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A criterion to classify void fraction profiles in bubbly flows based on averaged flow quantities for use in subchannel codes

Abstract : Hydrodynamics of multiphase flows can have a large influence on the overall design and operation of nuclear reactors. Interfacial forces drive bubbles towards or away from the wall, resulting in wall-peaked or core-peaked void fraction profiles. This change has a dramatic effect on the averaged flow quantities such as wall shear stress, velocity profiles and turbulence levels. Subchannel codes lack information of the underlying flow profiles, and rely on closure models to impose the correct averaged flow quantities. In this work, we propose a criterion to classify bubbly flows as wall-peaked or core-peaked, based on the turbulence level, and the ratio of surface tension to buoyancy forces. The criterion has been developed leveraging a large set of adiabatic air-water bubbly flow experiments, and can be used in subchannel codes to identify the underlying void fraction profile and apply the appropriate closure model. The criterion shows low sensitivity when applied to different operating fluids, and thus can be extended into a generalized formulation that can be applied to all adiabatic bubbly flows. An experiment has also been proposed to confirm the classification criterion at reactor conditions, using water with additives at atmospheric pressure to mitigate the high cost of high-pressure experiments. The new criterion proposed in this work, along with the improved understanding from new experiments, will allow us to develop more physically consistent closure models for subchannel codes.
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Contributor : Bibliothèque Cadarache <>
Submitted on : Wednesday, October 21, 2020 - 3:36:59 PM
Last modification on : Thursday, October 22, 2020 - 3:08:06 AM


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  • HAL Id : cea-02974095, version 1




R. Kommajosyula, G. Bois, Alan Burlot, M.-G. Rodio, B. Cariteau, et al.. A criterion to classify void fraction profiles in bubbly flows based on averaged flow quantities for use in subchannel codes. NURETH18 2019 - 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, ANS, Aug 2019, Portland, United States. ⟨cea-02974095⟩



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