TWO-PHASE FLOW STUDIES IN NUCLEAR POWER PLANT PRIMARY CIRCUITS USING THE THREE-DIMENSIONAL THERMAL-HYDRAULIC CODE BAGIRA.
Description:
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 …
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Date:
June 4, 2006
Creator:
KOHURT, P. (BNL), KALINICHENKO, S.D.; KROSHILIN, A.E.; KROSHILIN, V.E. & SMIRNOV, A.V.
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