Overview of microstructural evolution in neutron-irradiated austenitic stainless steels

Maziasz, P.J.

doi:10.1016/0022-3115(93)90077-c

Cited by 254 publications

(117 citation statements)

References 62 publications

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“…24,25 Studies on irradiation microstructure show that, although the saturation dose for various irradiation defects varies from 0.1 dpa to 30 or 40 dpa, the total dislocation density in austenitic SSs reaches a common saturation value at around 5 dpa. 26,27,28 The dose range for the BOR-60 and Halden specimens in this study is close to, or slightly below, the saturation dose for austenitic SSs. This work Chung et al [20] Chen et al [19] Alexander et al [21] Mills et al [22] Sindelar et al [23] YS ( Yield strength of SA austenitic SSs irradiated at 90 to 427°C and tested in various environments between room temperature and 427°C.…”

Section: Irradiation Effects On the Ssrt Propertiessupporting

confidence: 56%

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Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

Chen¹,

Chopra²,

Soppet³

et al. 2010

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Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320°C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in highdissolved-oxygen environment, and TEM characterization of irradiation defect structure.Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5-and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials.Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to 25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundaryengineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

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Section: Irradiation Effects On the Ssrt Propertiessupporting

confidence: 56%

“…The irradiation microstructure is a function of irradiation temperature. 30,31,32 The point-defect survival rate and the microstructural evaluation can be strongly affected by the irradiation temperature. For austenitic SSs, vacancies resulting from displacement damage become mobile at 300°C.…”

Section: Irradiation Effects On the Ssrt Propertiesmentioning

confidence: 99%

Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

Chen¹,

Chopra²,

Soppet³

et al. 2010

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“…For the same irradiation level, the CGRs for weld HAZ materials were higher than those for solution-annealed SSs. Results in the literature suggest that the CGRs of SSs irradiated to higher fluence levels (e.g., 8.67 x 10 21 n/cm 2 or 13 dpa) strongly depend on the stress intensity factor (K) and can be up to a factor of 30 higher than the NUREG-0313 disposition curve.…”

Section: Crack Growth Rate Testsmentioning

confidence: 99%

“…[7][8][9] At irradiation temperatures above 300°C (572°F), the microstructure consists of larger faulted loops, network dislocations, and cavities that are threedimensional clusters (voids) of vacancies and/or gas bubbles. The microchemistry of the material is also changed due to radiation-induced segregation (RIS).…”

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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“…These defects interact with each other during their nucleation and growing stages, 10) making the scenario complicated for mechanism analysis. Thus, such interactions need to be simplified before Si's role could be revealed.…”

Section: Introductionmentioning

confidence: 99%

Radiation Defects Formed in Ion-Irradiated 316L Stainless Steel Model Alloys with Different Si Additions

et al. 2015

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The 304/316 series of austenite stainless steels are used in light water reactors as structural materials. As a result of the high temperatures and neutron irradiation in reactor, dislocation defects will form in stainless steel, causing an increase in the hardness and a decrease in the ductility of the material. In this work, high purity 316L stainless steel model alloys with three different Si contents were ion irradiated at 290°C or 400°C to investigate the black dot and Frank loop formation mechanism influenced by Si addition. Black dot defect formation mainly occurs at 290°C. It is Frank loop in nature with its formation not affected by Si addition. Frank loop is the main defect at 400°C, and both loop density and the average size are substantially suppressed by Si addition. This may be caused by silicon's role in enhancing effective vacancy diffusivity and thus promoting recombination. The trend of irradiation hardening measured verses temperature matches the microstructure observed.

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Overview of microstructural evolution in neutron-irradiated austenitic stainless steels

Cited by 254 publications

References 62 publications

Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

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

Radiation Defects Formed in Ion-Irradiated 316L Stainless Steel Model Alloys with Different Si Additions

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