The stress corrosion cracking behavior of alloys 690 and 152 WELD in a PWR environment.
- Nuclear Engineering Division
Alloys 690 and 152 are the replacement materials of choice for Alloys 600 and 182, respectively. The latter two alloys are used as structural materials in pressurized water reactors (PWRs) and have been found to undergo stress corrosion cracking (SCC). The objective of this work is to determine the crack growth rates (CGRs) in a simulated PWR water environment for the replacement alloys. The study involved Alloy 690 cold-rolled by 26% and a laboratory-prepared Alloy 152 double-J weld in the as-welded condition. The experimental approach involved pre-cracking in a primary water environment and monitoring the cyclic CGRs to determine the optimum conditions for transitioning from the fatigue transgranular to intergranular SCC fracture mode. The cyclic CGRs of cold-rolled Alloy 690 showed significant environmental enhancement, while those for Alloy 152 were minimal. Both materials exhibited SCC of 10{sup -11} m/s under constant loading at moderate stress intensity factors. The paper also presents tensile property data for Alloy 690TT and Alloy 152 weld in the temperature range 25--870 C.
- Research Organization:
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Organization:
- USNRC
- DOE Contract Number:
- DE-AC02-06CH11357
- OSTI ID:
- 1013984
- Report Number(s):
- ANL/NE/CP-62061; TRN: US1102618
- Resource Relation:
- Conference: ASME Pressure Vessels and Piping Conference; Jul. 27, 2008 - Jul. 31, 2008; Chicago, IL
- Country of Publication:
- United States
- Language:
- ENGLISH
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