A safety evaluation of fluid induced vibration for APR1400 steam generator tube

The first one in the contamination by radioactivity due to the breakage of the pressure boundary between the primary system and the secondary system. The other is the possibility of the so called Loss Of Coolant accident (LOCA) whereby the primary system cooling water will be decreased because the primary system cooling water flowed into the secondary system due to the pressure difference between them when the heat transfer tubes of the steam generator was damaged. Besides the safety problem, the excessive wear of S/G tubes causes a sudden stoppage of operation in the nuclear power plant and requires frequent inspections of the fine tube plugging. Also, it results in diverse economic losses such as labor costs due to the tube plugging and its inspection, irradiation to workers, and long term outage, etc. In the current steam generators for KSHP, fretting wear due to the Fluid induced Vibration has occurred commonly and this takes mainly placed at the upper part of the U Bend above the central cavity. Therefore, for the UCN 5,6 and OPR1000's steam generators, in order to prevent Fluid Induced Vibration (FIV) by decreasing the fluid density and the velocity in this area., the Egg crate Flow Distribution Plate (EFDP) was installed in the central cavity area. Also, in order to prevent low natural frequency, tube support of the U bend region in APR1400 steam generator was modified. In this study, we have performed a Safety Evaluation of Fluid Induced Vibration for APR1400 S/G related to those design changes