Drift flux modeling of the VVER-1000 horizontal nuclear steam generator

Abstract

Nuclear steam generator is a critical component of the pressurized-water nuclear power plant that plays an important role in security and efficiency of the nuclear power plant. Therefore, using assured thermo-hydraulic model to simulate the nuclear steam generator has particular importance. In this paper, the numerical solution of void fraction in horizontal nuclear steam generator in the steady state analysis is presented using Drift-Flux Model. For a two phase mixture of a gas or vapor and liquid flowing together in a channel, different internal flow geometries or structure can occur depending on the size or orientation of the flow channel, the magnitudes of the gas and liquid flow parameters, the relative magnitudes of this flow parameter, and on the fluid properties of the two phases. The Drift-Flux Model (DFM) is able to predict the void fraction in different geometries. The drift velocity in various two phase flow regimes for small and big diameter pipes is explained. VVER-1000 nuclear steam generator is simulated by DFM using the FLUENT 6.3.26 code. It is explained that void fraction in horizontal steam generator is strongly affected by using a perforated sheet in the top of the horizontal hot tubes. Validity and superiority of the DFM compared to the other two-phase models is proved. Simulation results are compared with RELAP5 and BAGIRA codes results. The calculated void fraction is in good agreement with measured data. The accuracy of the prediction shows that it is possible to use the DFM for thermal-hydraulic analysis in advanced models in nuclear power plant and other industries. This model can be used for assessment of experimental data and licensing processes.