| Article ID | Journal | Published Year | Pages | File Type |
|---|---|---|---|---|
| 297460 | Nuclear Engineering and Design | 2012 | 7 Pages |
Sodium heated steam generator (SG) is a crucial component in the heat transport system of a fast breeder reactor (FBR). In case, one of its water/steam carrying tubes becomes defective, water/steam leaks into sodium, flowing in the shell side, causing sodium-water reaction, which is highly exothermic and producing corrosive NaOH. The reaction jet originating from a leaking tube may impinge on its adjacent tube, resulting in damage of the tube. Impingement wastage refers to this kind of damage, occurring to a tube of sodium heated SG, owing to a small water/steam leak from a neighboring tube. Extensive research works have been conducted all over the world to study various aspects of this phenomenon. Experimental studies were carried out in Indira Gandhi Centre for Atomic Research (IGCAR) to understand the effect of impingement wastage on Mod 9Cr 1Mo, which is the tube material of prototype fast breeder reactor (PFBR) SG. This paper brings out the data and experience gained through the experiments.
► Sodium heated steam generators are crucial components of fast breeder reactors. ► A leak in steam generator tube will cause sodium water reaction that damages the tubes. ► Experimental study was conducted to quantify the extent of damage on Mod 9Cr 1Mo tube due to a water leak.
