침식부식으로 인한 탄소강 배관의 잔여수명 예측시스템 개발 및 적용

황경모,노희영,진태은 ( K . M . Hwang,H . Y . Roh,T . E . Jin ) 한국화학공학회 1996 화학공학의이론과응용 Vol.2 No.2

급수가열기 충격판 설계변경에 따른 동체감육 완화에 관한 유동해석 연구

김경훈,황경모,진태은,Kim K. H.,Hwang K. M.,Jin T. E. 한국시뮬레이션학회 2005 한국시뮬레이션학회 논문지 Vol.14 No.2

Feedwater heaters of many nuclear power plants have recently experienced wall thinning damage, which will increase as operating time progresses. As it is judged that the wall thinning damages have generated due to local fluid behavior around the impingement baffle installed in downstream of the high pressure turbine extraction steam line to avoid colliding directly with the tubes, numerical analyses using PHOENICS code were performed for two models with original clogged impingement baffle and modified multi-hole impingement baffle. To identify the relation between wall thinning and fluid behavior, the local velocity components in x-, y-, and z-directions based on the numerical analysis for the model with the clogged impingement baffle were compared with the wall thickness data by ultrasonic test. From the comparison of the numerical analysis results and the wall thickness data, the local velocity component only in the y-direction, and not in the x- and z-direction, was analogous to the wall thinning configuration. From the result of the numerical analysis for the modified impingement baffle to mitigate the shell wall thinning, it was identified that the shell wall thinning may be controlled by the reduction of the local velocity in the y-direction.

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.

오리피스 유형별 유동가속부식 영향 분석

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.

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.

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.

주기적안전성평가를 위한 원전 열교환기 Fouling 평가

황경모(K. M. Hwang),진태은(T. E. Jin),한상길(S. G. Han),김병섭(B. S. Kim) 대한기계학회 2003 대한기계학회 춘추학술대회 Vol.2003 No.4

Fouling of heat exchangers is generated by water-borne deposits, commonly known as foulants<br/> including particulate matter from the air, migrated corrosion produces; silt, clays, and sand suspended in<br/> water; organic contaminants; and boron based deposits in plants. This fouling is known to interfere<br/> with normal flow characteristics and reduce thermal efficiencies of heat exchangers. This paper focuses<br/> on fouling analyses for six heat exchangers of two primary systems in two nuclear power plants; the<br/> regenerative heat exchangers of the chemical and volume control system and the component cooling<br/> water heat exchangers of the component cooling water system. To analyze the fouling for heat<br/> exchangers, fouling factor was introduced based on the ASME O&M codes and TEMA standards.<br/> Based on the results of the fouling analyses, the present thermal performances and fouling levels for<br/> the six heat exchangers were predicted.

원전 설계기준 사고시 냉각재계통 부분정체로 인한 비대칭 열유동 혼합해석에 관한 연구

황경모 ( K. M. Hwang ),진태은 ( T. E. Jin ),김경훈 ( K. H. Kim ) 한국분무공학회 2003 한국액체미립화학회지 Vol.8 No.1

N/A When a cold HPSI (High Pressure Safety Injection) fluid associated with a design basis accident, such as LOCA (Loss of Coolant Accident), enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena may arise due to incomplete mixing. If the stratified flow enters a reactor pressure vessel downcomer, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. Previous thermal-mixing analyses have assumed that the thermal stratification phenomena generated in stagnated loop of a partially stagnated coolant loop are neutralized in the vessel downcomer by strong flow from unstagnated loop. On the basis of these reasons, this paper presents the thermal-mixing analysis results in order to identify the fact that the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is affected by the strong flow of the unstagnated loop.

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.

급수가열기 충격판 설계변경에 따른 동체감육 완화에 관한 유동해석 연구

황경모(K. M. Hwang),진태은(T. E. Jin),김경훈(K. H. Kim) 대한기계학회 2004 대한기계학회 춘추학술대회 Vol.2004 No.8