Modeling of distribution parameter, void fraction covariance and relative velocity covariance for upward steam-water boiling flow in vertical rod bundle

Abstract This paper describes the computational fluid dynamics (CFD) methodology to simulate the boiling flow in a typical Pressurized Water Reactor (PWR) 5 ⨯ 5 rod bundle. The method includes the Eulerian-Eulerian two-fluid model coupled with the improved wall heat partitioning model. The NUPEC PWR Subchannel and Bundle Test (PSBT) International Benchmark are used for validation. The simulated surface averaged void fraction agree well with the experimental data, which indicate the promising application of the present method for modeling the boiling flow in the fuel rod bundle. The main emphasis of current research has been given to the analysis of the phase distribution around and downstream the spacer grid, the effect of the spacer grid structure, including the mixing vanes, the springs and the dimples on the void fraction distribution is investigated. The findings can contribute to a better understanding of three dimensional flow boiling characteristics and can be used to assist in optimizing the spacer grid.

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Void Fraction Distribution and Mass/Energy Transfer Characteristics in Rod Bundle

10.1615/tfec2018.cmd.022201 ◽

2018 ◽

Author(s):

Jun-Yi Zhang ◽

Xiao Yan ◽

Ze-Jun Xiao ◽

Yong-Ze Xu

Keyword(s):

Energy Transfer ◽

Void Fraction ◽

Mass Energy ◽

Transfer Characteristics ◽

Rod Bundle ◽

Void Fraction Distribution ◽

Fraction Distribution

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Numerical Simulation of Boiling Two-Phase Flow in Tight-Lattice Rod Bundle by 3-Dimensional Two-Fluid Model Code ACE-3D

Volume 3: Thermal Hydraulics; Instrumentation and Controls ◽

10.1115/icone16-48715 ◽

2008 ◽

Author(s):

Hiroyuki Yoshida ◽

Takeharu Misawa ◽

Kazuyuki Takase

Keyword(s):

Void Fraction ◽

Lift Force ◽

Fluid Model ◽

Phase Flow ◽

Two Phase ◽

Rod Bundle ◽

Fraction Distribution ◽

Two Fluid Model ◽

Tight Lattice ◽

Two Fluid

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

JSME International Journal Series B ◽

10.1299/jsmeb.49.279 ◽

2006 ◽

Vol 49 (2) ◽

pp. 279-286 ◽

Cited By ~ 5

Author(s):

Michio SADATOMI ◽

Keiko KANO ◽

Akimaro KAWAHARA ◽

Naoki MORI

Keyword(s):

Phase Equilibrium ◽

Pressure Drop ◽

Void Fraction ◽

Two Phase ◽

Rod Bundle

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Analysis of RELAP5 Drift-Flux Model for Vertical Subcooled Boiling Flow at Low Pressure Conditions

10.1115/imece2000-1535 ◽

2000 ◽

Author(s):

Boštjan Končar ◽

Ivo Kljenak ◽

Borut Mavko

Keyword(s):

Void Fraction ◽

Drift Velocity ◽

Bubbly Flow ◽

Low Pressure ◽

Subcooled Boiling ◽

Drift Flux Model ◽

Drift Flux ◽

Boiling Flow ◽

Flux Model ◽

Relap5 Code

Abstract The RELAP5/MOD3.2.2 Gamma code was assessed against low pressure boiling flow experiments performed by Zeitoun and Shoukri (1997) in a vertical annulus. The predictions of subcooled boiling bubbly flow showed that the present version of the RELAP5 code underestimates the void fraction increase along the flow and strongly overestimates the vapor drift velocity. It is shown that in the calculations, a higher vapor drift velocity causes a lower interphase drag and may be a possible reason for underpredicted void fraction development. A modification is proposed, which introduces the replacement of the EPRI drift-flux formulation, which is currently incorporated in the RELAP5 code, with the Zuber-Findlay (1965) drift-flux model for the experimental low pressure conditions of the vertical bubbly flow regime. The improved experiment predictions with the modified RELAP5 code are presented and analysed.

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