Development of tube rupture evaluation code for FBR steam generator (II); Modification of heat transfer model in sodium side

Hamada, Hirotsugu  ; Kurihara, Akikazu 

The thermal effect of sodium-water reaction jet on neighboring heat transfer tubes was examined to rationally evaluate the structural integrity of the tube for overheating rupture under a water leak in an FBR steam generator. Then, the development of new heat transfer model and the application analysis were carried out. Main results in this paper are as follows. (1)The evaluation method of heat flux and heat transfer coefficient (HTC) on the tube exposed to reaction jet was developed. By using the method, it was confirmed that the heat flux could be realistically evaluated in compatison with the previous method. (2)The HTC between reaction jet and the tube was theoretically examined in the two-phase flow model, and new heat transfer model considering the effect of fluid temperature and cover gas pressure was developed. By applying the model, a tentative experimental correlation was conservatively obtained by using SWAT-1R test data. (3)The new model was incorporated to the Tube Rupture Evaluation Code (TRUE), and the conservatism of the model was confirmed by using sodium-water reaction data such as the SWAT-3 tests. (4)In the application analysis of the PFR large leak event, there was no significant difference of calculation results between the new model and previous one; the importance of depressurization in the tube was confirmed. (5)In the application analysis of the Monju evaporator, it was confirmed that the calculation result in the previous model would be more conservative than that in the new one and that the maximum cumulative damage of 25% could be reduced in the new model.