PRECIPITATION-HARDENING STAINLESS STEELS IN WATER-COOLED REACTORS
A study is made of the stress corrosion susceptibiity of unirradiated precipitation-hardening stainless steels. This study is made because of the failures encouatered with these materials in the Dresden and Vallecltos boiling water reactors. Service experience, static steam autoclave tests, and dynamic water and steam corrosion loop tests have demonstrated that 17-4 PH in the high- hardened, H900 condition is subject to stress corrosion cracking even at relatively low applied stresses in high purity water at temperatures ranging from 200 to 550 deg F. The test results indicate, however, that in the H1100 condition, 17-4 PH is resistant to stress corrosion cracking at stresses up to the yield point; and that only at very high stresses approaching 1% strain does there appear to be any danger of stress corrosion cracking. The number of tests on the other materials do not support firm conclusions, but it is evident that 17-7 PH (1050), A-286 (1325), and AM-350 (850 and 1000) are all less susceptible to stress corrosion cracking in high-purity water than 17-4 PH (H900). No failures of any of the steels, except 17-4 PH (H900), have occurred at applied stresses up to the yield strength. Only very high stresses over yield have produced any cracking in the limited tests performed. (N.W.R.)
- Research Organization:
- General Electric Co., San Jose, Calif.
- NSA Number:
- NSA-16-007859
- OSTI ID:
- 4802248
- Journal Information:
- Nuclear Eng., Vol. Vol: 7; Other Information: Orig. Receipt Date: 31-DEC-62
- Country of Publication:
- Country unknown/Code not available
- Language:
- English
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