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장창희,장훈,홍종대,조현철,김태순,이재곤 한국원자력학회 2013 Nuclear Engineering and Technology Vol.45 No.7
Environmental fatigue of the metallic components in light water reactors has been the subject of extensive research andregulatory interest in Korea and abroad. Especially, it was one of the key domestic issues for the license renewal of operatingreactors and licensing of advanced reactors during the early 2000s. To deal with the environmental fatigue issue domestically,a systematic test program has been initiated and is still underway. The materials tested were SA508 Gr.1a low alloy steels,316LN stainless steels, cast stainless steels, and an Alloy 690 and 52M weld. Through tests and subsequent analysis, themechanisms of reduced low cycle fatigue life have been investigated for those alloys. In addition, the effects of temperature,dissolved oxygen level, and dissolved hydrogen level on low cycle fatigue behaviors have been investigated. In this paper, thetest results and key analysis results are briefly summarized. Finally, an on-going test program for hot-bending of 347 stainlesssteel is introduced.
영향계수를 이용한 원자로 압력용기의 운전제한곡선 작성 : 냉각곡선
장창희,Jang, Chang-Hui 대한기계학회 2002 大韓機械學會論文集A Vol.26 No.3
During heatup and cooldown of pressurized water reactor, thermal stress was generated in the reactor pressure vessel (RPV) because of the temperature gradient. To prevent potential failure of RPV, pressure was required to be maintained below the P-T limit curves. In this paper, several methods for constructing the P-T limit curves including the ASME Sec. XI, App. G method were explained and the results were compared. Then, the effects of the various parameters such as flaw size, flaw orientation, cooldown rate, existence of chad, and reference fracture toughness, were evaluated. It was found that the current ASME Sec. XI App. G method resulted in the most conservative P-T limit curve. As the more accurate fracture mechanics analysis results were used, some of the conservatism can be removed. Among the parameters analysed, reference flaw orientation and reference fracture toughness curve had the greatest effect on the resulting P-T limit curves.
영향계수를 이용한 원통용기 축방향 표면결함의 응력확대계수의 계산
장창희,문호림,정일석,김태룡,Jang, Chang-Heui,Moon, Ho-Rim,Jeong, Ill-Seok,Kim, Tae-Ryong 대한기계학회 2002 大韓機械學會論文集A Vol.26 No.11
For integrity analysis of nuclear reactor pressure vessel, including the Pressurized thermal shock analysis, the fast and accurate calculation of the stress intensity factor at the crack tip is needed. For this, a simple approximation scheme is developed and the resulting stress intensity factors for axial semi-elliptical cracks in cylindrical vessel under various loading conditions are compared with those of the finite element method and other approximation methods, such as Raju-Newman's equation and ASME Sec. Xl approach. For these, three-dimensional finite-element analyses are performed to obtain the stress intensity factors for various surface cracks with t/R = 0.1. The approximation methods, incorporated in VINTIN (Vessel INTegrity analysis-INner flaws), utilizes the influence coefficients to calculate the stress intensity factor at the crack tip. This method has been compared with other solution methods including 3-D finite clement analysis for internal pressure, cooldown, and pressurized thermal shock loading conditions. The approximation solutions are within $\pm$2.5% of the those of FEA using symmetric model of one-forth of a vessel under pressure loading, and 1-3% higher under pressurized thermal shock condition. The analysis results confirm that the VINTIN method provides sufficiently accurate stress intensity factor values for axial semi-elliptical flaws on the surface of the reactor pressure vessel.
장창희,김성환,사인진,Daejong Kim 대한기계학회 2016 JOURNAL OF MECHANICAL SCIENCE AND TECHNOLOGY Vol.30 No.10
The creep behavior of a nickel-base superalloy, Alloy 617, which is considered as a candidate material for the very high temperature gas cooled reactor, was studied. Creep rupture tests were carried out at 800 o C, 900 o C and 1000 o C in static and flowing helium environments as well as in air. Creep rupture life in static helium was longer than that in air, while it was shorter in flowing helium environments. Microstructure observation of the creep tested specimens showed that the shorter creep rupture life in flowing helium was associated with the thicker oxide layer, greater decarburization depth, and deeper internal oxidation happened during the creep tests. The degree of such oxidation damage was quantified for the creep tested specimens and correlated with the creep rupture life in different environments.