TWO-PHASE FLOW STUDIES IN NUCLEAR POWER PLANT PRIMARY CIRCUITS USING THE THREE-DIMENSIONAL THERMAL-HYDRAULIC CODE BAGIRA.

Conference · Sun Jun 04 04:00:00 EDT 2006

OSTI ID:882227

KROSHILIN, A E; KROSHILIN, V E; SMIRNOV, A V

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.

• Research Organization: BROOKHAVEN NATIONAL LABORATORY
• Sponsoring Organization: DOE/IPP
• DOE Contract Number: AC02-98CH10886
• OSTI ID: 882227
• Report Number(s): BNL--75556-2006-CP; NN-410-1010
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ABLATION
ACCIDENTS
BENCHMARKS
NUCLEAR POWER PLANTS
NUMERICAL ANALYSIS
REACTORS
SIMULATION
STEAM GENERATORS
TRANSIENTS
TWO-PHASE FLOW