Atom probe and transmission electron microscopy study of aging of cast duplex stainless steels

Auger, Pierre; Danoix, F.; Menand, A.; Bonnet, Stéphanie; Bourgoin, J.; Guttmann, M.

doi:10.1179/mst.1990.6.3.301

Cited by 109 publications

(37 citation statements)

References 8 publications

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“…Aging of cast SSs at temperatures <500°C (<932°F) leads to precipitation of additional phases in the ferrite, e.g., formation of a Cr-rich α' phase by spinodal decomposition; nucleation and growth of α'; precipitation of a Ni-and Si-rich G phase, M 23 C 6 , and γ 2 (austenite); and additional precipitation and/or growth of existing carbides at ferrite/austenite phase boundaries. [8][9][10][11][12] Thermal embrittlement is caused primarily by formation of the Cr-rich α' phase in the ferrite and, to some extent, by precipitation and growth of carbides at phase boundaries. Thermal embrittlement of cast SSs results in brittle fracture associated with either cleavage of the ferrite or separation of the ferrite/austenite phase boundary.…”

Section: Introductionmentioning

confidence: 99%

Effects of thermal aging on fracture toughness and Charpy-impact strength of stainless steel pipe welds

Gavenda¹,

Michaud²,

Galvin³

et al. 1996

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The degradation of fracture toughness, tensile, and Charpy-impact properties of Type 308 stainless steel (SS) pipe welds due to thermal aging has been characterized at room temperature and 290°C. Thermal aging of SS welds results in moderate decreases in Charpy-impact strength and fracture toughness. For the various welds in this study, upper-shelf energy decreased by 50-80 J/cm 2 . The decrease in fracture toughness J-R curve or J IC is relatively small. Thermal aging had little or no effect on the tensile strength of the welds. Fracture properties of SS welds are controlled by the distribution and morphology of second-phase particles. Failure occurs by the formation and growth of microvoids near hard inclusions; such processes are relatively insensitive to thermal aging. The ferrite phase has little or no effect on the fracture properties of the welds. Differences in fracture resistance of the welds arise from differences in the density and size of inclusions. Mechanical-property data from the present study are consistent with results from other investigations. The existing data have been used to establish minimum expected fracture properties for SS welds.

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Section: Introductionmentioning

confidence: 99%

Effects of thermal aging on fracture toughness and Charpy-impact strength of stainless steel pipe welds

Gavenda¹,

Michaud²,

Galvin³

et al. 1996

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“…Formation of Cr-rich ∃% phase in the ferrite is the primary mechanism for thermal embrittlement of cast austenitic SSs; [28][29][30][31][32][33][34][35][36] thermal aging has little or no effect on the austenite phase. Embrittlement of ferrite phase from neutron irradiation occurs at lower fluences than does embrittlement of the austenite phase.…”

Section: Synergistic Effect Of Thermal and Neutron Irradiationmentioning

confidence: 99%

“…Thermal aging of cast SSs at 250-400°C (482-752°F) leads to precipitation of additional phases in the ferrite (e.g., formation of Cr-rich ∃% phase by spinodal decomposition; nucleation and growth of ∃%; precipitation of a Ni-and Si-rich G phase, M 23 C 6 carbide, and & 2 austenite; and additional precipitation and/or growth of existing carbides at the ferrite/austenite phase boundaries). [33][34][35][36] The formation of the Cr-rich ∃% phase by spinodal decomposition of ferrite is the primary mechanism for thermal embrittlement; it strengthens the ferrite phase by increasing strain hardening and the local tensile stress. Thermal aging has little or no effect on the austenite phase.…”

Section: Introductionmentioning

confidence: 99%

Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

Chopra¹,

Shack²

2008

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In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heataffected-zone materials, that were irradiated to fluence levels as high as ≈ 2 x 10 21 n/cm 2 (E > 1 MeV) (≈ 3 dpa) in a boiling heavy water reactor at 288-300°C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated. Paperwork Reduction Act StatementThis NUREG does not contain information collection requirements and, therefore, is not subject to the requirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Public Protection NotificationThe NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a current valid OMB control number. ForewordThis report presents the results of a study of simulated light-water reactor coolants, material chemistry, and irradiation damage and their effects on the susceptibility to stress-corrosion cracking of various commercially available and laboratory-melted stainless steels. This report is one of a series dating back about 8 years, describing such results, which are required to support analysis of the structural integrity of reactor internal components, many of which are subject to irradiation-assisted stress-corrosion cracking (IASCC).The earlier reports detailed crack growth rates in heat-affected zones adjacent to stainless steel weldments, and they comprised the final publications based on specimens irradiated in Phase I (of two) in the Halden test reactor. Phase I irradiations principally involved stainless steels of wide-ranging chemistry (including commercial steels of typical chemistry) and conventional heat treatment and product form processing. By contrast, this report is the first to present data from specimens irradiated in Phase II, which featured a variety of innovatively fabricated and engineered ...

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“…Cast stainless steels with low activation energy and slow embrittlement at 400°C show G-phase precipitation after aging, and steels with high activation energy and fast embrittlement at 400°C do not contain a G phase. [8][9][10]17,18 The presence of Ni-Si-Mo clusters in the ferrite matrix of an unaged material may be considered a signature of steels that are potentially sensitive to thermal embrittlement, i.e., steels with Ni-Si-Mo clusters in the ferrite matrix show low activation energy for thermal embrittlement but take longer to embrittle at 400°C.…”

Section: Mechanism Of Thermal Embrittlementmentioning

confidence: 99%

“…1,2,[8][9][10][17][18][19][20][21] Thermal aging has little or no effect on the austenite phase. The formation of Cr-rich α' phase by spinodal decomposition of ferrite is the primary mechanism for thermal embrittlement.…”

Section: Mechanism Of Thermal Embrittlementmentioning

confidence: 99%

Tensile-property characterization of thermally aged cast stainless steels.

Michaud¹,

Toben²,

Soppet³

et al. 1994

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The effect of thermal aging on tensile properties of cast stainless steels during service in light water reactors has been evaluated. Tensile data for several experimental and commercial heats of cast stainless steels are presented. Thermal aging increases the tensile strength of these steels. The high-C Mo-bearing CF-8M steels are more susceptible to thermal aging than the Mo-free CF-3 or CF-8 steels. A procedure and correlations are presented for predicting the change in tensile flow and yield stresses and engineering stress-vs.-strain curve of cast stainless steel as a function of time and temperature of service. The tensile properties of aged cast stainless steel are estimated from known material information, i.e., chemical composition and the initial tensile strength of the steel. The correlations described in this report may be used for assessing thermal embrittlement of cast stainless steel components. Executive SummaryCast duplex stainless steels used in light water reactor (LWR) systems for primary pressure-boundary components are susceptible to thermal embrittlement at reactor operating temperatures. Thermal aging of cast stainless steels at these temperatures causes an increase in hardness and tensile strength and a decrease in ductility, impact strength, and fracture toughness of the material and the Charpy transition curve shifts to higher temperatures. Investigations at Argonne National Laboratory have shown that thermal embrittlement of cast stainless steel components may occur within the reactor design lifetime of 40 yr. Various grades and heats of cast stainless steel exhibit varying degrees of thermal embrittlement. In general, the low-C CF-3 steels are the most resistant t o thermal embrittlement, and the Mobearing, high-C CF-8M steels are the least resistant. An assessment of mechanical-property degradation due to thermal embrittlement is therefore required to evaluate the performance of cast stainless steel components during prolonged exposure to service temperatures, because rupture of the primary pressure boundary could lead to a loss-of-coolant accident and possible exposure of the public to radiation.This report presents tensile-property data on several heats of cast stainless steels aged up to 58,000 h at temperatures between 290 and 450°C (554 and 752°F). The tensile data are analyzed to establish the effects of thermal aging on tensile strength and engineering stress-strain behavior (represented by the Ramberg-Osgood equation) of cast stainless steels. A procedure and correlations are presented for predicting the change in tensile flow and yield stress, and in the engineering stress-vs.-strain curve of cast stainless steel components due to thermal aging during service in LWRs. The tensile properties of aged cast stainless steel are estimated from information that is readily available from certified material test records for the component, i.e., chemical composition and the initial tensile strength of the unaged material. The correlations described in this report may be used ...

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Atom probe and transmission electron microscopy study of aging of cast duplex stainless steels

Cited by 109 publications

References 8 publications

Effects of thermal aging on fracture toughness and Charpy-impact strength of stainless steel pipe welds

Effects of thermal aging on fracture toughness and Charpy-impact strength of stainless steel pipe welds

Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

Tensile-property characterization of thermally aged cast stainless steels.

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