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전건호(Kun-Ho Chun),전태현(Tae-Hyun Chun),인왕기(Wang-Kee In),양용식(Yong-Sik Yang) 대한기계학회 2006 대한기계학회 춘추학술대회 Vol.2006 No.11
The objective of this task is to conduct sensitivity of gap conductance for annular fuel rod. In the annular fuel with an internal cooling hole, part of the heat rate generated in a fuel pellet is transferred to the external coolant in the open fuel pin array and the remaining part proceeds into the coolant flowing through the inner channel. It is important to achieve the accurate determination of the heat flow split because it affects coolant enthalpy and Departure from Nuclear Boiling Ratio (DNBR) in each channel. In particular, when most of the heat rate leans to either inner or outer channel, it is out of thermal equilibrium. For minimizing thermal imbalance, placing another gap in the pellet is introduced. The heat flow distribution between internal and external channels as well as fuel and cladding temperature profiles is finely calculated for the annular fuel rod with or without the fuel gap between inner and outer pellets.
양용식(Yong Sik Yang),전태현(Tae Hyun Chun),신창환(Chang Hwan Shin),송근우(Kun Woo Song) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.11
The surface heat flux of nuclear fuel rod is the most important factor which can affect safety of reactor and fuel. If fuel rod surface heat flux exceeds the CHF(Critical Heat Flux), fuel can be damaged. In case of double cooled annular fuel, which is under developing, contains two coolant channels. Therefore, a generated heat in the fuel pellet can move to inner or outer channel and heat flow direction is decided by both sides heat resistance which varied by dimension and material property change which caused by temperature and irradiation. The new program(called DUO) was developed. For the calculation of surface heat flux, a both sides convection by inner/outer coolant, s gap temperature jump and conduction in the fuel are modeled. Especially, temperature and time dependent fuel dimension and material property change are considered during the iteration. A sample calculation result shows that the DUO program has sufficient performance for annular fuel thermal hydraulics design.
이중냉각연료에서 지지격자의 압력손실에 대한 엔탈피 증가
전건호(Kun-Ho Chun),전태현(Tae-Hyun Chun),신창환(Chang-Hwan Shin) 대한기계학회 2007 대한기계학회 춘추학술대회 Vol.2007 No.5
A dual side cooling annular fuel having internal and external coolant channels has many advantages basically due to low fuel temperature and high DNBR margin, which can make a significant increase of core power density possible. So recently a 12×12 square annular fuel array was proposed for the fuel assembly to be reloaded without structural interference with operating reactors of OPR-1000s. Even through the inherent potential of the annular fuel on the high power density, it may be seriously eroded in the case of a severe unbalanced mass flux split to the internal and external channels in standpoint of DNB. Mass flux split is determined pressure drop characteristics between inner and outer channels. The spacer grids binding fuel array influence greatly the pressure drop in outer channels and the mass flux split. As an important factor of DNB behavior, the enthalpy differences at both channel exits were evaluated using the mass flux splits.
CFX 코드를 이용한 핵연료 봉다발에서의 단상 및 2 상유동 열전달 분석
인왕기(Wang-Kee In),황대현(Dae-Hyun Hwang),전태현(Tae-Hyun Chun) 한국유체기계학회 2006 유체기계 연구개발 발표회 논문집 Vol.- No.-
A three-dimensional numerical analysis was performed to examine the coolant flow mixing and the heat transfer in a nuclear fuel assembly with a flow-mixing device on the grid spacer. The nuclear fuel assembly used in pressurized water reactor(PWR) is a square rod bundle which is supported by a spacer grid. The coolant flows axially through the subchannels formed between the rods. The fuel spacer affects the coolant flow distribution in the fuel rod bundle, and so the spacer geometry has a strong influence on a bundle’s thermal-hydraulic characteristics, such as the critical heat flux and pressure drop. In particular, the flow deflecting vanes on the grid spacer can improve the departure from a nucleate boiling(DNB) performance by increasing the coolant mixing and the rod heat transfer downstream of the vanes. The computational fluid dynamics(CFD) code, CFX-10 is used to perform the three-dimensional analysis of the flow mixing and heat transfer in a rod bundle with and without a flow-mixing vane. The CFX-10 code is also used to predict void distribution in a boiling water reactor(BWR) rod bundle which is an international benchmark problem.
인왕기(In Wang-Kee),오동석(Oh Dong-Seok),전태현(Chun Tae-Hyun) 대한기계학회 2003 대한기계학회 춘추학술대회 Vol.2003 No.11
A CFD analysis has been performed to investigate turbulent heat transfer in a triangular rod bundle with a<br/> pitch-to-diameter ratio(P/D) of 1.06. Anisotropic turbulence models predicted the turbulence-driven secondary<br/> flow in the triangular subchannel and the distributions of time mean velocity and temperature showing<br/> significantly improved agreement with the measurements over the linear standard k ε . model. The<br/> anisotropic turbulence models predicted turbulence structure in large flow region fairly well but could not<br/> predict the very high turbulent intensity of azimuthal velocity observed in narrow flow region(gap).
CFD를 이용한 이중냉각핵연료의 봉단마개 측면오리피스를 통과하는 내측수로 유동에 대한 연구
곽영균(Young Kyun Kwack),신창환(Chang Hwan Shin),이치영(Chi Young Lee),박주용(Ju Yong Park),전태현(Tae Hyun Chun),인왕기(Wang Kee In) 대한기계학회 2012 대한기계학회 춘추학술대회 Vol.2012 No.11
A bottom end plug with side orifices was developed for sufficient coolant flowing to inner channel in the case of complete entrance blockage of the inner channel. In this event, the flow rate of the inner channel should be greater than 60% of normal condition in order to avoid a damaged condition of the nuclear fuel. The flow rate and the pressure loss coefficient were obtained through simulation analysis of single rod for the end plug. According to the blockage rate of the outer channel, flow pattern and pressure loss coefficient of the side hole were investigated.