Development of LEAP-III code for evaluation of long-time event progress under tube failure accident in steam generators

Uchibori, Akihiro ; Yanagisawa, Hideki*; Takata, Takashi  ; Kurihara, Akikazu ; Hamada, Hirotsugu  ; Ohshima, Hiroyuki

For safety assessment of a steam generator of sodium-cooled fast reactors, it is necessary to evaluate the possibility of occurring tube failure propagation and of water leak rate under sodium-water reaction accident. In the previous studies, a computer code called LEAP-II calculating a wastage-type failure propagation and the water leak rate during long-time event progress was developed. In this study, a numerical method to evaluate the possibility of occurring overheating rupture was introduced into the LEAP-II code to expand application range of this code. The completed code is called LEAP-III. The test analysis on a tube bundle configuration demonstrated that the overheating rupture model could provide conservative prediction.