Flow boiling heat transfer in a helically coiled steam generator for nuclear power applications

Abstract

Forced convection boiling of water was experimentally investigated in a 24m long full-scale helically coiled steam generator tube, prototypical of the steam generators with in-tube boiling used in small modular nuclear reactor systems. Overall, 1575 axially local and peripherally averaged heat transfer coefficient measurements were taken, covering operating pressures in the range of 2–6MPa, mass fluxes from 200 to 800kgm−2s−1 and heat fluxes from 40 to 230kWm−2. The heat transfer coefficient was found to depend on the mass flux and on the heat flux, indicating that both nucleate boiling and convection are contributing to the heat transfer process. Seven widely quoted flow boiling correlations for straight tubes fitted the present helical coil databank with a mean absolute percentage error within 15–20%, which was comparable with the experimental uncertainty of the measured heat transfer coefficient values, thus indicating that curvature effects on flow boiling are small and negligible in practical applications.