United States Nuclear Regulatory Commission - Protecting People and the Environment

Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys (NUREG/CR-6687)

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Publication Information

Manuscript Completed: May 2000
Date Published: October 2000

Prepared by:
H. M. Chung, W. E. Ruther,
R. V. Strain, W. J. Shack

Argonne National Laboratory
9700 South Cass Avenue
Argonne, IL 60439

M. McNeil, NRC Project Manager

Prepared for:
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

NRC Job Code L1429

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This report summarizes work performed at Argonne National Laboratory on irradiationassisted stress corrosion cracking (IASCC) of model austenitic stainless steels (SSs) that were irradiated in the Halden boiling heavy water reactor in simulation of the neutronic and environmental conditions of boiling water reactor (BWR) core internal components. Slowstrain-rate tensile tests in simulated BWR water were conducted on model austenitic stainless steel alloys that were irradiated at 2890C in helium to ≈0.3 x 1021 n-cm-2 and ≈0.9 x 1021 n-cm-2 (E > 1 MeV). Fractographic analysis by scanning electron microscopy was conducted to determine susceptibility to irradiation-assisted stress corrosion cracking as manifested by the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC) fracture surface morphology. As fluence was increased from ≈0.3 x 1021 n-cm-2 to ≈0.9 x 1021 n-cm-2, IGSCC fracture surfaces emerged in many alloys, usually in the middle of and surrounded by TGSCC fracture surfaces. Alloy-to-alloy variations in susceptibility to TGSCC and IGSCC were significant at ≈0.9 x 1021 n-cm-2. Susceptibility to TGSCC and IGSCC was influenced by more than one alloying and impurity element in a complex manner. Results from this study indicate that for commercial heats of Types 304 and 304L SS, a high concentration of S is detrimental and that a sufficiently low concentration of S is a necessary condition for good resistance to IASCC. A laboratory alloy containing a high concentration of Cr exhibited excellent resistance to TGSCC and IGSCC, despite a high S content, indicating that Cr atoms in high concentration play a major role in suppressing susceptibility to IASCC under BWR conditions. Yield strength of the model alloys, measured in BWR-like water at 289°C, was nearly constant at ≈200 MPa in the unirradiated state and was more or less independent of Si concentration. However, as the alloys were irradiated, the degree of increase in yield strength was significantly lower for alloys that contain >0.9 wt.% Si than for alloys that contain <0.8 wt.% Si, indicating that the nature of irradiation-induced hardening centers and the degree of irradiation hardening is significantly influenced by the Si content of the alloy. Similar influence was not observed for C and N. Results also indicate that a Si content between ≈0.8 and ≈1.5 wt.% is beneficial in delaying the onset of or suppressing the susceptibility to IASCC. Although susceptibility to TGSCC and IGSCC was insignificant and the fracture surface morphology was mostly ductile, some alloys exhibited very low uniform elongation and poor work-hardening capability in water after irradiation to ≈0.9 x 1021 n-cm-2. Such alloys contained unusually high concentrations of O or unusually low concentrations of Si or both. To better understand such behavior, usually indicative of severe susceptibility to dislocation channeling or other forms of localized deformation, a systematic microstructural investigation by transmission and scanning electron microscopies is desirable.

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