Numerical investigation on thermal-hydraulic steady-state characteristics of steam generator in different operating conditions

Three-dimensional unit pipe physical model of fluid-solid coupling heat transfer on primary side fluid, secondary side fluid and tubes in the steam generator was established based on the similarity and modeling principle. Thermal-hydraulic steady-state characteristics of the steam generator of Daya Bay Nuclear Power Plant were investigated by numerical simulation under different operating conditions. Thermal phase change model was utilized to describe vapor-liquid two-phase flow and heat transfer, and heat transfer between primary side coolant and secondary side fluid through the tubes was calculated by fluid-solid coupled model. Numerical results show that tube inner wall temperature distribution profile is almost identical with that of primary side fluid at a full power, and so is secondary side fluid with the outer wall. Cross-section average void fraction increases along the height of tubes and the outlet mass fraction of vapor is consistent with the actual operating value of Daya Bay Nuclear Power Plant. With the power load lowering down, the primary outlet temperature maintains almost unchanged, the outlet temperature of the secondary side increases, but both the outlet mass fraction and heat transfer coefficient of vapor decrease. The average heat transfer coefficient is basically consistent with the result of Rohsenow empirical correlation.