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

Abstract

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.

Interfacial area concentration in gas–liquid bubbly to churn-turbulent flow regime

“…The experimental existing databases particular to IAC measurements have been provided extensively by Ishii (2001, 2002), Hibiki et al (2006) and Ozar et al (2008). Among the work presented in these publications, Akita and Yoshida (1974), Liu (1989), Serizawa et al (1991), Roy et al (1994) and Yun et al (2008) provide a valuable experimental database for different flow geometries and wide conditions. Very few reliable data exists in the literature that covers a wide range of flow conditions and flow regimes.…”

Vertical upward two-phase flow CFD using interfacial area transport equation

“…Besides the databases listed in Tables 5 and 6, several unique and useful data are available in literature. They are boiling flow data in a sub-channel of 3 Â 3 rod bundle (Yun et al, 2008) and interfacial area transport data of vertical upward steam-water twophase flow in an annulus at elevated pressures (Ozar et al, 2013) and interfacial area transport data of vertical upward airewater two-phase flow in an annulus channel (Jeong et al, 2008). More detailed review is found in Lin and Hibiki (2014).…”

Velocity measurements in convective boiling flow using radioactive particle tracking technique

“…The effective heat removal depends on the complex hydrodynamics of the two‐phase flow around the nuclear rods bundle. The hydrodynamics observed in the nuclear reactor core is complicated owing to the presence of nuclear rods (which is a challenging geometry), and also due to the complex phenomena of bubble nucleation, growth, detachment, and simultaneous motion of bubbles and the liquid phase …”