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

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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 ... continued below

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12 pages

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KOHURT, P. (BNL), KALINICHENKO, S.D.; KROSHILIN, A.E.; KROSHILIN, V.E. & SMIRNOV, A.V. June 4, 2006.

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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.

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12 pages

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  • 2006 INTERNATIONAL CONGRESS ON ADVANCES IN NUCLEAR POWER PLANTS (ICAPP'06); RENO, NV; 20060604 through 20060608

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  • Report No.: BNL--75556-2006-CP
  • Grant Number: DE-AC02-98CH10886
  • Office of Scientific & Technical Information Report Number: 882227
  • Archival Resource Key: ark:/67531/metadc892745

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  • June 4, 2006

Added to The UNT Digital Library

  • Sept. 23, 2016, 2:42 p.m.

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  • Nov. 17, 2016, 8:07 p.m.

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KOHURT, P. (BNL), KALINICHENKO, S.D.; KROSHILIN, A.E.; KROSHILIN, V.E. & SMIRNOV, A.V. TWO-PHASE FLOW STUDIES IN NUCLEAR POWER PLANT PRIMARY CIRCUITS USING THE THREE-DIMENSIONAL THERMAL-HYDRAULIC CODE BAGIRA., article, June 4, 2006; [Upton, New York]. (digital.library.unt.edu/ark:/67531/metadc892745/: accessed October 18, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.