Two-Phase Flow Studies in Nuclear Power Plant Primary Circuits Using the Three-Dimensional Thermal-Hydraulic Code BAGIRA

Conference · Sat Jul 01 04:00:00 EDT 2006

OSTI ID:21016403

Kalinichenko, S D; Kroshilin, A E; Kroshilin, V E; Smirnov, A V ^[1]; Kohut, P ^[2]

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1. All-Russian Research Institute for Nuclear Power Plant Operations (VNIIAES) 25 Ferganskaya St., 109507 Moscow (Russian Federation)
2. Brookhaven National Laboratory, Bldg. 475 Upton, NY 11973, (United States)

in this paper we present recent results of the application of the thermal-hydraulic code BAGIRA to the analysis of complex two-phase flows in nuclear power plants primary loops. In particular, we performed benchmark numerical simulation of an integral LOCA experiment performed on a test facility modeling the primary circuit of VVER-1000. In addition, we have also analyzed the flow patterns in the VVER-1000 steam generator vessel for stationary and transient operation regimes. For both of these experiments we have compared the numerical results with measured data. Finally, we demonstrate the capabilities of BAGIRA by modeling a hypothetical severe accident for a VVER-1000 type nuclear reactor. The numerical analysis, which modeled all stages of the hypothetical severe accident up to the complete ablation of the reactor cavity bottom, shows the importance of multi-dimensional flow effects. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 21016403
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ABLATION
BENCHMARKS
CAVITIES
LOSS OF COOLANT
NUCLEAR POWER PLANTS
NUMERICAL ANALYSIS
REACTOR OPERATION
SIMULATION
STEAM GENERATORS
TEST FACILITIES
THERMAL HYDRAULICS
THREE-DIMENSIONAL CALCULATIONS
TRANSIENTS
TWO-PHASE FLOW
WWER TYPE REACTORS