear regression is used to find the best fit of the available data. An iterative procedure between suggesting and evaluating a model is used until an optimum model is obtained. The developed model gives the FAC rate comparable to the measured FAC rate. The developed model is going to be refined using additional laboratory data in the future.

• 대한기계학회논문집B
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• 제35권10호
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• pp.1105-1110
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• 2011

유동가속부식(FAC)은 가장 잘 알려진 탄소강 배관 손상 메커니즘으로 현재 국내 전 원전에서는 유동가속부식으로 인한 감육현상을 관리할 수 있는 체계적인 방안이 수립되어 있다. 그러나, 발전소 배관은 다양한 침식손상 메커니즘에 의해 여전히 손상을 받고 있다. 대표적인 침식 메커니즘은 캐비테이션, 액적충돌침식(LDIE), 플래싱, 고체입자침식(SPE)이다. 본 논문에서 기술하는 액적충돌침식 은 손상예측이 어렵고, 관리를 위한 체계적인 방안도 수립되어 있지 않다. 본 논문에서는 실제 발전소 현장에서 발생한 사례를 바탕으로 기존에 개발된 예측 모델과 실험을 통해 얻어진 상관식을 비교하여 액적충돌침식으로 인한 손상을 평가할 수 있는 방법을 제시하였다.

• 한국마린엔지니어링학회:학술대회논문집
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• 한국마린엔지니어링학회 2012년도 전기공동학술대회 논문집
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• pp.152-153
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• 2012

While engine room arranging pipe which is used from the vessel, It measured the internal flow of 90 degree elbow which is used from the main condenser. Fluid flow in elbow of 90 degree is measured by PIV and Dewetron system. The Reynolds number adopts 50000 and experimental study of flow field in the elbow.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.2977-2982
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• 2007

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle - installed downstream of the high pressure turbine extraction steam line - inside number 5A and 5B feedwater heaters. At that point, the extracted steam from the high pressure turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of the number 5 high pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the down scale experimental data which effect on disclosing of the shell wall thinning of the high pressure feedwater heaters by porous plate.