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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.
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.