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| Home > Electronic Reading Room > Document Collections > NUREG-Series Publications > Publications Prepared by NRC Contractors         > NUREG/CR-6892
Irradiation-Assisted Stress         Corrosion Cracking Behavior of         Austenitic Stainless Steels         Applicable to LWR Core Internals (NUREG/CR-6892)
On this page:
- Publication Information
- Abstract

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- NUREG/CR-6892
Cover Pages, Chapters 1 - 8 (PDF - 2.76 MB)
Chapter 9 - References (PDF - 11.7 MB)

Publication Information
**Manuscript Completed: March 2005      Date Published: January 2006**

Prepared by        H. M. Chung, W. J. Shack

Energy Technology Division        Argonne National Laboratory        9700 South Cass Avenue      Argonne, IL 60439

W. H. Cullen, Jr., 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 Y6388

Availability         Notice
Abstract
This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress        corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in        simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components.        Slow-strain-rate tensile tests in BWR-like oxidizing water were conducted on 27 austenitic stainless steel        alloys that were irradiated at 288°C in helium to 0.4, 1.3, and 3.0 dpa. Fractographic analysis was conducted        to determine the fracture surface morphology. Microchemical analysis by Auger electron spectroscopy was        performed on BWR neutron absorber tubes to characterize grain-boundary segregation of important elements        under BWR conditions. At 0.4 and 1.4 dpa, transgranular fracture was mixed with intergranular fracture. At        3 dpa, transgranular cracking was negligible, and fracture surface was either dominantly intergranular, as in        field-cracked core internals, or dominantly ductile or mixed. This behavior indicates that percent        intergranular stress corrosion cracking determined at ≈3 dpa is a good measure of IASCC susceptibility. At        ≈1.4 dpa, a beneficial effect of a high concentration of Si (0.8-1.5 wt.%) was observed. At ≈3 dpa, however,        such effect was obscured by a deleterious effect of S. Excellent resistance to IASCC was observed up to ≈3        dpa for eight heats of Types 304, 316, and 348 steel that contain very low concentrations of S. Susceptibility        of Types 304 and 316 steels that contain >0.003 wt.% S increased drastically. This indicates that a sulfurrelated        critical phenomenon plays an important role in IASCC. A sulfur content of <0.002 wt.% is the        primary material factor necessary to ensure good resistance to IASCC. However, for Types 304L and 316L        steel and their high-purity counterparts, a sulfur content of <0.002 wt.% alone is not a sufficient condition to        ensure good resistance to IASCC. This is in distinct contrast to the behavior of their high-C counterparts. At        S concentrations >0.002 wt.%, the deleterious effect of S is so dominant that a high concentration of C is not        an important factor. A two-dimensional map was developed in which susceptibility or resistance to IASCC is        shown as a function of bulk concentrations of S and C. Data reported in the literature are consistent with the        map. The map is helpful to predict relative IASCC susceptibility of Types 304 and 316 steels. A similar but        somewhat different map is helpful to predict IASCC behavior of Type 348 steels. Grain-boundary        segregation of S was observed for BWR neutron absorber tubes irradiated to ≈3 dpa. On the basis of the        results of the stress-corrosion-cracking tests and the microstructural characterization, a mechanistic IASCC      model has been developed. |
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