The evolution of mechanical property change in irradiated austenitic stainless steels

Lucas, G.E.

doi:10.1016/0022-3115(93)90129-m

Cited by 398 publications

(171 citation statements)

References 42 publications

Supporting

Mentioning

168

Contrasting

Order By: Relevance

“…As a result, the material exhibits a substantial increase in yield stress (Fig. 2), reduction in ductility, swelling loss of corrosion resistance which may be damaging to the proper operation of the reactor [1][2][3][4][5][6][7]. For instance cracks were observed in bolts (Fig.…”

Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

View full text Add to dashboard Cite

Abstract:Even if temperature, pressure and chemistry of the cooling water are not very high and aggressive, materials used in PWRs (Pressurized Water Reactors) are exposed to different degradation mechanisms. One of the main goals of the research programs in this field is to develop physical model of the mechanisms down to the atomic scale. Such approach needs a clear description and understanding of the degradation mechanisms at the same small scale. This paper illustrates the benefits of different microscopies and of their last improvements up to the promising possibilities of monochromated and aberrations corrected TEM/STEM. A specific focus is placed on four different degradation mechanisms observed in austenitic stainless steel: irradiation ageing, corrosion fatigue, stress corrosion cracking and corrosion.

show abstract

Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

View full text Add to dashboard Cite

show abstract

“…572°F). 10 In Fig. 73, the J Ic values for all of the data at neutron exposures greater than 20 dpa may overestimate the toughness for irradiation temperatures of 290-320°C (554-608°F) because the irradiation temperatures were above 300°C (572°F).…”

Section: Comparison With Fracture Toughness Data In the Literaturementioning

confidence: 95%

“…10 This process is also termed "dislocation channeling," whereby dislocation motion along a narrow band of slip planes clears the irradiation-induced defect structure, creating a defect-free channel that offers less resistance to subsequent dislocation motion or deformation. The enhanced planar slip leads to a pronounced degradation in the fracture toughness of austenitic SSs.…”

Section: Introductionmentioning

confidence: 99%

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

Chopra¹,

Shack²

2008

View full text Add to dashboard Cite

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 ...

show abstract

“…Post-irradiation tests show that for 300 series SS substantial yield stress (YS) increases and ductility and fracture toughness decreases occur in these alloys as reviewed by Lucas [15]. Yield strengths as much as 4 to 6 times the unirradiated value have been observed [16], while uniform elongation can drop to < 3%.…”

Section: Tensile Behavior Of Irradiated Stainless Steelmentioning

confidence: 98%

Nuclear energy plant optimization (NEPO) final scientific/technical report

Cole¹,

Daum²,

Tsai³

et al. 2004

View full text Add to dashboard Cite

The evolution of mechanical property change in irradiated austenitic stainless steels

Cited by 398 publications

References 42 publications

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

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

Nuclear energy plant optimization (NEPO) final scientific/technical report

Contact Info

Product

Resources

About