Influence of two-phase flow characteristic on critical heat flux in low pressure

Book · Thu Aug 01 04:00:00 EDT 1996

OSTI ID:271902

Inoue, Akira; Lee, Sangryoul ^[1]

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1. Tokyo Inst. of Tech., Yokohama (Japan). Interdisciplinary Graduate School of Science and Engineering

Estimation of the critical heat flux (CHF) in a boiling two-phase flow is one of the important subjects for the safety of water-cooled reactors and other energy systems. In the case of a boiling two-phase flow at low pressure, flow pattern and void fraction are easy to change by the power input and the flow becomes more complex due to low density of gas phase. The CHF is affected by the flow pattern. In this study, the CHFs were measured over wide quality range from the subcooled boiling to the annular-mist flow. By using Pyrex glass tube as a test channel, the two-phase flow situation was observed. Graphite rod or stainless steel tube was used as a heater rod and installed at the center of the glass tube. Two-phase flow was formed by steam injection to circulating water at an upstream region of the test section. The flow pattern was kept nearly constant over the length of test section due to the low input power density into the fluid. Then, the characteristics of CHF could be investigated at each flow patterns of bubbly, slug, annular and annular-mist flow. In the subcooled boiling region of bubbly flow, the CHF decreased with increase of quality and was less sensitive to flow rate. In the slug flow region, the CHF showed a minimum value. With more increase of quality in the annular flow, the CHF increased and reached a peak value at a certain quality depending on a flow rate. The peak of CHF occurred almost at a constant vapor mass velocity. In the annular-mist flow region, the CHF decreased with increase of quality. In the region, the effect of heated length on the CHF was systematically measured and validity of an analytical model considering dryout of liquid film based on formation of a dry patch was investigated.

• OSTI ID: 271902
• Report Number(s): CONF-960306--; ISBN 0-7918-1226-X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
ANNULAR SPACE
CRITICAL HEAT FLUX
ENERGY SYSTEMS
EXPERIMENTAL DATA
REACTOR SAFETY
SAFETY ANALYSIS
SUBCOOLED BOILING
TWO-PHASE FLOW
VARIATIONS
WATER COOLED REACTORS