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이준(J. Lee),강한옥(H. O. Kang),서재광(J. K. Seo),유승엽(S. Y. Ryu),윤주현(J. Yoon),김긍구(K. K. KIM) 대한기계학회 2003 대한기계학회 춘추학술대회 Vol.2003 No.11
In this paper, the thermal hydraulic design characteristics of the power conversion system for SMART 330 MWt plant which the once-through helically-coiled steam generators have been installed was analyzed. The feedwater temp., total feedwater flow rate, steam pressure, total steam flow rate and steam temp. was analyzed vs. steam generator power on conditions of a constant feedwater temp. and steam pressure, and was compared with the commercial nuclear power plants. After the present analysis, it was evaluated that at the range of low power(about below 20%) the steam temp. is nearly same as the core outlet temp., and at the range of meddle power(about between 20 ~ 80%) it show gradually a upward trend, and at the range of high power(about between 80 ~ 100%) it show gradually a downward trend. These phenomena result from the once-through steam generator's model which consists of the 3 different type of regions, i.e., subcooled, boiling, and superheated region.
일체형원자로에서 냉각재펌프의 전력측정을 이용한 실시간 유량산정 방법에 관한 연구
이준(J. Lee),윤주현(J.-H. Yoon),지성균(S.-Q. Zee) 한국유체기계학회 2003 유체기계 연구개발 발표회 논문집 Vol.- No.-
It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of coolant pump power has been introduced in this study. Up to now, we did not found out a precedent which the coolant pump power is used for the real-time flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the real-time flow-rate calculation method by the measurement of coolant pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs, So, it has been concluded that it is possible to calculate the real-time flow-rate by the measurement of pump motor inputs. In addition, the compensation for a above new method. can be made by HBM being now used in the commercial nuclear power plants.
연구로의 PCS discharge header 설계를 위한 CFD 해석
서경우(K.W. Seo),윤현기(H.G. Yoon),지대영(D.Y. Chi),윤주현(J.H. Yoon) 한국전산유체공학회 2012 한국전산유체공학회 학술대회논문집 Vol.2012 No.5
The heat generated from the core of a research reactor is removed by a primary cooling system (PCS). Most reactors of relatively low power are designed using an open type reactor and pool. Thus, the primary coolant can be discharged to the reactor pool. However, the discharged coolant toward the top of the open type pool should be minimized for operator safety because it contains many kinds of radio-nuclides. This means that it is important for a PCS discharge header to be designed for minimizing bulk flow in the pool. ANSYS-CFD was employed to design a PCS discharge header and to analyze the flow characteristics in the pool. Several discharge headers were simulated with a similar scaled geometry and the boundary conditions of an actual research reactor. To minimize the bulk flow toward the top of the pool, it was analyzed that the magnitude and direction of velocity at the hole of the discharge header should be reduced. The results showed that the CFD analysis can be used as a tool for designing and modifying the PCS discharge according to various types of reactor pools.
박용철(Y.C. Park),윤현기(H.G. Yoon),서경우(K.W. Seo),지대영(D.Y. Chi),윤주현(J.H. Yoon) 한국전산유체공학회 2012 한국전산유체공학회 학술대회논문집 Vol.2012 No.5
In nuclear power plant, the reactor cooling system has maintained high-Reynolds-number flow above 1E+07 to cool a heat generated by the reactor. To minimize uncertainty for flow calibration, it is necessary to install a flow simulation system to maintain the high-Reynolds-number flow. Y-connection is selected to connect four (4) parallel high flow circulation pumps for minimizing system pressure loss. This paper describes the characteristics for Y-connection by computer flow simulation. It was confirmed through the results that the pressure loss of the Y-connection was lower than that of T-connection, and that the pressure loss of the connection was similar sixty degree (60°) and below of connection angle.
서경우(K.W. Seo),윤현기(H.G. Yoon),김성훈(S.H. Kim),지대영(D.Y. Chi),윤주현(J.H. Yoon) 한국전산유체공학회 2013 한국전산유체공학회 학술대회논문집 Vol.2013 No.5
In a research reactor, the large pool water is at the heart of safety because a thermal design flow during a normal operation or a flow path for a natural circulation during an abnormal reactor trip shall be supplied for core cooling. Most research reactors include a reactor in the enormous pool, and are designed using an open type reactor and pool. Thus, the primary coolant containing many kinds of radio-nuclides can be mixed to the reactor pool. Because operators can operate near the pool top during a normal operation, the bulk flow toward the top of the open type pool should be designed to be minimized.