Y-JET 2-유체 분무노즐 내부유동의 모델링

인왕기,이상용,송시홍,In, Wang-Kee,Lee, Sang-Yong,Song, Si-Hong 대한기계학회 1993 대한기계학회논문집 Vol.17 No.7

A simplified one-dimensional analysis has been performed to predict the local pressure distributions in Y-Jet twin-fluid atomizers. Fluid compressibility was considered both in the gas(air) and two-phase(mixing) ports. The annular-mist flow model was adopted to analyze the flow in the mixing port. A series of experiments also has been performed; the results show that the air flow rate increases and the liquid flow rate decreases with the increase of the air injection pressure and/or with the decrease of the liquid injection pressure. From the measured injection pressures and flow rates, the appropriate constants for the correlations of the pressure loss coefficients and the rate of drop entrainment were decided. The local pressures inside the nozzle by prediction reasonably agree with those by the experiments.

지지격자가 있는 봉다발 구조에서 압력 손실에 대한 실험적 연구

이치영(Chi Young Lee),신창환(Chang Hwan Shin),박주용(Ju Yong Park),오동석(Dong Seok Oh),인왕기(Wang Kee In) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.10

In the present experimental study, OFEL (Omni Flow Experimental Loop) was built up, and then the friction factor in rod bundle and the pressure loss coefficient of spacer grid were investigated in range of Re=40,000?200,000. As a test section, 25 smooth rods of 9.48 mm in diameter and 2 m in length was prepared, and installed in 5×5 square array in a square channel. In this case, P/D (Pitch-to-Diameter ratio) was 1.35. In this work, three kinds of spacer grids were tested: SG0MV (Spacer Grid without Mixing Vane), SG1MV (Spacer Grid with 1-pair Mixing Vane) and SG2MV (Spacer Grid with 2-pair Mixing Vane). In the bare rod bundle (i.e., no spacer grid), the measured friction factors were in good agreement with the previous correlations proposed by Blasius and McAdams. Among the spacer grids tested in this study, the SG2MV appeared the largest friction factor in rod bundle and pressure loss coefficient, which were approximately 0.019 and 0.94 at Re=200,000, respectively.

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.

가열 봉다발의 난류 열전달에 대한 전산유체역학 해석

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).

가압경수로 핵연료집합체의 유동혼합과 열전달 실험 및 수치 해석

인왕기(Wang Kee In) 대한기계학회 2021 대한기계학회 논문집. Transactions of the KSME. C, 산업기술과 혁신 Vol.9 No.2

가압경수로 원통형 핵연료집합체와 이중냉각 환형핵연료집합체를 각각 모의한 표준 및 협소 4x4 봉 다발에서의 유동혼합과 열전달 특성을 실험과 CFD 방법으로 분석하였다. 실험 결과와 CFD 계산 결과 모두 부수로 중심 영역에서 큰 회전유동과 간극 영역에서 교차류가 발생하는 것을 알 수 있다. 표준 봉 다발의 경우 CFD 결과들은 실험 결과에 비해 다소 큰 온도 변화를 예측하고 있으며, 특히 간극에서의 온도가 큰 차이를 보이고 있다. 혼합날개의 영향으로 지지격자 하류에서 봉 표면 온도가 감소하는 즉 열전달이 증가하는 것을 실험 및 CFD 결과가 잘 보여주고 있다. Experiments and CFD calculations were performed to investigate flow mixing and heat transfer in regular and tight-lattice rod bundles, which simulate cylindrical fuel assembly and dual-cooled annular fuel assembly for pressurized water reactor, respectively. The experiments and CFD calculations showed a large swirl in core region of subchannel and crossflow in gap region. The CFD prediction showed somewhat larger variation of rod temperature than the experimental result and significantly large discrepancy particularly in rod gap. The experimental and CFD results showed the decrease of rod temperature, i.e., the increase of heat transfer downstream of mixing-vane grid.

경수로핵연료 열수력 연구개발 분석 및 연산학 협력 성과

인왕기(Wang Kee In),신창환(Chang Hwan Shin),이치영(Chi Young Lee),이찬(Chan Lee),전태현(Tae Hyun Chun),오동석(Dong Seok Oh) 대한기계학회 2016 大韓機械學會論文集B Vol.40 No.12

가압경수로에 장전되는 핵연료집합체는 연료 봉 다발과 지지격자 및 상하단 고정체로 구성되어 있다. 고온 고압의 냉각수는 원자로 하부로 유입되어 연료 봉 사이로 형성된 부수로를 따라 노심 상부로 흐른다. 경수로핵연료의 주요 열수력 성능인자는 정상운전시 압력강하 및 임계열속이며 사고시에는 급랭 시간이다. 한국원자력연구원에서는 경수로핵연료의 성능을 향상시키고 국산화를 위해 고성능 경수로핵연료, 이중냉각 핵연료 및 사고저항성 핵연료를 개발하였다. 경수로핵연료의 열수력 핵심기술을 개발하기 위해 압력강하실험, 난류 유동혼합/열전달 실험, 임계열속 및 급랭 시험을 수행하였으며 전산유체역학 방법도 활용하였다. 더불어 사용후핵연료의 임시저장을 위한 건식저장 용기의 열유동에 대한 전산유체해석을 수행하였다. 한편, 경수로핵연료의 열수력 기반기술을 개발하고 실용화를 위해 대학 및 산업체와 협력연구도 진행하였다. The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

가압경수로용 이중냉각핵연료 열수력 특성 실험 및 CFD 해석

인왕기(Wang-Kee In),신창환(Chang-Hwan Shin),박주용(Ju-Yong Park),오동석(Dong-Seok Oh),전태현(Tae-Hyun Chun) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.5

온도측정을 통한 정사각 채널내 단일봉과 평행한 유동의 부수로간 열혼합에 관한 실험적 연구

박주용(Ju Yong Park),신창환(Chang Hwan Shin),이치영(Chi Young Lee),인왕기(Wang Kee In) 대한기계학회 2013 대한기계학회 춘추학술대회 Vol.2013 No.12

Dual-cooled annular fuel is being developed in KAERI. Dual-cooled annular fuel has small gaps between fuel rods compared to circular fuel rod. Because of this, a pulsation occurred between subchannels. In this paper, thermal mixing by pulsation occurred at the between subchannels in parallel flow to single rod in square channel has been performed through temperature measurement. With decreased of the distance between rod and wall, the thermal mixing effect was increased. In addition, the thermal mixing effect at subchannel was increased by mixing vane

지지격자가 있는 봉다발과 축방향으로 평행한 유동의 압력손실에 관한 실험적 연구

이치영(Chi Young Lee),신창환(Chang Hwan Shin),박주용(Ju Yong Park),인왕기(Wang Kee In) 대한기계학회 2012 大韓機械學會論文集B Vol.36 No.7

지지격자가 있는 봉다발과 축방향으로 평행한 유동에서, 봉다발 마찰계수와 지지격자 손실계수를 평가하였다. 시험부는 외경 9.5 mm, 길이 2000 mm 인 봉 25 개를 5×5 정사각 구조로 배열하여 제작하였으며, 봉 중심간 거리와 봉 외경의 비는 1.35 였다. 지지격자로는 plain 지지격자, split-vane 지지격자, hybrid-vane 지지격자를 이용하였다. 지지격자가 없는 봉다발의 마찰계수는 기존 상관식과 비교적 잘 일치하였다. 지지격자가 있는 봉다발 실험의 경우, hybrid-vane 지지격자에서 봉다발 마찰계수 및 지지격자손실계수가 가장 크게 측정되었으며, 이는 지지격자의 유동단면 막음비 증가와 혼합날개 형상에 의한 유동 교란이 증가되기 때문인 것으로 판단된다. Re=5×105 조건에서 plain 지지격자, split-vane 지지격자, hybrid-vane 지지격자의 손실계수는 약 0.79, 0.80, 0.88 로 예측되었다. The friction factor in a rod bundle and the loss coefficient at a spacer grid were examined. As a test section, 25 smooth rods, 9.5 mm in diameter and 2000 mm in length, were prepared and installed in a 5 × 5 square array in a square channel. In this case, the P/D (Pitch-to-Diameter ratio) was 1.35. In this work, plain (i.e., no mixing vanes), split-vane, and hybrid-vane spacer grids were tested. In a bare rod bundle (i.e., no spacer grid), the measured friction factors were in good agreement with the previous correlations. Among the spacer grids tested, the hybrid-vane spacer grid presented the largest friction factor in the rod bundle and loss coefficient. This may be because of the flow pattern change induced by large relative plugging of the flow cross section and mixing vane geometry. At Re = 5 × 105, the predicted loss coefficients of plain, splitvane, and hybrid-vane spacer grids were approximately 0.79, 0.80, and 0.88, respectively.

환형 연료의 열 및 유량분리 수치해석

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.