Two-Phase Flow Studies in Nuclear Power Plant Primary Circuits Using the Three-Dimensional Thermal-Hydraulic Code BAGIRA
- All-Russian Research Institute for Nuclear Power Plant Operations (VNIIAES) 25 Ferganskaya St., 109507 Moscow (Russian Federation)
- 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
- Resource Relation:
- Conference: 2006 International congress on advances in nuclear power plants - ICAPP'06, Reno - Nevada (United States), 4-8 Jun 2006; Other Information: Country of input: France; 7 refs; Related Information: In: Proceedings of the 2006 international congress on advances in nuclear power plants - ICAPP'06, 2734 pages.
- Country of Publication:
- United States
- Language:
- English
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