Void Fraction and Pressure Drop in Two-Phase Equilibrium Flows in a Vertical 2*3 Rod Bundle Channel-Assessment of Correlations against the Present Subchannel Data

Two-fluid model can simulate two phase flow less computational cost than inter-face tracking method and particle interaction method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as a rod bundle. Japan Atomic Energy Agency (JAEA) develops three dimensional two-fluid model analysis code ACE-3D, which adopts boundary fitted coordinate system in order to simulate complex shape channel flow. In this paper, boiling two-phase flow analysis in a tight lattice rod bundle is performed by ACE-3D code. The parallel computation using 126CPUs is applied to this analysis. In the results, the void fraction, which distributes in outermost region of rod bundle, is lower than that in center region of rod bundle. At height z = 0.5 m, void fraction in the gap region is higher in comparison with that in center region of the subchannel. However, at height of z = 1.1m, higher void fraction distribution exists in center region of the subchannel in comparison with the gap region. The tendency of void fraction to concentrate in the gap region at vicinity of boiling starting point, and to move into subchannel as water goes through rod bundle, is qualitatively agreement with the measurement results by neutron radiography. To evaluate effects of two-phase flow model used in ACE-3D code, numerical simulation of boiling two-phase in tight lattice rod bundle with no lift force model (neglecting lift force acting on bubbles) is also performed. From the comparison of numerical results, it is concluded that the effects of lift force model are not so large on overall void fraction distribution in tight lattice rod bundle. However, higher void fraction distribution in center region of the subchannel was not observed in this simulation. It is concluded that the lift force model is important for local void fraction distribution in rod bundles.

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Void fraction measurement of the air-water two-phase flow in the sub-channel of a rod bundle geometry based on an impedance meter

Annals of Nuclear Energy ◽

10.1016/j.anucene.2018.02.018 ◽

2018 ◽

Vol 115 ◽

pp. 480-486 ◽

Cited By ~ 1

Author(s):

Bin Yu ◽

Wenxiong Zhou ◽

Liangming Pan ◽

Hang Liu ◽

Quanyao Ren ◽

...

Keyword(s):

Void Fraction ◽

Two Phase Flow ◽

Phase Flow ◽

Two Phase ◽

Rod Bundle ◽

Fraction Measurement ◽

Impedance Meter

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Void fraction measurements of steam–water two-phase flow in vertical rod bundle: Comparison among different techniques

Experimental Thermal and Fluid Science ◽

10.1016/j.expthermflusci.2019.109881 ◽

2019 ◽

Vol 109 ◽

pp. 109881 ◽

Cited By ~ 6

Author(s):

Miao Gui ◽

Zhaohui Liu ◽

Bo Liao ◽

Teng Wang ◽

Yong Wang ◽

...

Keyword(s):

Void Fraction ◽

Two Phase Flow ◽

Phase Flow ◽

Two Phase ◽

Rod Bundle

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Two-Phase Flow Through Square and Circular Microchannels—Effects of Channel Geometry

Journal of Fluids Engineering ◽

10.1115/1.1777227 ◽

2004 ◽

Vol 126 (4) ◽

pp. 546-552 ◽

Cited By ~ 59

Author(s):

Peter M.-Y. Chung ◽

Masahiro Kawaji ◽

Akimaro Kawahara ◽

Yuichi Shibata

Keyword(s):

Pressure Drop ◽

Void Fraction ◽

Two Phase Flow ◽

Flow Characteristics ◽

Separated Flow ◽

Phase Flow ◽

Two Phase ◽

Channel Geometry ◽

Frictional Pressure Drop ◽

Channel Shape

An adiabatic experiment was conducted to investigate the effect of channel geometry on gas-liquid two-phase flow characteristics in horizontal microchannels. A water-nitrogen gas mixture was pumped through a 96 μm square microchannel and the resulting flow pattern, void fraction and frictional pressure drop data were compared with those previously reported by the authors for a 100 μm circular microchannel. The pressure drop data were best estimated using a separated-flow model and the void fraction increased non-linearly with volumetric quality, regardless of the channel shape. However, the flow maps exhibited transition boundaries that were shifted depending on the channel shape.

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Two-Phase Void Fraction and Pressure Drop in Horizontal Crossflow Across a Tube Bundle

Journal of Fluids Engineering ◽

10.1115/1.2819638 ◽

1998 ◽

Vol 120 (1) ◽

pp. 140-145 ◽

Cited By ~ 14

Author(s):

G. P. Xu ◽

K. W. Tou ◽

C. P. Tso

Keyword(s):

Pressure Drop ◽

Void Fraction ◽

Mass Velocity ◽

Flow Model ◽

Tube Bundle ◽

Diameter Ratio ◽

Two Phase ◽

Friction Pressure ◽

Oil Flow ◽

Friction Pressure Drop

Void fraction and friction pressure drop measurements were made for an adiabatic, horizontal two-phase flow of air-water, air-oil across a horizontal in-line, 5 × 20 tube bundle with pitch-to-diameter ratio, P/D, of 1.28. For both air-water and air-oil flow, the experimental results showed that the average void fraction were less than the values predicted by a homogenous flow model, but were well correlated with the Martinelli parameter Xtt and liquid-only Froude number FrLO. The two-phase friction multiplier data exhibited an effect of flow pattern and mass velocity, and they could be well-correlated with the Martinelli parameter.

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INVESTIGATION ON VOID FRACTION FOR TWO-PHASE FLOW PRESSURE DROP OF EVAPORATIVE R-290 IN HORIZONTAL TUBE

Jurnal Teknologi ◽

10.11113/jt.v78.9590 ◽

2016 ◽

Vol 78 (8-4) ◽

Cited By ~ 1

Author(s):

Agus Sunjarianto Pamitran ◽

Sentot Novianto ◽

Normah Mohd-Ghazali ◽

Nasruddin Nasruddin ◽

Raldi Koestoer

Keyword(s):

Pressure Drop ◽

Void Fraction ◽

Two Phase Flow ◽

Flow Boiling ◽

Homogeneous Model ◽

Saturation Temperature ◽

Horizontal Tube ◽

Phase Flow ◽

Two Phase ◽

Experimental Conditions

Two-phase flow boiling pressure drop experiment was conducted to observe its characteristics and to develop a new correlation of void fraction based on the separated model. Investigation is completed on the natural refrigerant R-290 (propane) in a horizontal circular tube with a 7.6 mm inner diameter under experimental conditions of 3.7 to 9.6 °C saturation temperature, 10 to 25 kW/m2 heat flux, and 185 to 445 kg/m2s mass flux. The present experimental data was used to obtain the calculated void fraction which then was compared to the predicted void fraction with 31 existing correlations. A new void fraction correlation for predicting two-phase flow boiling pressure drop, as a function of Reynolds numbers, was proposed. The measured pressure drop was compared to the predicted pressure drop with some existing pressure drop models that use the newly developed void fraction model. The homogeneous model of void fraction showed the best prediction with 2% deviation

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Comparative Study in the Stability Analysis Code of LAPUR5.2 and LAPUR6.0 for the Kuosheng NPP

Volume 3: Thermal-Hydraulics; Turbines, Generators, and Auxiliaries ◽

10.1115/icone20-power2012-54183 ◽

2012 ◽

Cited By ~ 1

Author(s):

Chang-Lung Hsieh ◽

Guan-Yu Chen ◽

Chunkuan Shih ◽

Jong-Rong Wang ◽

Hao-Tzu Lin

Keyword(s):

Stability Analysis ◽

Pressure Drop ◽

Void Fraction ◽

Boiling Water ◽

Boiling Water Reactors ◽

Two Phase ◽

Flow Area ◽

Comparison Results ◽

Fuel Assemblies ◽

Water Reactors

Under certain conditions, boiling water reactors (BWRs) would be susceptible to couple neutron-thermalhydraulic instability. It is important to predict such potential problems as early as possible and prevent the core instability from happening. In each BWR reload core design, fuel vendors are required to provide instability boundaries on power/flow map to assure safety operation of the nuclear reactor. In Taiwan, a LAPUR5.2 methodology had been adapted to build up the remarkable analysis mode for different types BWRs to verify vendor’s results. However, with upgrading nuclear safety technology, most of boiling water reactors has been adopting partial length fuel assemblies to reduce two-phase pressure drop and void fraction, to improve reactor stability. The question is that LAPUR5.2 methodology cannot precisely analysis stability characteristics from the variation of flow area in fuel assemblies. From the reasons of upgrading stability analysis, a LAPUR6.0 methodology had built to do the related researches. This research was based on a comparison study between LAPUR5.2 and LAPUR6.0 to realize the major differences and their effects on stability characteristics. According to the comparison results for Kuosheng Nuclear Power Plant Unit 2 Cycle 21 reload design, it shows that LAPUR6.0 can completely present pressure drop, void fraction and density reactivity coefficient from the changing of flow area and fuel spacers.

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