Axial
and radial vapor void fraction profiles in forced convection flow boiling

Ashutosh Yadav and Shantanu Roy*

Department of Chemical Engineering,
Indian Institute of Technology - Delhi, New Delhi 110016, India.

 (ashutoshy701@gmail.com, *roys@chemical.iitd.ac.in)

Keywords:  Boiling
Flow, Forced Convection. Void Fraction, Gamma ray densitometry

Boiling
flow is of great importance to nuclear reactor systems and have been the
subject of numerous theoretical and experimental investigations [1]. The
particular relevance to the nuclear industry is for applications in thermal
hydraulics in boiling water reactors. In these applications, liquid phase water
is brought in typically under sub-cooled conditions and made to flow around
nuclear fuel pins (vertical rods) held concentrically within a large vertical
cylindrical vessel. The fuel pins serve as the principal source of heat,
usually with very high energy fluxes. The liquid water undergoes phase change
in the vicinity of the heated fuel rods, even when other parts of the column may
continue to be under sub-cooled or saturated liquid conditions. This leads to
differential distribution of vapor and liquid phases, with the vapor tending to
segregate both radially and axially. Indeed, in turn this segregated vapor
drives the liquid circulation, and in natural circulation boiling water
reactors, is the sole cause of the flow of the two-phase mixture to occur.

One
of the main challenges in operating this kind of a reactor system are in the
complexities of two-phase flow around the rods driven by a vertically
distributed heat flux in the rods [2]. This is because the void fraction (vapor
fraction) distribution significantly affects the reactor power and is one of
the important parameters that determine the heat transfer capability and the
possible occurrence of critical heat flux [3]. Knowledge of the time-averaged
void fraction distribution as well as the velocity profiles of the liquid phase
are of great relevance in design of these systems, for providing validation
data for thermal-hydraulic CFD codes, as well as for design of nuclear safety
systems.

In
this contribution, measurements for radial void fraction distribution will be reported
for a vertical upward flowing boiling flows in different configurations of
heater and flow channels. In one realization, we present an annulus channel consisting
of an inner electrical-heater rod with a diameter of 40 mm and an outer round pipe
with an inner diameter of 73.6 mm. A schematic of the experimental loop is
shown in Figure 1(b), and a photograph of the setup is shown in Figure 1(a). In
another configuration (Figure 2), we have a similar vertical tube but fitted
with different configurations of heater rods (varying both in number and in
heat flux). Using a variable rheostat, the heat flux from the electrical rods
in either case is varied, as is the incoming (sub-cooled) water temperature, to
?dial in? into different flow boiling regimes in the vertical tubes. Note that
in this work, all cases presented are for forced convection boiling, i.e., the
liquid water is independently pumped into the system.

For
making the void fraction measurements, gamma ray densitometry is used. This is a
well suited technique for measurement of void fraction non-invasively and has
been extensively used for measuring void fractions in many non-boiling
gas-liquid flows [4], both in terms of time-averaged axial and radial profiles.
In this contribution, the technique is adapted for measurements in boiling
flows in the aforementioned setup. Figure 3 shows a typical result of radial
variation of time averaged chordal void fraction at different axial locations. Comparative
results for the different geometries, flow conditions, and heater fluxes will
be presented in the final contribution. Analysis of the results will be
presented.

Aside
from the direct measurements done exhaustively under a wide variety of
conditions, the contribution also tries to assess the evolution of flow regimes
along the vertical length of the vessel, and attempts to map that qualitatively
into what might be happening in large-scale boiling water reactors.

  
           

Figure
1. a) Photograph of the Experimental setup b) Schematic diagram of experimental
set up.  

Figure
2. a) 9 Heater rods arrangement b) 10 Heater rods arrangement.

Figure
3. Radial variation of time averaged chordal void fraction for q=28.43 kW/m²,
T_in = 45⁰C and v_in=4.17 mm/s.

References

1.                 
Todreas, N. E., Kazimi, M. S., 2012.
Nuclear Systems. Taylor and Francis Group, New York.

2.                 
Collier,
J.G., Thome, J.R., 2001. Convective Boiling and Condensation. Oxford Science
Publications, London.

3.                 
Situ, R., Hibiki, T., Sun, H., Mi, Y.,
Ishii M., 2004. Flow structure of subcooled boiling flow in an internally
heated annulus. Int. J. Heat Mass Transfer 47, 5351-5364.

4.                 
Chan, A.M.C., Banerjee, S., 1981. Design
aspects of gamma densitometers for void fraction measurements in small scale
two-phase flow. Nucl. Instrum. Methods Phys. Res. 190, 135-148.