http://chineseinput.net/에서 pinyin(병음)방식으로 중국어를 변환할 수 있습니다.
변환된 중국어를 복사하여 사용하시면 됩니다.
열전도 물체가 존재하는 캐비티내 자연대류 열전달에 대한 수치적 연구
명현국(H. K. Myong),전태현(T. H. Chun) 한국전산유체공학회 2005 한국전산유체공학회지 Vol.10 No.3
The present study numerically investigates the natural convection heat transfer in a 2-D square cavity containing a centered heat conducting body. Special emphasis is given to the influences of the Rayleigh number, the dimensionless conducting body size, and the ratio of the thermal diffusivity of the body to that of the fluid on the natural convection heat transfer in overall concerned region. The analysis reveals that the fluid flow and heat transfer processes are governed by all of them. Results for isotherms, vector plots and wall Nusselt numbers are reported for Pr = 0.71 and relatively wide ranges of the other parameters. Heat transfer across the cavity, in comparison to that in the absence of a body, are enhanced (reduced) in general by a body with a thermal diffusivity ratio less (greater) than unity. It is also found that the heat transfer attains a minimum as the body size is increased with a thermal diffusivity ratio greater than unity.
OPR-1000 원자로의 출력증강을 위한 18×18 원통형 핵연료의 열수력적 타당성 평가
신창환(C. H. Shin),김효일(H. I. Kim),인왕기(W. K. In),전태현(T. H. Chun) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.11
The thermal hydraulic analysis of the 18×18 solid fuel assembly has been carried out for the power uprate of OPR-1000. The suggested 18x18 solid fuel assembly has a structural compatibility for reloading to operating PWR reactors of OPR-1000. In the thermal hydraulic analysis, the mixing effect between the neighboring channels has to be carried out in a subchannel analysis. A subchannel analysis code, MATRA has been developed by KAERI. The main parameters for the thermal hydraulic design, such as a pressure drop and DNBR, and the maximum temperature in a fuel pellet have been estimated. The 18×18 solid fuel in the 120% power uprate showed an increased pressure drop and a similar DNBR behavior. The peak temperature in the fuel centerline, however, was slightly higher than that of the 16×16 solid fuel assembly for the normal operation condition. The peak temperature of a fuel pellet in the solid fuel should be seriously considered for increasing power density.
자유낙하에 의한 경수로용 핵연료집합체의 동적 거동 해석
윤경호(K-H Yoon),김형규(H-K Kim),전태현(T-H Chun),안창기(C-K Ahn) 대한기계학회 2006 대한기계학회 춘추학술대회 Vol.2006 No.11
The objective of this research is to propose the methodology that could predict the dynamic failure behavior on the pressurized water reactor (PWR) fuel assembly structure for the reactor coolant pump (RCP) over-speed transient event. To perform this objective, the finite element analysis method for predicting the axial impact behavior on the fuel structure is established by a commercial finite element codes ANSYS and DYNA-3D. In this FE analysis method, appropriate boundary conditions and impact loading conditions are applied to simulate the actual core conditions. The drop impact analysis of a fuel assembly for PWR power plant is executed by the finite element analysis method. The analysis results are compared with the previous experimental results. The impact force results differed from the analysis condition depending on how many fuel rods slipped down to the bottom nozzle. And the joint and connection parts between components of a fuel assembly are very important for determining the stiffness of it. In the analysis, the fuel assembly experienced an impact force of approximately 111.53 kN when dropped from a height 25 mm with the fuel rods on the bottom nozzle. The secondary peak value is about 57.55 kN and the duration is about 6 milliseconds. After comparing with the previous test results, the developed finite element model and analysis procedure will be useful tool for evaluating the dynamic stiffness and strength of a fuel assembly. It is found that the joint and connection characteristics among parts are dominant factor which determine the non-linear behavior of a fuel during free fall event.
과냉 비등유동에 대한 CFD 모의계산에서의 벽 인접격자 영향
인왕기(W.K. In),신창환(C.H. Shin),전태현(T.H. Chun) 한국전산유체공학회 2010 한국전산유체공학회 학술대회논문집 Vol.2010 No.5
A multiphase CFD analysis is performed to investigate the effect of near-wall grid for simulating a subcooled boiling flow in vertical tube. The multiphase flow model used in this CFD analysis is the two-fluid model in which liquid(water) and vapor(steam) are considered as continuous and dispersed fluids, respectively. A wall boiling model is also used to simulate the subcooled boiling heat transfer at the heated wall boundary. The diameter and heated length of tube are 0.0154 m and 2 m, respectively. The system pressure in tube is 4.5 MPa and the inlet subcooling is 60 K. The near-wall grid size in the non-dimensional wall unit (y<SUB>w</SUB><SUP>+</SUP>) was examined from 64 to 172 at the outlet boundary. The CFD calculations predicted the void distributions as well as the liquid and wall temperatures in tube. The predicted axial variations of the void fraction and the wall temperature are compared with the measured ones. The CFD prediction of the wall temperature is shown to slightly depend on the near-wall grid size but the axial void prediction has somewhat large dependency. The CFD prediction was found to show a better agreement with the measured one for the large near-wall grid, e.g., y<SUB>w</SUB><SUP>+</SUP> > 100.
벽 비등모델을 이용한 과냉비등 유동에 대한 CFD 모의계산에서 벽 인접격자의 영향
인왕기(W.K. In),신창환(C.H. Shin),전태현(T.H. Chun) 한국전산유체공학회 2010 한국전산유체공학회지 Vol.15 No.3
A multiphase CFD analysis is performed to investigate the effect of near-wall grid for simulating a subcooled boiling flow in vertical tube. The multiphase flow model used in this CFD analysis is the two-fluid model in which liquid(water) and gas(vapour) are considered as continuous and dispersed fluids, respectively. A wall boiling model is also used to simulate the subcooled boiling heat transfer at the heated wall boundary. The diameter and heated length of tube are 0.0154 m and 2 m, respectively. The system pressure in tube is 4.5 ㎫ and the inlet subcooling is 60 K. The near-wall grid size in the non-dimensional wall unit for lqiuid phase (y?<SUB>w,l</SUB>) was examined from 101 to 313 at the outlet boundary. The CFD calculations predicted the void distributions as well as the liquid and wall temperatures in tube. The predicted axial variations of the void fraction and the wall temperature are compared with the measured ones. The CFD prediction of the wall temperature is shown to slightly depend on the near-wall grid size but the axial void prediction has somewhat large dependency. The CFD prediction was found to show a better agreement with the measured one for the large near-wall grid, e.g., y?<SUB>w,l</SUB> > 300 at the tube exit.
박주용(J.Y. Park),신창환(C.H. Shin),이치영(C.Y. Lee),곽영균(Y.K. Kwack),전태현(T.H. Chun),오동석(D.S. Oh),인왕기(W.K. In) 대한기계학회 2012 대한기계학회 춘추학술대회 Vol.2012 No.11
For power uprate of Pressurized Water Reactor(PWR), a dual-cooled annular fuel is being developed in Korea Atomic Energy Research Institute(KAERI). The annular fuel rod is configured to allow the coolant flow through the inner channel as well as outer channel. Since the inner channel is isolated from the neighbor channels unlike the outer channels, an inner channel blockage is one of the principal technical issues of a dual-cooled annular fuel. As a hypothetical event, if the inner channel is completely blocked by the debris, the inner cladding may experience a rapid temperature increase because of no further coolant supply. To complement the entrance blockage of an inner channel a long end plug with side holes is conceptually suggested. The experiment is performed to estimate the flow rate through side holes at the side hole blockage event. In this paper, the inner channel flow rate by the side holes was measured and the loss coefficient of side holes was evaluated.
가압경수로 이중냉각핵연료의 내측수로 막힘에 대한 전산유체역학 해석
인왕기(W.K. In),신창환(C.H. Shin),박주용(J.Y. Park),오동석(D.S. Oh),이치영(C.Y. Lee),전태현(T.H. Chun) 한국전산유체공학회 2011 한국전산유체공학회 학술대회논문집 Vol.2011 No.5
A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet(UO₂) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid fuel by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.
한국원자력연구소에서 개발한 가압경수로용 핵연료 지지격자의 성능 해석 및 시험
송기남(K-N Song),윤경호(K-H Yoon),강흥석(H-S Kang),최명환(M-H Choi),전태현(T-H Chun) 대한기계학회 2004 대한기계학회 춘추학술대회 Vol.2004 No.4
KAERI has contrived 16 kinds of spacer grid shapes of its own since 1997 and applied for domestic and foreign patents. To date, KAERI has obtained US and ROK patents for 11 kinds of spacer grid shapes among them and the others are under review in USA, EC, China, and ROK. In this study, detailed performance analysis and test on two spacer grid shapes that are assumed to be the most effective candidates for the spacer grid of the next generation nuclear fuel in Korea was carried out. The result has shown that the performances of the candidates are better or not worse than those of the current spacer grid.