Chromium enrichment on the habit plane of dislocation loops in ion-irradiated high-purity Fe–Cr alloys

Bhattacharya, Arunodaya; Meslin, E.; Henry, J.; Pareige, C.; Décamps, B.; Genevois, Cécile; Brimbal, Daniel; Barbu, A.

doi:10.1016/j.actamat.2014.06.050

Cited by 48 publications

(30 citation statements)

References 41 publications

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“…Carbon can trap vacancies, which has been shown to inhibit swelling and loop growth [69,70]. Thus, M 2 X can interact both directly, by acting as a trap for vacancies, and indirectly, by removing carbon from solution and reducing the number of traps in the bulk.…”

Section: Accepted Manuscriptmentioning

confidence: 99%

“…To understand this interaction, the role of M 2 X was considered both as a sink for defects and as a sink for chromium and carbon. Both chromium and carbon have been observed to suppress loop growth relative to that in pure iron [56,69,70] due to enrichment and pinning of loops.…”

Section: X Precipitatesmentioning

confidence: 99%

“…M 2 X can also be interpreted as both a sink for defects ( Figure 12) and as a consumer of carbon. Carbon in solution can trap vacancies, which has been demonstrated to limit both loop and void growth [56,69,70]. As carbon is precipitated out of solution, there should be a commensurate decrease in carbon available to pin dislocations, thus allowing for increased loop growth; a possible interaction mechanism and its effect on voids and dislocations will be further addressed later.…”

Section: A C C E P T E D Accepted Manuscriptmentioning

confidence: 99%

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The co-evolution of microstructure features in self-ion irradiated HT9 at very high damage levels

Getto

VanCoevering

Was

2017

Journal of Nuclear Materials

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Understanding the void swelling and phase evolution of reactor structural materials at very high damage levels is essential to maintaining safety and longevity of components in Gen IV fast reactors. A combination of ion irradiation and modeling was utilized to understand the microstructure evolution of ferritic-martensitic alloy HT9 at high dpa. Self-ion irradiation experiments were performed on alloy HT9to determine the co-evolution of voids, dislocations and precipitates up to 650 dpa at 460°C. Modeling of microstructure evolution was conducted using the modified Radiation Induced Microstructure Evolution (RIME) model, which utilizes a mean field rate theory approach with grouped cluster dynamics.Irradiations were performed with 5 MeV raster-scanned Fe 2+ ions on samples pre-implanted with 10 atom parts per million He. The swelling, dislocation and precipitate evolution at very high dpa was determined using Analytical Electron Microscopy in Scanning Transmission Electron Microscopy (STEM) mode.Experimental results were then interpreted using the RIME model. A microstructure consisting only of dislocations and voids is insufficient to account for the swelling evolution observed experimentally at high damage levels in a complicated microstructure such as irradiated alloy HT9. G phase was found to have a minimal effect on either void or dislocation evolution. M 2 X plays two roles; a variable biased sink for defects, and as a vehicle for removal of carbon from solution, thus promoting void growth. When accounting for all microstructure interactions, swelling at high damage levels is a dynamic process that continues to respond to other changes in the microstructure as long as they occur.

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Section: Accepted Manuscriptmentioning

confidence: 99%

Section: X Precipitatesmentioning

confidence: 99%

Section: A C C E P T E D Accepted Manuscriptmentioning

confidence: 99%

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The co-evolution of microstructure features in self-ion irradiated HT9 at very high damage levels

Getto

VanCoevering

Was

2017

Journal of Nuclear Materials

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“…The choice is based on excellent mechanical properties at high temperatures, good weldability, low thermal expansion coefficient, high thermal conductivity, adequate corrosion and stress-corrosion cracking resistances and high inherent void swelling resistance on neutron irradiation [3][4][5]. The ASME grade 91 and 92 (9Cr-1.8W-0.5Mo-VNb) steels are used extensively in the advanced fossil fired power plants and is being considered for structural applications in advanced sodium cooled fast reactors [6].…”

Section: Introductionmentioning

confidence: 99%

An assessment of creep deformation and rupture behaviour of 9Cr–1.8W–0.5Mo–VNb (ASME grade 92) steel

Sakthivel

Selvi

Laha

2015

Materials Science and Engineering: A

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“…Multiple studies have measured radiation induced segregation (RIS) in BCC Fe-Cr steels by collecting line scans using STEM-EDX. [5][6][7][8][9] This section will describe the methods contributing to a sound RIS measurement, including experimental setup, error analysis, and measurement limitations. Examples of STEM-EDX RIS measurements in BCC Fe-Cr steels will also be illustrated.…”

Section: Ris Measurementsmentioning

confidence: 99%

Application of STEM characterization for investigating radiation effects in BCC Fe-based alloys

Parish

Field

Certain³

et al. 2015

J. Mater. Res.

107

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This paper provides an overview of advanced scanning transmission electron microscopy (STEM) techniques used for characterization of irradiated BCC Fe-based alloys. Advanced STEM methods provide the high-resolution imaging and chemical analysis necessary to understand the irradiation response of BCC Fe-based alloys. The use of STEM with energy dispersive x-ray spectroscopy (EDX) for measurement of radiation-induced segregation (RIS) is described, with an illustrated example of RIS in proton-and self-ion irradiated T91. Aberration-corrected STEM-EDX for nanocluster/nanoparticle imaging and chemical analysis is also discussed, and examples are provided from ion-irradiated oxide dispersion strengthened (ODS) alloys. Finally, STEM techniques for void, cavity, and dislocation loop imaging are described, with examples from various BCC Fe-based alloys.

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Chromium enrichment on the habit plane of dislocation loops in ion-irradiated high-purity Fe–Cr alloys

Cited by 48 publications

References 41 publications

The co-evolution of microstructure features in self-ion irradiated HT9 at very high damage levels

The co-evolution of microstructure features in self-ion irradiated HT9 at very high damage levels

An assessment of creep deformation and rupture behaviour of 9Cr–1.8W–0.5Mo–VNb (ASME grade 92) steel

Application of STEM characterization for investigating radiation effects in BCC Fe-based alloys

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