Dispersed two-phase flows are part of many fluid systems such as the cooling of nuclear reactors. Depending on the heat flux in the reactor core, the flow rate, the subcooling or the pressure, different flows may occur: single phase, bubbly or annular flows (with a liquid film on the wall and a vapour core).
During a loss of primary coolant accident, the reactor core, containing the fuel rods, increases in temperature until the boiling crisis when the heat flux is high enough. The different regimes of two-phase flows that occur in this type of accident are illustrated in figure 1. A vapour film appears rapidly and thermally insulates the rods, while some liquid remains in the centre of the flow. The rods are dried up, thus their surface are cooled down by the single vapour, and the heat exchange at the wall is reduced [1], which corresponds to the « inverted annular film boiling » flow. When the liquid gradually vaporises, the vapour film thickens and the induced turbulence tends to form waves at the vapour-liquid interface, and to destabilise the interface until the formation of liquid slugs (inverted slug film boiling). Then, the evaporation and fragmentation of these slugs lead to the formation of a dispersed flow with droplets (dispersed film boiling).

The transitions of flow regimes in this configuration are not well-identified [1], [2] although their understanding is significant to study the cooling of a nuclear reactor core. One of the main obstacles in experimental studies is that the walls need to be strongly heated up in order to form and maintain a vapour film, which leads to opaque test sections. Thus, a direct visualisation is particularly complex to obtain, as much as measuring local parameters such as temperature and velocity fields. The experimental results available in the literature on this topic are insufficient to develop a physical model [1], [3], [4], [5].
As a first step towards an accurate identification of the regime transitions, this thesis focuses on the single effect of the hydrodynamics, by coupling experimental and analytical approaches. In order to clarify the physics of the different phenomena, the configuration of a liquid flow inside a gas flow is proposed. Indeed, the interface deformation and the gas and liquid velocities may influence the transition from one regime to another [6], [7]: the smooth interface is therefore perturbed by waves (Kelvin-Helmholtz instabilities) and droplets could be entrained from the interface. A parametric analysis is considered by varying the gas and liquid flow rates and the thickness of the gas film, in order to observe these different phenomena and to understand the influence of each parameter on the regime transitions. An experimental facility has recently been conceived at DM2S/STMF/LE2H to study these transitions by a visualisation of the interface deformations, and may be adapted with new measurements or new methodology if necessary.
Dimensionless numbers will be identified or defined from the experimental results to describe the phenomena. Then, the regime transitions will be characterized, based on these dimensionless numbers, in order to establish a diagram of the transitions of flow regimes.
The combination of the results obtained in this thesis will enable to reinforce the physical models used in the system code CATHARE, developed at CEA for thermal-hydraulic studies about nuclear safety. This thesis presents a strong academic interest thanks to an innovative experimental facility and production of original results. Besides, it also presents an interest on the industrial level since it contributes to enhance the expertise of significant phenomena in the demonstration of nuclear reactor safety.

References:
[1] M. Ishii et G. De Jarlais, « Flow visualization study of inverted annular flow of post-dryout heat transfer region », Nuclear Engineering and Design, 1987.
[2] G. De jarlais, M. Ishii, et J. Linehan, « Hydrodynamic stability of inverted annular flow in an adiabatic simulation », Argonne National Laboratory, CONF-830702-9, 1983.
[3] T. G. Theofanous, « The boiling crisis in nuclear reactor safety and performance », International Journal of Multiphase Flow, vol. 6, no 1, p. 69-95, févr. 1980, doi: 10.1016/0301-9322(80)90040-3.
[4] N. Takenaka, T. Fujii, et others, « Flow pattern transition and heat transfer of inverted annular flow », Int. J. Multiphase Flow, 1989.
[5] M. A. El Nakla, D. C. Groeneveld, et S. C. Cheng, « Experimental study of inverted annular film boiling in a vertical tube cooled by R-134a », International Journal of Multiphase Flow, vol. 37, p. 37-75, 2011.
[6] Q. Liu, J. Kelly, et X. Sun, « Study on interfacial friction in the inverted annular film boiling regime », Nuclear Engineering and Design, vol. 375, 2021.
[7] K. K. Fung, « Subcooled and low quality film boiling of water in vertical flow at atmospheric pressure », PhD Thesis, Argonne National Laboratory, 1981.