Inhibition of stress corrosion cracking of Alloy X-750 by prestrain
Tests of precracked and as-notched compact tension specimens were conducted in 3600C hydrogenated water to determine the effect of prestrain on the stress corrosion cracking (SCC) resistance of Alloy X-750 in the HTH, AH and HOA heat treated conditions. Prestraining is defined as the intentional application of an initial load (or strain) that is higher than the final test load. Prestrain was varied from 10% to 40% (i.e., the initial to final load ratios ranged from 1.1 to 1.4). Other variables included notch root radius, stress level and irradiation. Specimens were bolt-loaded to maintain essentially constant displacement conditions during the course of the test. The frequent heat up and cooldown cycles that were necessary for periodic inspections provided an opportunity to evaluate the effect of test variables on rapid low temperature crack propagation to which this alloy is subject. For Condition HTH, application of 20% to 40% prestrain either eliminates or significantly retards SCC initiation in as-notched specimens and the onset of crack growth in precracked specimens. In addition, this procedure reduces the propensity for low temperature crack growth during cooldown. Similar results were observed for precracked HOA specimens. Application of 20% prestrain also retards SCC in as-notched and precracked AH specimens, but the effects are not as great as in Condition HTH. Prestraining at the 10% level was found to produce an inconsistent benefit. In-reactor SCC testing shows that prestrain greatly improves the in-flux and out-of-flux SCC resistance of Condition HTH material. No SCC was observed in precracked specimens prestrained 30%, whereas extensive cracking was observed in their nonprestrain counterparts.
- Research Organization:
- Westinghouse Electric Corp., West Mifflin, PA (United States)
- Sponsoring Organization:
- USDOE Assistant Secretary for Nuclear Energy, Washington, DC (United States)
- DOE Contract Number:
- AC11-93PN38195
- OSTI ID:
- 501536
- Report Number(s):
- WAPD-T-2988; CONF-970832-1; ON: DE97004595; TRN: 97:013013
- Resource Relation:
- Conference: 8. international symposium on environmental degradation of materials in nuclear power systems-water reactors, Amelia Island, FL (United States), 10-14 Aug 1997; Other Information: PBD: Mar 1997
- Country of Publication:
- United States
- Language:
- English
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