Development of thermal hydraulic computer code for steam-water flow in steam generator of fast breeder reactor

Yoshikawa, Ryuji  ; Hamada, Hirotsugu  ; Ohshima, Hiroyuki; Yanagisawa, Hideki*

In Japan Atomic Energy Agency, investigations on heat transfer characteristics of steam generator are being performed for the development of commercialized sodium-cooled fast breeder reactor. In this study, the computer code for flow instability analysis in single heat transfer tube was developed with drift-flux model which included the effects of subcooled boiling and two phase slip. The special algorithm to calculate inlet flow rate with inlet pressure, outlet pressure and heat flux as boundary conditions for the density-wave instability analysis has been established. The subcooled model was verified by calculating the void fraction distribution of steady heat transfer flow. The capability of drift flux model for simulating density-wave instability in single tube was confirmed.