A thermal-hydraulic code for transient analysis in a channel with a rod bundle

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

DOI:https://doi.org/10.2172/107769· OSTI ID:107769

Khodjaev, I. D. ^[1]

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1. Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)

The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.

• 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); European Nuclear Society (ENS), Bern (Switzerland); Atomic Energy Society of Japan, Tokyo (Japan); Japan Society of Multiphase Flow, Kyoto (Japan)
• OSTI ID: 107769
• Report Number(s): NUREG/CP--0142-Vol.2; CONF-950904--Vol.2; ON: TI95017078
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
COMPUTER CODES
EXPERIMENTAL DATA
FUEL ELEMENT CLUSTERS
HYDRAULICS
LENINGRAD-1 REACTOR
MATHEMATICAL MODELS
REACTOR SAFETY
SAFETY ANALYSIS
THEORETICAL DATA
TWO-PHASE FLOW