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ToSPACE 프로그램을 이용한 FAC 해석결과와 실험결과 비교
황경모,윤훈,서혁기,정의제,김경모,김동진 한국부식방식학회 2020 Corrosion Science and Technology Vol.19 No.3
A number of piping components in the secondary system of nuclear power plants (NPPs) are exposed toaging mechanisms, such as flow-accelerated corrosion (FAC), cavitation, flashing, solid particle erosion,and liquid droplet impingement erosion. Those mechanisms may lead to thinning, leaking, or rupture ofthe components. Due to the pipe ruptures caused by wall thinning in Surry unit 2 in the USA in 1986and Mihama unit 3 in Japan in 1994, pipe wall thinning management has emerged as one of the mostimportant issues in the nuclear industry. To manage pipe wall thinning, a foreign program has been utilizedfor NPPs in Korea since 1996. As our experience and knowledge of pipe wall thinning management haveaccumulated, our program needs to reflect our experience, requests from users, and the result of recentexperiments using Flow Accelerated Corrosion Testing System (FACTS). FACTS is the empirical experimentalfacility developed by Korea Atomic Energy Research Institute (KAERI) for tests. Accordingly, KEPCO-E&Cdeveloped a 3D-based pipe wall thinning management program called ToSPACE in 2016. This paper describesa comparison between the FAC analysis results using ToSPACE and the experimental results using FACTSto verify their applicability to pipe wall thinning management in NPPs.
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.
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.
필수냉각기 응축기에 대한 Fouling 평가법 개발 및 적용
황경모(Kyeong Mo Hwang),진태은(Tae Eun Jin) 대한기계학회 2004 대한기계학회 춘추학술대회 Vol.2004 No.11
Heat exchangers in nuclear power plants are used for various purposes, such as safe shutdown of nuclear reactor, increase of thermal efficiency, maintenance of temperature inside building, final heat sink, reduction of thermal stress by cold water injection, etc. As operating time of these heat exchangers progresses, fouling generated by water-borne deposits increases and thermal performance decreases. Even though the thermal performance tests for heat exchangers without phase change in domestic nuclear power plants have performed with a fixed interval, thermal performance tests for finned tube heat exchangers with condensation have not performed to date. This paper describes the basic theory of fouling evaluation method for finned tube heat exchangers and the application result for an essential chiller condenser using the freon-1,2,3 as refrigerant.
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.
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.
탄소강배관 다중 UT 측정두께를 활용한 감육여부 판별법 개발
황경모 ( Kyeong Mo Hwang ),윤훈 ( Hun Yun ),이찬규 ( Chan Kyoo Lee ) 한국부식방식학회(구 한국부식학회) 2012 Corrosion Science and Technology Vol.11 No.5
To manage the wall thinning of carbon steel piping in nuclear power plants, the utility of Korea has performed thickness inspection for some quantity of pipe components during refueling outages and determined whether repair or replacement after evaluating UT (Ultrasonic Test) data. When the existing UT data evaluation methods, such as Band, Blanket, PTP (Point to Point) Methods, are applied to a certain pipc component, unnecessary re-inspecting situations may be generated even though the component does not thinned. In those cases, economical loss caused by repeated inspection and problems of maintaining the pipe integrity followed by decreasing of newly inspected components may be generated. EPRI(Electric Power Research Institute) in USA has suggested several statistical methods, TPM(Total Point Method), LSS(Least Square Slope) Method, etc. to distinguish whether multiple inspecting components have thinned or not. This paper presents the analysis results for multiple inspecting components over three times based on both NAM (Near Area of Minimum) Method developed by KEPCO-E&C and the and the other methods suggested by EPRI.
황경모(Kyeong Mo Hwang),진태은(Tae Eun Jin),김경훈(Kyung Hoon Kim) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.11
To mitigate the effects of cavitation and flashing, several types of orifices have been installed in the pipeline of new nuclear power plants. To review the effects of wall thinning caused by flow-accelerated corrosion by the types of orifices, which are cone and plate, and the relation between flow behavior and local wall thinning, experiments and numerical analyses for the downstream pipe of two types of orifices were performed. The experimental results in terms of static pressure obtained for the experimental facilities were compared with those of three-dimensional (3D) numerical analyses using the FLUENT code. As the results of review of flow-accelerated corrosion effects based on the experiment and numerical analysis, it was identified that the orifice of cone-type can be comparatively mitigated the effects of cavitation and flashing, but can not be mitigated the effect of flow-accelerated corrosion.