A void fraction correlation for generalized applications

Chexal, B.; Lellouche, G.S.; Horowitz, J.; Healzer, J.M.

doi:10.1016/0149-1970(92)90007-p

Cited by 130 publications

(65 citation statements)

References 9 publications

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“…The first 4 sets such as the constitutive equations of Bestion,11) Chexal et al, 12) Inoue et al, 13) and Maier and Coddington 14) have been derived from rod bundle data (mainly differential pressure measurements) and have been successfully tested against different rod bundle databases 16) in a wide range of void fraction conditions. The last set such as the constitutive equations of Hibiki et al 15) was derived for annulus channel geometry and obtained by local probe measurements.…”

Section: Survey Of Existing Work 1 One-dimensional Drift-flux Mmentioning

confidence: 99%

“…14) Chexal et al (1992) 12) This correlation is based on the Zuber-Findlay model 10) and the main modifications are focused on the distribution parameter. In their work, Chexal and Lellouche developed correlations for both upward and downward flows in vertical, inclined and horizontal pipes with different fluid types.…”

Section: Survey Of Existing Work 1 One-dimensional Drift-flux Mmentioning

confidence: 99%

See 1 more Smart Citation

Flow Structure of Subcooled Boiling Water Flow in a Subchannel of 3×3 Rod Bundles

Yun

Park

Juliá

et al. 2008

J. Nucl. Sci. Technol.

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In this paper, the interfacial flow structure of subcooled water boiling flow in a subchannel of 3 Â 3 rod bundles is presented. The 9 rods are positioned in a quadrangular assembly with a rod diameter of 8.2 mm and a pitch distance of 16.6 mm. Local void fraction, interfacial area concentration, interfacial velocity, Sauter mean diameter, and liquid velocity have been measured using a conductivity probe and a Pitot tube in 20 locations inside one of the subchannels. A total of 53 flow conditions have been considered in the experimental dataset at atmospheric pressure conditions with a mass flow rate, heat flux, inlet temperature, and subcooled temperature ranges of 250-522 kg/m 2 s, 25-185 kW/m 2 , 96.6-104.9 C, and 2-11 K, respectively. The dataset has been used to analyze the effect of the heat flux and mass flow rate on the local flow parameters. In addition, the area-averaged data integrated over the whole subchannel have been used to validate some of the distribution parameter and drift velocity constitutive equations and interfacial area concentration correlations most used in the literature.

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Section: Survey Of Existing Work 1 One-dimensional Drift-flux Mmentioning

confidence: 99%

Section: Survey Of Existing Work 1 One-dimensional Drift-flux Mmentioning

confidence: 99%

Flow Structure of Subcooled Boiling Water Flow in a Subchannel of 3×3 Rod Bundles

Yun

Park

Juliá

et al. 2008

J. Nucl. Sci. Technol.

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“…For vertical pipe flows and conditions studied in this work, RELAP5/MOD3 use the Chexal-Lellouche correlation [20,21]:…”

Section: Drift-velocity Approach (Dva)mentioning

confidence: 99%

On the One-Dimensional Modeling of Vertical Upward Bubbly Flow

Peña-Monferrer

Gómez-Zarzuela

Chiva

et al. 2018

Science and Technology of Nuclear Installations

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The one-dimensional two-fluid model approach has been traditionally used in thermal-hydraulics codes for the analysis of transients and accidents in water-cooled nuclear power plants. This paper investigates the performance of RELAP5/MOD3 predicting vertical upward bubbly flow at low velocity conditions. For bubbly flow and vertical pipes, this code applies the driftvelocity approach, showing important discrepancies with the experiments compared. Then, we use a classical formulation of the drag coefficient approach to evaluate the performance of both approaches. This is based on the critical Weber criteria and includes several assumptions for the calculation of the interfacial area and bubble size that are evaluated in this work. A more accurate drag coefficient approach is proposed and implemented in RELAP5/MOD3. Instead of using the Weber criteria, the bubble size distribution is directly considered. This allows the calculation of the interfacial area directly from the definition of Sauter mean diameter of a distribution. The results show that only the proposed approach was able to predict all the flow characteristics, in particular the bubble size and interfacial area concentration. Finally, the computational results are analyzed and validated with cross-section area average measurements of void fraction, dispersed phase velocity, bubble size, and interfacial area concentration.

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“…[10]. From the drift-flux relationship, the relative velocity between phases, which appears in field equations to account for the mass, momentum, and energy diffusion, is expressed as…”

Section: Constitutive Relationsmentioning

confidence: 99%

“…TAPINS uses the Chexal-Lellouche model [10] as a kinematic constitutive equation to predict the drift velocity. It has been reported that this model is applicable for a wide range of flow conditions and is accurate for most applications.…”

Section: Constitutive Relationsmentioning

confidence: 99%

Assessment of TAPINS code for application to thermal-hydraulic analysis of an integral reactor, REX-10

Lee

Park

2013

Journal of Nuclear Science and Technology

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The thermal-hydraulic analysis program for integral reactor system (TAPINS) is a thermal-hydraulic system code developed by Seoul National University for transient analysis of an integral reactor, REX-10. Specialized for a fully passive integral pressurized water reactor, TAPINS adopts a one-dimensional fourequation drift-flux model for two-phase flows. It also consists of component models for the core, the helicalcoil steam generator, and the steam-gas pressurizer. This paper presents the developmental assessment of TAPINS to validate its applicability to the thermal-hydraulic analysis of REX-10. Assessment problems are determined by taking into account thermal-hydraulic phenomena expected during design basis accidents of REX-10, including the loss-of-feedwater accident and the small-break loss-of-coolant accident. To confirm the predictive capability of TAPINS for these phenomena, the TAPINS model is validated against four sets of separate effects problems, including the pressurizer insurge test, the subcooled boiling experiment, the critical flow test, and the Edwards pipe problem. In addition, the calculation results of TAPINS are compared with the experimental data obtained from a series of integral effects tests using a scaled apparatus of REX-10. From the validation results, it is demonstrated that TAPINS can provide the reasonable prediction on the thermal-hydraulic responses of REX-10 during the transient and accident conditions.

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A void fraction correlation for generalized applications

Cited by 130 publications

References 9 publications

Flow Structure of Subcooled Boiling Water Flow in a Subchannel of 3×3 Rod Bundles

Flow Structure of Subcooled Boiling Water Flow in a Subchannel of 3×3 Rod Bundles

On the One-Dimensional Modeling of Vertical Upward Bubbly Flow

Assessment of TAPINS code for application to thermal-hydraulic analysis of an integral reactor, REX-10

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