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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.
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.
증기발생기 전열관의 구조 건전성 평가를 위한 확률론적 프로그램 개발
김현수(H.S. Kim),심희진(H.J. Shim),오창균(C.K. Oh),정성규(S.G. Jung),장윤석(Y.S. Chang),김홍덕(H.D. Kim),이재봉(J.B. Lee) 대한기계학회 2012 대한기계학회 춘추학술대회 Vol.2012 No.11
The steam generator in a nuclear power plant is a large heat exchanger with thousands of thin-walled tubes and comprises over 50% of the total primary pressure-retaining boundary. Failure of the steam generator tubes can result in release of fission products to the environment outside. Therefore, an accurate integrity assessment of the tubes is of great importance for maintaining the safety of a nuclear power plant. With regard to the structural integrity assessment for a cracked and/or flawed tube, most of the preceding researches have been focused on the deterministic approach. However, due to inherent uncertainties such as material property, crack/flaw sizing and so on, the deterministic approach can not produce realistic result. In this research, therefore, a computer program for probabilistic integrity assessment was developed in MS Windows environment. This program covers axial crack, circumferential crack and wear flaw, and can be used in practical integrity assessment of steam generator tube.
오창균(C.K. Oh),심희진(H.J. Shim),김현수(H.S. Kim),정성규(S.G. Jung),허남수(N.S. Huh) 대한기계학회 2012 대한기계학회 춘추학술대회 Vol.2012 No.11
Threaded members have been regarded as an important structural elements and can influence significantly the strength and endurance of the whole structure. Further, in terms of crack initiation and propagation, threaded members have to maintain the enough safety margin under variable loads during operation. Therefore it is necessary to perform an accurate safety analysis for the component integrity and reliability. In this study, an integrity evaluation procedure for bolting products in the component supports of the operating nuclear power plants was presented. In addition, detailed finite element analyses were performed for component support in order to determine actual stresses under the preload condition. Based on these results, the safety margin for a proloaded threaded fastener was evaluated for two kinds of operating conditions. Analysis results show that the threaded fastener has enough safety margin for both normal and accident conditions.