Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals.

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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 ... continued below

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Chung, H. M.; Shack, W. J. & Technology, Energy January 31, 2006.

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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 {approx}3 dpa is a good measure of IASCC susceptibility. At {approx}1.4 dpa, a beneficial effect of a high concentration of Si (0.8-1.5 wt.%) was observed. At {approx}3 dpa, however, such effect was obscured by a deleterious effect of S. Excellent resistance to IASCC was observed up to {approx}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 sulfur related 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 {approx}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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  • Report No.: ANL-04/10
  • Grant Number: DE-AC02-06CH11357
  • DOI: 10.2172/915725 | External Link
  • Office of Scientific & Technical Information Report Number: 915725
  • Archival Resource Key: ark:/67531/metadc877362

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  • January 31, 2006

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  • Sept. 22, 2016, 2:13 a.m.

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  • Sept. 28, 2016, 6:16 p.m.

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Chung, H. M.; Shack, W. J. & Technology, Energy. Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals., report, January 31, 2006; United States. (digital.library.unt.edu/ark:/67531/metadc877362/: accessed December 18, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.