Abstract
The properties of LiF-LiCl-LiBr salt make it attractive as a solvent for extracting tritium from a fusion reactor lithium blanket. Consequently, the corrosion of type 316 stainless steel by flowing (about 15 mm/s) LiF-LiCl-LiBr at a maximum temperature of 535°C was studied to determine whether compatibility with the structural material would be limiting in such a system. The corrosion rate was found to be low (<2 μm/year) except immediately after the addition of a small amount of lithium metal to the salt. The lithium addition increased the corrosion rate to 13.5 μm/year at 535°C (approximately that of type 316 stainless steel exposed to lithium flowing at a similar velocity). At the proposed operating temperature (≤ ~535°C), however, it appears that type 316 stainless steel has acceptable compatibility with the tritium-processing salt LiF-LiCl-LiBr for use with a lithium blanket.
| Original language | English |
|---|---|
| Pages (from-to) | 675-680 |
| Number of pages | 6 |
| Journal | Journal of Nuclear Materials |
| Volume | 103 |
| Issue number | C |
| DOIs | |
| State | Published - 1981 |
Funding
l&search sponsored by the Office of Fusion Energy, U.S. Department of Energy under contract W-7405-eng-26 with the Union Carbide Corporation.