Void fraction and flow regime in adiabatic upward two-phase flow in large diameter vertical pipes

Abstract

In pipes with very large diameters, slug bubbles cannot exist. For this reason, the characteristics of two-phase flow in large pipes are much different than those in small pipes. Knowledge of these characteristics is essential for the prediction of the flow in new nuclear reactor designs which include a large chimney to promote natural circulation. Two of the key parameters in the prediction of the flow are the void fraction and flow regime. Void fraction measurements were made in a vertical tube with diameter of 0.15 m and length of 4.4 m. Superficial gas and liquid velocities ranged from 0.1 to 5.1 m/s and from 0.01 to 2.0 m/s, respectively. The measured void fractions ranged from 0.02 to 0.83. Electrical impedance void meters at four axial locations were used to measure the void fraction. This data was verified through comparison with previous data sets and models. The temporal variation in the void fraction signal was used to characterize the flow regime through use of the Cumulative Probability Density Function (CPDF). The CPDF of the signal was used with a Kohonen Self-Organized Map (SOM) to classify the flow regimes at each measurement port. The three flow regimes used were termed bubbly, cap-bubbly, and churn flow. The resulting flow regime maps matched well with the maps developed previously through other methods. Further, the flow regime maps matched well with the criteria which were proposed based on Mishima and Ishii's (1984) criteria.

Introduction

The safety characteristics of the next generation of nuclear reactors will be evaluated using new advanced best-estimate thermal hydraulic system codes. Using these codes will allow for streamlining the licensing process by allowing the safety of a reactor design to be evaluated without the construction of expensive full-scale models. This will reduce the expense and time associated with the construction of new power plants, increasing the economic viability of nuclear power.

Many of these reactor designs will be natural circulation Boiling Water Reactors (BWRs). These designs include a large chimney section above the reactor. A large pipe is defined as a pipe in which slug bubbles can no longer exist. When the hydraulic diameter of a two-phase flow channel reaches a certain size, slug bubbles bridging the entire diameter can no longer be sustained due to Taylor instability. For this reason Kataoka and Ishii (1987) defined the critical diameter at which this occurs in terms of the Taylor wavelength as

$$D_H^{*}=\frac{D_H}{\sqrt{\sigma /g\Delta \rho }}\ge 40$$

When the pipe diameter is larger than this value, the result is dramatic changes in the structure and dynamics of the two-phase flow. For this reason, experimental data obtained in small diameter pipes in which slug flow exists cannot be scaled for use in understanding flow in large diameter pipes (Hibiki and Ishii, 2003). Therefore experimental data and correlations must be developed for analysis of systems which include flow channels of large diameter.

Natural circulation depends on the density difference between the hot water or two-phase mixture in the chimney and the cold water in the downcomer, thus the chimney section provides the gravity head necessary to drive natural circulation through the reactor core. It is therefore necessary that accurate models exist for the prediction of two-phase flow in such systems, as this will affect the prediction of the two-phase flow rate in new Boiling Water Reactor designs. Additionally, the prediction of void fraction in large pipes is significant for the analysis of accident scenarios, especially loss of coolant accidents, because the void fraction plays a large role in the determination of the liquid inventory in the reactor pressure vessel. This is one of the major factors in the ability of a reactor to cool itself in the event of an accident.

In addition to the development of correlations for the prediction of void fraction, the ability to accurately identify the flow regimes and characterize the physical mechanisms governing each flow regime is necessary for the development of accurate two-phase flow models. Extensive work has been done regarding flow regime transition in small pipes, but the flow regimes in large pipes are significantly different and require a separate analysis (Mishima and Ishii, 1984).

Two-phase flows always involve some relative motion between the phases. For this reason, problems in two-phase flows should be formulated in terms of two velocity fields: one for each phase. In general, the two-fluid model or drift-flux model can be adapted for a solution, depending on how strongly the two phases are coupled. While the drift-flux model is approximate compared to the two-fluid model, it has the advantages of being simple and allowing for rapid calculation as well as applicability to a wide range of two-phase flow conditions. These advantages make the drift-flux model of considerable practical interest, and models based on drift-flux formulations are often used as constitutive models in advance thermal–hydraulic analysis codes.

A review of the current literature has revealed very little past work in the objective identification of the two-phase flow regimes in large pipes, as well as insufficient work in developing void fraction models and data at intermediate void fractions in such systems. In view of this, research has been performed to provide additional void fraction data in a large diameter test section. This research includes data on the axial development of void fraction and on objective flow regime identification.