Radiation-induced segregation in FFTF-irradiated austenitic stainless steels

Kenik, E.A.; Hojou, K.

doi:10.1016/0022-3115(92)90691-d

Cited by 35 publications

(22 citation statements)

References 7 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Furthermore, a sink effect link to GB boundary disorientation was also pointed out by Watanabe et al [14]. RID of Cr was also observed close to dislocation loops (Frank loops) in austenitic steels [15,16]. Both references provided support for a vacancy-driven RID mechanism.…”

Section: Introductionmentioning

confidence: 66%

“…Both references provided support for a vacancy-driven RID mechanism. Indeed, analytical measurements performed by Kenik et al [16] in the dilatational strain field of edge-on dislocation loops have not shown any enrichment, suggesting that the mechanism is not driven by the interstitial fluxes. More recently, torus-shaped clusters and rounded clusters were detected by atom probe tomography (APT) due to segregation on dislocation loops [15].…”

Section: Introductionmentioning

confidence: 93%

See 1 more Smart Citation

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

et al. 2014

View full text Add to dashboard Cite

Section: Introductionmentioning

confidence: 66%

Section: Introductionmentioning

confidence: 93%

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

et al. 2014

View full text Add to dashboard Cite

“…Contrary to grain boundary segregation, analyses on edge- 32) They found clear Cr depletion and Ni and Si enrichment at loops as large as 100 nm but no clear change at smaller loops. Qualitative evidence of Ni and Si enrichment at Frank loops was also reported in LWR-irradiated stainless steels although the detailed data were not described.…”

Section: Segregation and Precipitationmentioning

confidence: 99%

Evolution of Microstructure and Microchemistry in Cold-worked 316 Stainless Steels under PWR Irradiation

Fukuya¹,

Fujii²,

Nishioka³

et al. 2006

J. Nucl. Sci. Technol.

View full text Add to dashboard Cite

The evolution of microstructures and microchemistry was examined by transmission electron microscopy in coldworked SUS 316 stainless steel components irradiated in a pressurized water reactor to 1-73 dpa at 565-596 K. Homogenous nucleation of dislocation loops, helium bubbles and 0 precipitates was detected. The dislocation loops consisted of a high density of Frank loops and black dots. The black dots are considered to be small Frank loops, some fraction of which could be vacancy-type. The size distribution showed a double peak, which remained unchanged up to 73 dpa, suggesting that the dislocation structure was saturated under a balance of nucleation and disappearance. The helium bubbles were extremely dense and fine, resulting in swelling of less than 0.1%. Such bubble structure was formed under irradiation with a high He/dpa ratio. The 0 precipitation was detected at doses higher than 4 dpa with an increasing density for higher doses. The measured radiation hardening was almost explained by these visible microstructural features. Radiation-induced segregation at grain boundaries was confirmed to continuously develop up to 73 dpa whereas no significant segregation could be detected at Frank loops.

show abstract

“…However, to ensure that the yield stress evolution is only due to the Frank loops formation, a chemical characterization of the microstructure was performed. γ' precipitation [20,22] or Ni-Si enrichment at Frank loops [23] observations using conventional TEM are reported in the literature. This was completed more recently by A. Etienne owing to the development of the LATAP (Laser Assisted Tomographic Atom Probe) at the GPM …”

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

Radiation-induced segregation in FFTF-irradiated austenitic stainless steels

Cited by 35 publications

References 7 publications

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

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

Evolution of Microstructure and Microchemistry in Cold-worked 316 Stainless Steels under PWR Irradiation

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

Contact Info

Product

Resources

About