액적충돌침식 영향 배관의 설계변경에 관한 연구

황경모(Kyeong Mo Hwang),이찬규(Chan Gyu Lee),방극진(Keug Jin Bhang),임영식(Young Sig Yim) 대한기계학회 2011 大韓機械學會論文集B Vol.35 No.10

액적충돌침식은 증기나 공기에 포함된 액적이 금속 소재에 고속으로 충돌할 때 모재가 손상되는 현상이다. 액적충돌침식 손상은 증기터빈이나 빗방울과 부딪치는 항공기에서 주로 발생되어 왔으나 최근에는 원전 배관에서도 발생하고 있다. 원전 배관 중에서도 특히 높은 압력강하가 발생하고 2상 증기가 흐르는 배관에서 주로 발생한다. 실제 2011년 초반 국내 한 원전에서는 2상 증기가 흐르는 배관에서 액적충돌침식 손상으로 인한 누설이 발생한 바 있다. 본 논문에서는 액적충돌침식 손상이 발생한 배관에 대하여 손상을 억제할 수 있는 설계변경 방안에 관한 연구를 수행하였다. 설계변경은 유체 유동측면에서 분석하였으며, 상용 수치해석 코드인 FLUENT를 이용하였다. Liquid droplet impingement erosion (LDIE) is caused by the impact of high-velocity droplets entrained in steam or air on metal. The degradation caused by the LDIE has been experienced in steam turbine internals and high-velocity airplane components (particularly canopies). Recently, LDIE has also been observed in the pipelines of nuclear plants. LDIE among the pipelines occurs when two-phase steam experiences a high pressure drop (e.g., across an orifice in a line to the condenser). In 2011, a nuclear power plant in Korea experienced a steam leak caused by LDIE in a pipe through which a two-phase fluid was flowing. This paper describes a study on the design change of a pipe affected by LDIE in order to mitigate the damage. The design change has been reviewed in terms of fluid dynamics by using the FLUENT code.

수치해석 기법을 활용한 FAC 예측 프로그램 보완

황경모(Kyeong Mo Hwang),진태은(Tae Eun Jin),박원(Won Park),오동훈(Dong Hoon Oh) 대한기계학회 2010 大韓機械學會論文集B Vol.34 No.4

고온, 고압의 유체가 흐르는 탄소강 배관에서는 유동가속부식으로 인한 배관감육 현상이 발생할 수 있다. 화력 및 원자력발전소에서 유동가속부식으로 인한 배관 손상시 고비용의 보수와 발전 정지를 유발할 뿐 아니라 발전소 신뢰도 및 안전성에 영향을 미칠 수도 있다. CHECWORKS 프로그램은 국내 발전소에서 유동가속부식에 의한 배관 손상을 예방하기 위하여 배관 두께검사 데이터를 평가하고 검사 계획을 수립하는데 이용되어 왔다. 그러나 상기 프로그램은 원전 차측 배관 모두를 데이터베이스화한 후에 배관라인 그룹별로 유동가속부식 손상을 예측하기 때문에 국부적으로 감육에 민감한 부위를 찾는데 어려움이 있다. 본 논문에서는 CHECWORKS 프로그램을 이용하여 해석을 수행하고 수치해석을 통하여 검증할 수 있는 방법론을 기술하였다. 또한 국내 원전 2개의 배관 라인그룹에 대하여 CHECWORKS 프로그램을 이용한 유동가속부식 민감 부위를 FLUENT를 이용한 수치해석 결과와 비교하였다. Flow-accelerated corrosion (FAC) leads to thinning of steel pipe walls that are exposed to flowing water or wet steam. From experience, it is seen that FAC damage to piping at fossil and nuclear plants can result in outages that require expensive repairs and can affect plant reliability and safety. CHECWORKS have been utilized in domestic nuclear plants as a predictive tool to assist FAC engineers in planning inspections and evaluating the inspection data so that piping failures caused by FAC can be prevented. However, CHECWORKS may be occasionally ignore local susceptible portions when predicting FAC damage in a group of pipelines after constructing a database for all the secondary side piping in nuclear plants. This paper describes the methodologies that can complement CHECWORKS and the verifications of CHECWORKS prediction results using numerical analysis. FAC susceptible locations determined using CHECWORKS for two pipeline groups of a nuclear plant was compared with determined using the numerical-analysis-based FLUENT.

탄소강배관 설계 변경시에 발생한 두께편차와 국부감육의 상관성에 관한 연구

황경모(Kyeong Mo Hwang),윤훈(Hun Yun) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.10

Flow accelerated corrosion (FAC) leads to wall thinning of carbon steel piping exposed to flowing water or wet steam. Experience has shown that FAC damage to piping at fossil and nuclear plants can lead to costly outages and repairs and can affect plant reliability and safety. To protect the wall thinning damage, the utility performs periodic inspect for the susceptible piping and replaces the thinned pipe if the thickness exceeds the managing criteria. This paper describes the relationship between local wall thinning and thickness difference generated inevitably in design modification with thick piping on a prevention basis.

가압경수로형 원전 급수가열기 출구헤더 연결배관에 대한 국부감육 원인분석에 관한 연구

황경모(Kyeong Mo Hwang),이찬규(Chan Kyoo Lee) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.10

Carbon steel piping exposed to flowing high temperature, high pressure, and high speed water or wet steam experience wall thinning phenomena. Wall thinning damages caused by FAC (Flow Accelerated Corrosion) may be generated in the piping of ordinary industry plants, fossil power plants, and nuclear power plants. The pipe wall thinning in nuclear power plants should be considered more strictly than that of other plants. This is because the pipe wall thinning leads to leakage, rupture, and unplanned shutdown in nuclear power plants, which can affect plant reliability and safety. This paper describes the cause analysis result of local wall thinning for the piping connected to the outlet header of a feedwater heater installed in a PWR nuclear power plant. The results are based on the flow behaviors inside piping simulated by numerical analysis and the UT thickness data analysis.

[기술논문특집] : 원자력발전소에서의 침부식 ; 원자력발전소 2차계통 탄소강배관 감육관리 현황 및 전망

황경모 ( Kyeong Mo Hwang ) 한국부식방식학회(구 한국부식학회) 2012 부식과 방식 Vol.11 No.1

주증기계통 오리피스 후단 소구경 배관의 감육 및 누설 발생

황경모 ( Kyeong Mo Hwang ) 한국부식방식학회(구 한국부식학회) 2013 Corrosion Science and Technology Vol.12 No.5

A number of components installed in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), Cavitation, Flashing, and LDIE (Liquid Droplet Impingement Erosion). Those aging mechanisms can lead to thinning of the components. In April 2013, one (1) inch small bore piping branched from the main steam line experienced leakage resulting from wall thinning in a 1,000 MWe Korean PWR nuclear power plant. During the normal operation, extracted steam from the main steam line goes to condenser through the small bore piping. The leak occurred in the downstream of an orifice. A control valve with vertical flow path was placed on in front of the orifice. This paper deals with UT (Ultrasonic Test) thickness data, SEM images, and numerical simulation results in order to analyze the extent of damage and the cause of leakage in the small bore piping. As a result, it is concluded that the main cause of the small bore pipe wall thinning is liquid droplet impingement erosion. Moreover, it is observed that the leak occurred at the reattachment point of the vortex flow in the downstream side of the orifice.

액적충돌침식 영향 배관의 설계변경에 관한 연구

황경모(Kyeong Mo Hwang),이찬규(Chan Gyu Lee),방극진(Keug Jin Bhang),임영식(Young Sig Yim) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.4

액적충돌침식은 증기나 공기에 포함된 액적이 금속 소재에 고속으로 충돌할 때 모재가 손상되는 현상이다. 액적충돌침식 손상은 증기터빈이나 빗방울과 부딪치는 항공기에서 주로 발생되어 왔으나 최근에는 원전 배관에서도 발생하고 있다. 원전 배관 중에서도 특히 높은 압력강하가 발생하고 2상 증기가 흐르는 배관에서 주로 발생한다. 실제 2011년 초반 국내 한 원전에서는 2상 증기가 흐르는 배관에서 액적충돌침식 손상으로 인한 누설이 발생한 바 있다. 본 논문에서는 액적충돌침식 손상이 발생한 배관에 대하여 손상을 억제할 수 있는 설계변경 방안에 관한 연구를 수행하였다. 설계변경은 유체 유동측면에서 분석하였으며, 상용 수치해석 코드인 FLUENT를 이용하였다. Liquid droplet impingement erosion is caused by the impact to metal of high velocity droplets entrained in steam or air. The degradations caused by liquid droplet impingement erosion have experienced in steam turbine internals and high-velocity airplane components (particularly canopies) by rain. Recently, the liquid droplet impingement erosion is also evoked in the pipelines of nuclear plants. Liquid droplet impingement erosion among the pipelines occurs when a two-phase steam experiences a high-pressure drop (e.g., across an orifice in a line to the condenser). A nuclear power plant in Korea experienced a steam leak caused by liquid droplet impingement erosion in a pipe flowing 2-phase fluid in 2011. This paper describes a study on the design change of a pipe affected by liquid droplet impingement erosion in order to mitigate the damage. The design change has reviewed in terms of fluid dynamic using FLUENT code.

다관원통형 열교환기의 파울링 및 관막음 여유 평가법 개발 연구

황경모(Kyeong Mo Hwang),진태은(Tae Eun Jin) 대한기계학회 2003 대한기계학회 춘추학술대회 Vol.2003 No.11

As operating time of heat exchangers progresses, fouling generated by water-borne deposits increases and thermal<br/> performance decreases. The fouling is known to interfere with normal flow characteristics and reduce thermal<br/> efficiencies of heat exchangers. The heat exchangers of nuclear power plants have been analyzed in terms of<br/> the heat flux and heat transfer coefficient at test conditions based on the ASME OM-S/G-Part 2 as a means<br/> of heat exchanger management. It is hard to estimate the heat performance trend and to establish the future<br/> management plan. This paper describes the fouling evaluation method which can evaluate the thermal<br/> performance for heat exchangers and estimate the future fouling variations and the plugging margin evaluation<br/> method which can reflect the current fouling level developed in this study. To develop the fouling and<br/> plugging margin evaluation methods for heat exchangers, fouling factor was introduced based on the ASME<br/> O&M codes and TEMA standards. For the purpose of verifying the two evaluation methods, the fouling and<br/> plugging margin evaluations were performed for a component cooling heat exchanger in a nuclear power<br/> plant.

원전 배관감육 평가를 위한 새로운 기법의 도입 및 타당성

황경모 ( Kyeong Mo Hwang ),윤훈 ( Hun Yun ),박현철 ( Hyun Cheol Park ) 한국부식방식학회(구 한국부식학회) 2014 Corrosion Science and Technology Vol.13 No.2

A huge number of carbon steel piping components installed in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), Cavitation, Flashing, and LDIE (Liquid Droplet Impingement Erosion). Those aging mechanisms can lead to thinning of the piping components. To manage the wall thinning degradation, most of utilities in the world predict the wall thinning rate based on the computational program such as CHECWORKS, COMSY, and BRT-CICERO, evaluate the UT (Ultrasonic Test) data, and determine next inspection timing, repair or replacement, if needed. There are several evaluation methods, such as band, blanket, and strip methods, commonly used for determining the wear of piping components from single UT inspection data. It has been identified that those single UT evaluation methods not only do not consider the manufacturing features of pipes, but also may exclude the data of the most thinned point when determining the representative wear rate of piping components. This paper describes a newly developed single UT evaluation method, E-Cross method, for solving above problems and introduces application examples for several pipes and elbows. It was identified that the E-Cross method using the length and width of UT data excluded the most thinned points appropriate as the single UT evaluation method for thinned piping components.

주급수격리밸브 하부몸체의 감육현상 분석을 위한 실측 및 수치해석 연구

황경모(Kyeong Mo Hwang),진태은(Tae Eun Jin) 대한기계학회 2005 대한기계학회 춘추학술대회 Vol.2005 No.5

A numerical analysis study has performed in terms of fluid dynamics to identify the wall thinning generated in the main feedwater isolation valve body of a nuclear power plant. To review the relations between flow characteristics and the wall thinning induced by flow accelerated corrosion (FAC), numerical analysis using FLUENT code and ultrasonic tests (UT) were performed. The local velocities according to the analysis results were compared with the distribution of the measured wall thickness by ultrasonic tests. The comparison results show that the local velocity in the x-direction had no correlation with the wall thinning but the local velocity in the y-direction and turbulence intensity had a great influence on that. These results provide a good match to those of the previous studies - locations colliding vertically against components undergo severe wall thinning. These results may be utilized to the design modification and the wall thinning management for main feedwater isolation valves for preventing the wall thinning degradation.