Flooding characteristics of gas-liquid two-phase flow in a horizontal U bend pipe

Technical Report · Fri Sep 01 04:00:00 EDT 1995

DOI:https://doi.org/10.2172/107025· OSTI ID:107025

Sakaguchi, T.; Hosokawa, S.; Fujii, Y. ^[1]

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1. Kobe Univ. (Japan); and others

For next-generation nuclear reactors, hybrid safety systems which consist of active and passive safety systems have been planned. Steam generators with horizontal U bend pipelines will be used as one of the passive safety systems. It is required to clarify flow characteristics, especially the onset of flooding, in the horizontal U bend pipelines in order to examine their safety. Flooding in vertical pipes has been studied extensively. However, there is little study on flooding in the horizontal U bend pipelines. It is supposed that the onset of flooding in the horizontal U bend pipelines is different from that in vertical pipes. On the other hand, liquid is generated due to condensation of steam in pipes of the horizontal steam generators at the loss of coolant accident because the steam generators will be used as a condenser of a cooling system of steam from the reactor. It is necessary to simulate this situation by the supply of water at the middle of horizontal pipe. In the present paper, experiments were carried out using a horizontal U bend pipeline with a liquid supply section in the midway of pipeline. The onset of flooding in the horizontal U bend pipeline was measured. Effects of the length of horizontal pipe and the radius of U bend on the onset of flooding were discussed.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; American Nuclear Society, La Grange Park, IL (United States); American Inst. of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers, New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); Japan Society of Multiphase Flow, Kyoto (Japan)
• OSTI ID: 107025
• Report Number(s): NUREG/CP--0142-Vol.1; CONF-950904--Vol.1; ON: TI95017077
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
DIAGRAMS
EXPERIMENTAL DATA
HYDRAULICS
LOSS OF COOLANT
PROCEEDINGS
PWR TYPE REACTORS
REACTOR SAFETY
SAFETY ANALYSIS
STEAM GENERATORS
TUBES
TWO-PHASE FLOW