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용접부 잔류응력에 기인한 경로의존성 J-적분의 측정에 대한 연구
강선예(Seon-Ye Kang),유완(One Yoo),박성호(Sung-Ho Park),윤기석(Ki-Seok Yoon),최택상(Taek-Sang Choi),이봉상(Bong-Sang Lee) 대한기계학회 2005 대한기계학회 춘추학술대회 Vol.2005 No.5
The J-integral is an important fracture mechanics parameter applicable for many areas which provide the integrity evaluation methods. Fracture of welding structure usually starts at weld region due to high weld residual stress and defects easily found in it. However, it is difficult to calculate the correct J-integral of the structure containing weld because J-integral is path dependent due to the residual stress. It is also true that there are no evidence that the J-integral measuring methods suggested by the ASTM or others are appropriate for J-integral specimens with the residual stress. Therefore, in this paper, the J-integral measuring methods suggested by the ASTM or others were evaluated and the CMOD based J-integral measuring method is suggested as an appropriate method for measuring J-integral of specimens containing weld residual stress.
원자로냉각재 환경영향을 고려한 고리1호기 가압기 밀림관 피로평가
양준석(Jun-Seog Yang),박치용(Chi-Yong Park),강선예(Seon-Ye Kang) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.5
The original design of Kori Unit 1 did not consider the thermal stratification phenomena in the surge line piping and the insurge / outsurge out of limit in the pressurizer lower head during plant heat up and cooldown condition. For the plant life extension over 20 years, the metal fatigue evaluation considering the effects of reactor water environment was performed to determine the impact of the insurge / outsurge out of transients and thermal stratification transients, in conjunction with design transient effects, on the environmental fatigue usage factors at the critical locations of the surge line and pressurizer surge nozzle. In this paper, the effects of the environmental fatigue were evaluated and discussed by confining to the thermal stratification transients in the surge line and insurge / outsurge transients in the pressurizer surge nozzle. Also, the use of fatigue life correction factor to incorporate the effects of environment into the ASME Code fatigue evaluation for the surge line piping and pressurizer surge nozzle was discussed in this paper.
가압기 밀림관 환경피로평가를 위한 피로보정계수 적용에 관한 연구
양준석(Jun-Seog Yang),박치용(Chi-Yong Park),강선예(Seon-Ye Kang) 대한기계학회 2009 大韓機械學會論文集A Vol.33 No.10
Nuclear power plants applying for the continued operation over design life are required to address the effects of reactor water environment in fatigue design requirement of the ASME Code. Reactor water environmental effects are generally evaluated by calculating fatigue correction factors on fatigue usage. This paper describes the application for pressurizer surge line of environmental fatigue correction factors and the strain rate impact in the application. From this paper, the environmental fatigue correction factors resulted from the assumption of a step change in temperature are especially compared with those calculated from the data measured during plant startup. As a conclusion of this paper, the design transient conditions applied to the fatigue design may be conservative in case of the environmental fatigue evaluation.
〈기술논문〉 설계기준 초과 과도조건을 가정한 한국 표준형 원전 가압기 밀림노즐 구조건전성에 관한 연구
박정수(Jeong Soo Park),김종민(Jong Min Kim),강선예(Seon Ye Kang),박성호(Sung Ho Park),손갑헌(Gap Heon Sohn) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.5
Out-of-Limit Transients due to thermal stratification are discovered in the pressurizer of the nuclear power plants supplied by Westinghouse. The occurring of out-of-limit transients are not reported in the pressurizer of OPR1000. However, the occurring of out-of-limit transients are assumed for the pressurizer of OPR1000 and are evaluated for the pressurizer surge nozzle structural integrity. The evaluation results show that the ASME Code requirements are met even though out-of-limit transients are considered additionally.
양준석(Jun-Seog Yang),박치용(Chi-Yong Park),강선예(Seon-ye Kang) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.5
원자력발전소 가압기와 고온루프배관의 운전조건에 따른 계통온도차로 인해서 밀림관에서는 서로 다른 밀도차에 의해서 층이 분리된 채 존재하는 열성층 현상이 발생한다. 원자력발전소 운전 중 가장 큰 온도차는 가열 및 냉각 운전기간에 발생하며 이 때 원자로냉각재가 가압기로 유입될 때 가압기로 들어오는 차가운 유체는 가압기 밀림노즐과 하부헤드를 냉각하면서 온도구배를 야기시키며 반대로 유출이 되는 경우 같은 부위를 다시 가열시킬 수 있다. 본 논문에서는 가열 및 냉각 운전기간 동안 원자로냉각재가 유입 또는 유출되는 과도조건을 작용하여 ASME Code Section Ⅲ 방법론에 따라 가압기 하부헤드의 구조건전성을 평가하였다. 평가 결과 가압기 하부헤드 모든 영역에서 설계수명기간 동안 ASME Code Sec. Ⅲ의 허용요건을 만족하였다. When the temperature difference between the reactor coolant system and the pressurizer during heat-up and cool down in the nuclear power plant is large, these out of limit conditions can produce significant temperature transients in the pressurizer lower head. This paper evaluates the impact of pressurizer out-of-limit insurge/outsurge transients on the structural integrity of the pressurizer lower head in the nuclear power plant. The stresses to be considered are those resulting from the specified loadings of internal pressure, deadweight, thermal, seismic, and thermal gradients. The evaluation on the primary plus secondary stress intensity ranges and the cumulative fatigue usage factors were included. The calculated stress intensity range and usage factor values are compared with the applicable limits of Section Ⅲ of the ASME Boiler and Pressure Vessel Code. According to the evaluation results, all requirements of the ASME Section Ⅲ, Subsection NB are satisfied.