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김영범(Y. B. Kim),권갑주(K. J. Kwon),윤인식(I. S. Yoon),최종귀(J. K. Choi),황지혁(J. H. Hwang),손기철(K. Ch. Son),김성영(S. Y. Kim) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.11
Domestic Enforcement Regulation of the Atomic Energy Act requires the Equipment Qualification (EQ) of safety-related equipment in nuclear power plants to be included at Periodic Safety Review (PSR). The purpose of this study is to verify the integrity of AOV (Air Operated Valve) actuator by equipment qualification, which will be followed by the localization of the AOV and its actuator. The AOV actuator, whose model number is SDD-VA-1, was developed based on the requirements of the relevant ASME and ASNI Codes. The procedure for the equipment qualification was established based on the EQ Rule (10CFR50.49), Reg. Guide 1.89, IEEE 323(1983) and IEEE 382(1972). Detailed test procedure was based on that of Korea Institute of Machinery and Materials (KIMM) and the EQTR of a referenced valve was used.
김영범(Y. B. Kim),권갑주(K. J. Kwon),윤인식(I. S. Yoon),이상민(S. M. Lee),장훈(H. Jang),강영미(Y. M. Kang),김상표(S. P. Kim),박종선(J. S. Park) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.11
In domestic industry, more than 95% of high temperature and pressure control valves are being imported and particularly, all of the AOV actuators for nuclear power plants are being imported. There is, therefore, an urgent need to localize them. As a part of the effort for localization, System Design and Development (SDD) designed and manufactured main components of Q Class control valve for nuclear power plants such as body, yoke, bonnet, and actuator based on ASME code. In this study, seismic qualification was performed for the AOV actuator to validate integrity. The fundamental frequency of the equipment was identified through analysis and operating basis earthquake (OBE) and safe shutdown earthquake (SSE) test was performed according to the requirements of IEEE 344 (1987) and 382 (1996).
장훈(H Jang),김영범(Y.B Kim),권갑주(K.J Kwon),윤인식(I.S Yoon),윤재원(J.W Yoon) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.11
Unsteady state fluid flow is caused by the very high-potential energy of fluid like cavitation, flashing and hammering in the Nuclear Power Plants. The main components of a valve are damaged by them. The control valve under high pressure drop that limites velocity of fluid on 30m/s is used to solve the problems. In this case, the reduction of generating power is caused by fail to the control of fluid. In this paper, the control valve trim under high pressure drop which is called Helical Trim was developed to control fluid flow. The pressure, velocity, and movement Pattern of fluid in control valve were analyzed by using numerical analysis program which is called ANSYS CFX. It was performed to compare calculated valve flow coefficient using dynamic characteristics test facility and design data and relevance of high-precision fluid control was analyzed.