Void Fraction Measurement and Prediction of Two-Phase Boiling Flows in a Tubular Test Section

Author:

Liu Qingqing1,Diaz Julio1,Petrov Victor1,Burak Adam1,Manera Annalisa1,Kelly Joseph2,Sun Xiaodong1

Affiliation:

1. Department of Nuclear Engineering & Radiological Sciences, University of Michigan , Ann Arbor, MI 48109-2104

2. Office of Nuclear Regulatory Research, The U.S. Nuclear Regulatory Commission , Washington, DC 20555-0001

Abstract

Abstract Void fraction is one of the most important parameters that affect two-phase flow heat transfer and pressure drop. In this paper, a commercial gamma densitometer and a high-speed X-ray radiography system developed at the University of Michigan (UM) are used to measure the void fraction in two-phase boiling flows, with water as the working fluid, in a tubular test section. The test section is made of Incoloy 800H/HT with a total length of 1.589 m, an inner diameter of 12.95 mm, and a wall thickness of 3.05 mm. These two instrumentation systems are installed on a traversing platform that travels along the vertical test section to perform measurements at multiple elevations. Subcooled flow boiling and natural convection boiling experiments are performed to measure the void fraction in the test section. Flow visualization images are obtained for bubbly and slug flows from the X-ray radiography system. The wall temperature of the test section is measured at 17 elevations by thermocouples. In addition to the experiments, a multiphase computational fluid dynamics (MCFD) model is developed using ansysfluent to simulate the subcooled flow boiling. The measured wall temperature and void fraction from the experiments are compared with the MCFD simulation results. The root-mean-square (RMS) relative deviations are 3.6% and 16.1% for the wall temperature and void fraction, respectively, between the experimental data and MCFD simulations.

Funder

U.S. Nuclear Regulatory Commission

Publisher

ASME International

Subject

Nuclear Energy and Engineering,Radiation

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