Oxide-Induced Initiation of Stress Corrosion Cracking in Irradiated Stainless Steel

Cookson, John; Was, Gary S.; Andresen, Peter L.

doi:10.5006/1.3284856

Cited by 14 publications

(6 citation statements)

References 28 publications

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“…Although C significantly increases the yield strength of irradiated SSs, higher C content seems to be either benign or conducive to lower susceptibility to intergranular cracking of irradiated materials. 6 5 Deleterious effects of 0 in steels have been reported by Chung et al 6 5 and Cookson et al 66 Indications of the deleterious effect of grain-boundary segregation of N have been reported for BWR neutron absorber tubes. 6 5 Similar reports suggest that a higher concentration of N is deleterious, at least under BWR conditions.…”

Section: Irradiation-assisted Stress Corrosion Cracking Of Austeniticmentioning

confidence: 94%

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

Chopra¹,

Chung²,

Gruber³

et al. 1999

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This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiationassisted stress corrosion cracking of Type 304 SS, IC) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288°C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRS to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289°C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.iii 4. 5. 6. 7.8. 18.19. 20.21. 22.23. 24.25. 26. 27.28. 29.30. 31. 32.33. Plain carbon steels (CSs) and low-alloy steels (LASS) are used extensively in steam supply systems of pressurized and boiling water nuclear reactors (PWRS and BWRs) as piping and pressure vessel materials. Fatigue tests are being conducted on C S s and LASs in water and in air to establish the effects of material and loading variables on fatigue life. The results indicate a significant decrease in fatigue life when five conditions are satisfied simultaneously, viz.. when applied strain range, service temperature, dissolved-oxygen (DO) in the water, and s u l h r content of the steel are above a minimum threshold level, and the loading strain rate is below a threshold value. Only a moderate decrease in fatigue life is observed in light water reactor (LWR) environments when any one threshold condition is not satisfied. This report presents (a) results of fatigue tests that examine the influence of the reactor environment on crack initiation and crack growth of C S s and LASs, and (b) data on austenitic stainless steel (SS) that establish the effects of various loading and environmental variables on fatigue lives of these steels.Crack lengths as a function of fatigue cycles were determined for A533-Gr B LAS and A106-Gr B CS in air and water environments. The results indicate that, in air, fatigue cracks that are 10 p m or longer, form quite early during fatigue (

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Section: Irradiation-assisted Stress Corrosion Cracking Of Austeniticmentioning

confidence: 94%

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

Chopra¹,

Chung²,

Gruber³

et al. 1999

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“…Studies by both Cookson and Was (1998) and Chung, Ruther, Sanecki, and Kassner (1992) showed that oxide inclusions were excellent sites for nucleation of IASCC cracks. IASCC is promoted by the fracture of brittle oxides that intersect grain boundaries, leading to crack extension along the grain boundary.…”

Section: Low Inclusion Densitymentioning

confidence: 96%

Irradiation-assisted stress corrosion cracking

Was¹,

Ashida²,

Andresen³

2011

Corrosion Reviews

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Irradiation-assisted stress corrosion cracking (IASCC) refers to the process(es) by which irradiation enhances the inherent susceptibility of alloys to stress corrosion cracking (SCC). This article begins with an introduction to the IASCC problem experienced in light water reactors, followed by a summary of existing plant and laboratory data that reveal its dependence on irradiation, environment and alloy parameters. The specifi c effects of irradiation on IASCC are classifi ed into two categories: the effects on water chemistry and the effects on microstructure. Water chemistry effects include radiolysis and its effects on corrosion potential, and effects of corrosion potential on IASCC. Microstructure effects include radiationinduced segregation (RIS), irradiated microstructure, swelling and creep, and hydrogen and helium generation. Leading mechanisms proposed to explain the role of radiation in the SCC process are: radiolysis and crack tip strain rate, grain boundary chromium depletion, irradiation hardening, localized deformation and RIS of minor elements. As nuclear power plants operate for longer, an increased incidence of SCC can be expected unless active mitigation steps are taken. Drawing on the accumulated understanding over the past four decades of the key processes believed to control IASCC, a set of attributes of an ideal, IASCC-resistant alloy are identifi ed as a fi rst step.

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“…23,24 Deleterious effects of O on IGSCC of proton-irradiated steels have been reported by Cookson et al, who showed that oxides near the surface facilitate crack nucleation. 25 High concentrations of O and internal oxide precipitates were also reported for grain boundaries in the heat-affected zones of unirradiated Types 304 and 304L SSs and field-cracked BWR core shroud welds that were fabricated by the shielded-metal-arc procedure. 26,27 Yet, the exact role and the source of O on grain boundaries in IASCC of a solution-annealed material or intergranular failure of a BWR core internal weld remain poorly understood.…”

Section: Introductionmentioning

confidence: 93%

“… [19][20][21][22][23][24][25][26][27]. strongly indicate that S concentration of ≤0.002 wt.% is a necessary but not a sufficient condition for a 304-or 316-type steel to be resistant to IASCC under BWR-like conditions.Figure 20.…”

mentioning

confidence: 99%

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

Chung¹,

Shack²

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 ≈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.This page is intentionally blank.iv Foreword v This report examines the effects of simulated light-water reactor coolants, material chemistry, and irradiation damage on the susceptibility to stress corrosion cracking of a series of commercially available and laboratory-melted stain...

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Oxide-Induced Initiation of Stress Corrosion Cracking in Irradiated Stainless Steel

Cited by 14 publications

References 28 publications

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

Irradiation-assisted stress corrosion cracking

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

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