Degradation susceptibility of steam generator tubing materials under ACR steam generator secondary side crevice conditions

Alloy 800 material has been selected as the preferred steam generator (SG) tubing material for the Advanced CANDU Reactor (ACR). Alloy 800 tubing has a good service record in SGs of both German pressurized water reactors and CANDU 6 reactors. On the other hand, thermally treated Alloy 690 (Alloy 690 TT) is also a favored material for replacing Alloy 600 SG tubing in the world nuclear industry. Both alloys have excellent stress corrosion cracking (SCC) resistance to primary water environments and to most off-specification (faulted) secondary side environments. No in-service SCC has been observed for Alloy 690 since its first application in 1989. However, both Alloy 800 and Alloy 690 are susceptible to SCC and localized corrosion under some extreme off-specification chemistry conditions in laboratory tests. In order to confirm that Alloy 800 and Alloy 690 TT SG materials are qualified SG tubing materials for the applications under ACR SG condition, laboratory tests were performed to evaluate the corrosion susceptibility of both alloys under plausible ACR SG crevice chemistry conditions. Constant Extension Rate Tests and accelerated crevice corrosion tests were performed under a range of ACR SG conditions. It was found that both Alloy 800 and 690 TT have good SCC and crevice corrosion resistance under plausible ACR crevice chemistry conditions. The experimental results support the ACR design of using either Alloy 800 or Alloy 690 TT tubing materials for the ACR SGs. (author)