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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.
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.
정성규(S.G. Jung),진태은(T.E. Jin),정명조(MJ. Jhung) 대한기계학회 2002 대한기계학회 춘추학술대회 Vol.2002 No.5
The structural integrity of the reactor vessel with the approaching end of life must be assured for pressurized thermal shock. The regulation specifies the screening criteria for this and requires that specific analysis be performed for the reactor vessel which is anticipated to exceed the screening criteria at the end of plant life. In case the screening criteria is exceeded by the deterministic analysis. probabilistic analysis must be performed to show that failure probability is within the limit. In this study. probabilistic fracture mechanics analysis of the reactor vessel for pressurized thermal shock is performed and the effects of residual stress and master curve on the failure probability are investigated.
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.
CF8M 주조 오스테나이트 스테인리스강의 열취화에 따른 재료물성치 평가
김철(C. Kim),방흥배(H. B. Park),진태은(T. E. Jin),정일석(I. S. Jeong),석창성(C. S. Seok),박재실(J. S. Park) 대한기계학회 2003 대한기계학회 춘추학술대회 Vol.2003 No.4
CF8M cast austenitic stainless steel is used for several components such as primary coolant piping,<br/> elbow, pump casing, and valve bodies in light water reactors. These components are subject to thermal<br/> aging at the reactor operating temperature. Thermal aging results in spinodal decomposition of the<br/> delta-ferrite leading to increased strength and decreased toughness. In this study, three kinds of the<br/> aged CF8M specimen were prepared using an artificially simulated aging method. The objective of this<br/> study is to summarize the method of estimating ferrite contents, Charpy impact energy and J-R curve,<br/> and to evaluate the thermal embrittlement of the CF8M cast austenitic stainless steel piping used in the<br/> domestic nuclear power plants.
열취화에 따른 주조 스테인리스강의 재료물성치 저하 예측
이승건(S.G. Lee),박흥배(H.B. Park),진태은(T.E. Jin),송택호(T.H. Song),장창희(C.H. Jang),정일석(I.S. Jeong) 대한기계학회 2001 대한기계학회 춘추학술대회 Vol.2001 No.3
주조 오스테나이트 스테인리스강은 기계적 특성, 부식성 및 용접성 등이 뛰어나 원자력발전소의 1차 계통에 널리 사용되고 있다. 그러나 주조 오스테나이트 스테인리스강은 고온에서 장시간 사용될 때 ‘475℃ 취성’이라고 알려져 있는 열취화 현상이 발생하게 되며 이로 인해 인장강도는 증가하고 연성과 충격에너지 값은 감소하게 되어 안전성에 영향을 미치게 된다. 따라서 열취화 발생기준을 결정하는 것은 주조 스테인리스강의 건전성 평가에 매우 중요하다. 본 논문에서는 일반적인 열취화 기구 및 GALL에서 제시한 열취화 발생기준, 재료물성치 저하 예측 방안을 소개하고 이들 기준을 국내 주조 오스테나이트 스테인리스강 배관에 적용하여 열취화 발생 여부를 검토하였다. 이는 열취화를 고려한 원자력발전소 원자로냉각계통 배관의 파괴역학 해석시 유용하게 활용될 수 있을 것이다.
김종성(J.S Kim),박진석(J.S Park),진태은(T.E. Jin) 대한기계학회 2002 대한기계학회 춘추학술대회 Vol.2002 No.8
An analytical procedure for hydrogen induced cracking(HIC) on weld was developed. This procedure can be applied to evaluate weld integrity and identify its mitigation method due to the HIC damage. The analysis procedure consists of temperature analysis, residual stress analysis, and hydrogen diffustion analysis. To demonstrate the validation of the developed procedure, it applied to carbon steel weld. It shows that the results of thermal analysis were in good agreement with those of the previous study. Then, the hydrogen diffusion resulting from thermal and stress gradient were analyzed in order to calculate the diffusible hydrogen concentration and determine the most susceptible area. Also, electrotransport treatment which can substitute for conventional post weld heat treatment was represented as a potential method for mitgating the HIC susceptibility.
정성규(S.G. Jung),김현수(H.S. Kim),진태은(T.E. Jin),박영섭(Y.S. Park),김풍식(P.S. Kim),이수득(S.D. Lee),박천명(C.M. Park) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.5
In order to investigate the feasibility of continued operation of CANDU nuclear power plants, the evaluation for the aging management plans(AMPs) has been accomplished being based on the law for continued operation of the domestic nuclear power plants. Total 38 AMPs, presented in MEST Notice 2008-17, were evaluated according to the reference standards such as USNRC guidelines, CSA Code of Canada, and IAEA technical documents, etc. As a result of the assessment, it was thought that 5 items including the onetime inspection should be established and implemented before the continued operation. In addition, the establishment and execution of aging management plans for 6 items including the flow-accelerated corrosion (FAC) was necessary. In conclusion, the continued operation of a CANDU nuclear power plant is possible if aged major components are refurbished and these recommendations are implemented.