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원전 배관용 SA 106 Gr.C의 협개자동용접 적용에 관한 연구
우승완(Seung Wan Woo),권재도(Jae Do Kwon),이춘열(Choon Yeol Lee),강석철(Suk Chull Kang),신호상(Ho Sang Shin) 대한기계학회 2004 대한기계학회 춘추학술대회 Vol.2004 No.11
Conventionally, SMAW process was applied to join pipes of RCL, which caused lot of loss in time and cost due to excessive heat input and defects in joining section. Recently, narrow-gap welding(NGW) process was introduced to overcome the disadvantages of SMAW. However, the application of NGW to nuclear power plant is not yet common because safety of NGW process is not proven. In present paper, the welded coupons are made of carbon steel. They are manufactured under different processes; general welding(GW), post-weld heat treatment(PWHT) after GW, repair welding after GW and PWHT with repair welding after GW in carbon steel. Performed are various mechanical tests investigation of microstructure, hardness test, tensile test at room and high temperature, bending test, impact test and J-R test. It is verified that the mechanical properties of carbon steel are greatly changed after repair welding process due to applied heat flux, and that the effect of PWHT is beneficial.
CF8M과 SA508 용접재의 열화에 따른 파괴특성 평가
우승완(Seung Wan Woo),권재도(Jae Do Kwon),최성종(Sung Jong Choi),최영환(Young Hwan Choi) 대한기계학회 2004 대한기계학회 춘추학술대회 Vol.2004 No.4
In a primary reactor cooling system(RCS), a dissimilar weld zone exists between cast stainless steel(CF8M) in a pipe and low-alloy steel(SA508 cl.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time to a reactor operating temperature between 290 and 330℃, while no effect is observed in SA508 cl.3. The specimens are prepared by an artificially accelerated aging technique maintained for 300, 1800 and 3600 hrs at 430℃, respectively. The specimens for elastic-plastic fracture toughness tests are prepared one type, which notch is created in the heat affected zone(HAZ) of CF8M. And, the specimens for fatigue crack growth tests are prepared in three classes, which notches are created at the center of deposited zone, the HAZ of CF8M, and the HAZ of SA508 cl.3. From the experiments, the J-integral values with the increase of aging time decrease, and the differences of the fatigue crack growth behaviors are relatively small in the three classes specimens.
SA312 TP347 재료 협개용접부의 응력부식에 관한 연구
우승완(Seung Wan Woo),이춘열(Choon Yeol Lee),권재도(Jae Do Kwon) 대한기계학회 2005 대한기계학회 춘추학술대회 Vol.2005 No.10
Conventionally, SMAW process was applied to join pipes of RCL, which caused lot of loss in time and cost due to excessive heat input and defects in joining section. Recently, narrow-gap welding(NGW) process was introduced to overcome the disadvantages of SMA W. However, the application of NGW to nuclear power plant is not yet common because safety of NGW process is not proven. In present paper, the welded coupons are made of stainless steel. They are manufactured under different processes; general welding(GW) and repair welding after GW in stainless steel. The stress corrosion cracking(SCC) tests are performed at 50, 80 and 100%,σ<SUB>y</SUB> under 42% MgCl₂ environment. The failure time of SAW base-metal at 100%,σ<SUB>y</SUB> is twice longer than that of welded-material. It is verified that welded-material of stainless steel is largely affected in effect of SCC environment. In addition, The failure time of SAW welded-material is longer than SRW welded-material. It is verified that SA 312 TP 347 welded material is largely exposed in SCC environment when it has repair welding.
CF8M과 SA508 용접재의 열화거동에 따른 파괴인성 평가
우승완(Seung Wan Woo),권재도(Jae Do Kwon),최영환(Young Hwan Choi) 대한기계학회 2005 대한기계학회 춘추학술대회 Vol.2005 No.8
In a primary reactor cooling system(RCS), a dissimilar weld zone exists between cast stainless steel(CF8M) in a pipe and low-alloy steel(SA508 c1.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time to a reactor operating temperature between 290 and 330℃, while no effect is observed in SA508 cl.3. The specimens are prepared by an artificially accelerated aging technique maintained for 300, 1800 and 3600 hrs at 430℃, respectively. The specimens for elastic-plastic fracture toughness tests are prepared one type, which notch is created in the heat affected zone(HAZ) of CF8M and deposited zone. From the experiments, the J<SUB>IC</SUB> value notched in HAZ of CF8M presented a rapid decrease up to 300 hours at 430℃ and slowly decreased according to the process in the thermal aging time. Also, the J1c value presented a lower value than that of the CF8M base metal. And, the J<SUB>IC</SUB> of the deposited zone presented the lowest value of all other cases.
우승완(Seung-Wan Woo),장윤석(Yoon-Suk Chang),최재붕(Jae-Boong Choi),김영진(Young-Jin Kim),정명조(Myung-Jo Jhung),최영환(Young-Hwan Choi) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.5
The well-known flaw evaluation criteria embodied in ASME Sec. XI are not applicable to secondary system piping of nuclear power plants, mainly, due to different R/t ratios comparing with those of primary system piping. Despite of previous activities to resolve this specific issue, there are still arguments to determine whether flaws detected in the secondary system are acceptable or not for continued service. In the present study, pros and cons of relevant researches are fully discussed and unique features of secondary system piping are assessed by retrieving a database enveloping eight nuclear power plants. Subsequently, for pipe geometries beyond the ASME applicable limits, finite element analyses are carried out by changing sizes of surface crack. The analysis results showed limitations of current flaw evaluation schemes and addressed necessity of establishing a new one which will be available from further detailed finite element analyses.
열성층 및 냉각재 환경이 오스테나이트 배관의 피로수명에 미치는 영향 평가
최신범(Shin-Beom Choi),우승완(Seung-Wan Woo),장윤석(Yoon-Suk Chang),최재붕(Jae-Boong Choi),김영진(Young-Jin Kim),이진호(Jin-Ho Lee),정해동(Hae-Dong Chung) 대한기계학회 2008 大韓機械學會論文集A Vol.32 No.8
During the last two decades, lots of efforts have been devoted to resolve thermal stratification phenomenon and primary water environment issues. While several effective methods were proposed especially in related to thermally stratified flow analyses and corrosive material resistance experiments, however, lack of details on specific stress and fatigue evaluation make it difficult to quantify structural behaviors. In the present work, effects of the thermal stratification and primary water are numerically examined from a structural integrity point of view. First, a representative austenitic nuclear piping is selected and its stress components at critical locations are calculated in use of four stratified temperature inputs and eight transient conditions. Subsequently, both metal and environmental fatigue usage factors of the piping are determined by manipulating the stress components in accordance with NUREG/CR-5704 as well as ASME B&PV Codes. Key findings from the fatigue evaluation with applicability of pipe and three-dimensional solid finite elements are fully discussed and a recommendation for realistic evaluation is suggested.
IMC부를 고려한 칩-스케일 패키지의 충격 신뢰성 평가
김종민(Jong-Min Kim),우승완(Seung-Wan Woo),김영진(Young-Jin Kim),최재붕(Jae-Boong Choi),지금영(Kum-Young Ji) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.5
CSP(Chip-Scale Package) shows various failure behavior according to the way of surface treatment, number of solder balls, etc. Among them, the characteristics of IMC(intermetallic compound) layer which is generated during the joining process between the solder ball and the board are known as key factors that affect the failure behavior. The present paper investigates the failure behavior of CSP by evaluating the stress, displacement and acceleration of solder ball, IMC, copper and nickel pad through a detailed 3-dimensional drop simulation according to the Joint Electron Device Engineering Council(JEDEC) standard under different load levels. Two different surface finishing conditions of ENEPIG(5/0.4/0.12) and ENEPIG(0.1/0.8/0.12) are considered, and SAC305 is used for solder ball material properties. While ENEPIG(0.1/0.8/0.12) shows more resistant tendency against the drop impact failure due to softening, ENEPIG(5/0.4/0.12) shows weakness to failure due to the brittle characteristics of Nickel pad.
INCOLOY 800 합금의 프레팅 피로수명에 관한 연구
박대규(Dae-Kyu Park),우승완(Seung-Wan Woo),배용탁(Yong-Tak Bae),정일섭(Il-Sub Chung),채영석(Young-Suck Chai),권재도(Jae-Do Kwon) 대한기계학회 2006 대한기계학회 춘추학술대회 Vol.2006 No.11
Mechanical breakdown often comes from the fatigue in many structural parts and nuclear power plants. Among the fatigue phenomenon, especially fretting fatigue occurs in mechanical joints showing small relative movements between contact surfaces. Although the research was developed for one hundred years, occurrence mechanism is not clearly identified yet. INCOLOY alloy 800 is a iron-nickel-chromium alloy having excellent resistance to many corrosive aqueous media and high-temperature atmospheres. This alloy is used extensively in the nuclear power plants industry, the chemical industry, the heat-treating industry and the electronic industry. In this paper, the effect of fretting damage on fatigue behavior for INCOLOY alloy 800 was studied. Also, various kinds of mechanical tests such as tension and plain fatigue tests are performed. Fretting fatigue tests were carried out with flat-flat contact configuration using a bridge type contact pad and plate type specimen. Through these experiments, it is found that the fretting fatigue strength decreased about 50% compared to the plain fatigue strength. In fretting fatigue, the oblique micro-cracks at an earlier stage are initiated. These results can be used as basic data in a structural integrity evaluation of heat and corrosion resisting alloy considering fretting damages.
Inconel 690 합금의 피로균열 발생 및 진전거동에 관한 연구
정한규(Han-Kyu Jeung),권재도(Jae-Do Kwon),우승완(Seung-Wan Woo),이주석(Ju-Seok Lee) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.11
The purpose of this study is to experimentally investigate the fatigue crack initiation and propagation characteristics when nuclear power plant steam generator tube materials, Inconel 690 alloy, receive fretting fatigue. In addition, the study aims to enhance the overall reliability of plant equipments while developing fretting fatigue life and integrity evaluation methods. By using the developed experiment equipment, fretting fatigue crack growth experiments under various stress conditions, were conducted on Inconel 690 alloy to evaluate the characteristics. In addition, the study aimed not only to observe microstructures in accordance with fretting fatigue and contact pressure but also examine fretting fatigue crack initiation mechanism.
INCONEL 600 과 690 합금의 프레팅 피로수명에 관한 연구
박대규(Dae Kyu Park),배용탁(Yong Tak Bae),최성종(Sung Jong Choi),우승완(Seung Wan Woo),권재도(Jae Do Kwon) 대한기계학회 2005 대한기계학회 춘추학술대회 Vol.2005 No.11
The fretting fatigue occurs at lower stress amplitude and at lower cycles of cyclic loading then plain fatigue. INCNEL 600 and 690 alloy are high-chromium nickel alloy having excellent resistance to many corrosive aqueous media and high-temperature atmospheres. In this paper, the effect of fretting damage on fatigue behavior for INCNEL 600 and 690 alloy were studied. Also, various kinds of mechanical tests such as hardness, tension and plain fatigue tests are per formed. Through these experiments, it is found that the fretting fatigue strength decreased about 40 ~ 70% compared to the plain fatigue strength in two materials. In fretting fatigue, the wear debris is observed on the contact surface and the oblique micro-cracks at on earlier stage are initiated.