Modelling of the effect of dislocation channel on intergranular microcrack nucleation in pre-irradiated austenitic stainless steels during low strain rate tensile loading

Evrard, Pierre; Sauzay, Maxime

doi:10.1016/j.jnucmat.2010.06.006

Cited by 30 publications

(15 citation statements)

References 61 publications

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“…Most of the prior FE modeling efforts [78][79][80] have explicitly modeled dislocation channels in the material, i.e., explicitly embedded 'softer' shear bands in a 'harder' matrix. In contrast, we started out with a bulk consisting of uniformly distributed irradiation-induced defects and let the microstructure evolve according to constitutive rules based on the physics of the material behavior under mechanical loading.…”

Section: Discussionmentioning

confidence: 99%

Continuum modeling of localized deformation in irradiated bcc materials

Patra

McDowell

2013

Journal of Nuclear Materials

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Section: Discussionmentioning

confidence: 99%

Continuum modeling of localized deformation in irradiated bcc materials

Patra

McDowell

2013

Journal of Nuclear Materials

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“…Recent EBSD measurements of misorientations in PWR-irradiated CW type 316 SSs showed that macroscopic elastic deformation at ∼50% of the yield strength caused detectable strains near GBs [312]. Recent modeling studies for channel-GB interactions also showed a similar role to that of dislocation pileups [313,314]. Although more Figure 27.…”

Section: Role Of Deformationmentioning

confidence: 97%

Current understanding of radiation-induced degradation in light water reactor structural materials

Fukuya¹

2013

Journal of Nuclear Science and Technology

161

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Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.

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“…The high amount of strain in the dislocation channels is accompanied by elevated levels of stress as predicted by computer modeling. Evrard and Sauzay [14], [15] used a finite element model to simulate the intersection between channels and grain boundaries. Dislocation channels were modeled as soft regions (low critical resolved shear stress) in a hard grain (high critical resolved shear stress).…”

Section: Introductionmentioning

confidence: 99%

Mechanism of dislocation channel-induced irradiation assisted stress corrosion crack initiation in austenitic stainless steel

McMurtrey

Cui

Robertson

et al. 2015

Current Opinion in Solid State and Materials Science

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The mechanism by which dislocation channeling induces irradiation assisted stress corrosion cracking was determined using Fe-13Cr15Ni austenitic stainless steel irradiated with protons to a dose of 5 dpa and strained at high temperature in both argon and simulated boiling water reactor normal water chemistry environments. Straining induced dislocation channels that were characterized by digital image correlation and confocal microscopy. Dislocation channels were found to be either continuous across the boundary, discontinuous, or discontinuous with slip in the boundary. Discontinuous channels were found to contain the least amount of strain but have the highest propensity for initiating cracks. Discontinuous dislocation channel-grain boundary intersections were shown to have the highest local stress. TEM in-situ straining of irradiated steels and atomistic simulation of dislocation-grain boundary interaction provided supporting evidence that channels that were unable to transfer strain underwent cracking. The inability of channels to relieve stress, by either slip in the adjacent grain or in the grain boundary, resulted in high local stresses and increased susceptibility to stress corrosion cracking initiation.

show abstract

Modelling of the effect of dislocation channel on intergranular microcrack nucleation in pre-irradiated austenitic stainless steels during low strain rate tensile loading

Cited by 30 publications

References 61 publications

Continuum modeling of localized deformation in irradiated bcc materials

Continuum modeling of localized deformation in irradiated bcc materials

Current understanding of radiation-induced degradation in light water reactor structural materials

Mechanism of dislocation channel-induced irradiation assisted stress corrosion crack initiation in austenitic stainless steel

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