Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys

Haley, Jack C.; Briggs, Samuel A.; Edmondson, Philip D.; Sridharan, Kumar; Roberts, Steve; Lozano-Pérez, Sergio; Field, Kevin G.

doi:10.1016/j.actamat.2017.07.011

Cited by 105 publications

(21 citation statements)

References 33 publications

(86 reference statements)

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The 〈100〉 type loops are more stable at higher temperatures [34], with the loop Burgers vectors transitioning from majority 〈111〉 to majority 〈100〉 between 300°C and 400°C [35]. The observation of a greater number of 〈100〉 loops for a constant temperature usually suggests that 〈111〉 defects (produced in-cascades) were more mobile [74,75], as it implies 〈100〉 loop production via the 111-mechanism [72] occurs more frequently. This suggests that the loops produced during self-ion irradiation may have been more mobile than loops produced during neutron irradiation, despite the neutron irradiation being the hotter experiment.…”

Section: -On the Formation Of Dislocation Loopsmentioning

confidence: 99%

Microstructural examination of neutron, proton and self-ion irradiation damage in a model Fe9Cr alloy

Haley

Shubeita

Wady

et al. 2020

Journal of Nuclear Materials

View full text Add to dashboard Cite

Transmission electron microscopy (TEM) was used to compare the microstructural defects produced in an Fe9Cr model alloy during exposure to neutrons, protons, or self-ions. Samples from the same model alloy were irradiated using fission-neutrons, 2MeV Fe+ ions or 1.2MeV protons at similar temperatures (~300°C) and similar doses (~2.0dpa). The neutron-irradiated alloy contained visible interstitial dislocation loops with b=〈111〉, and on average ~5nm in size. The density varied from 2±1x10 20 m -3 (in the matrix far from dislocations and boundaries) to 1.2±0.3x10 23 m -3 (close to helical dislocation lines). Chromium α'-phase precipitates were also identified at a density of 7.4±0.4x10 23 m -3 . Self-ion irradiation produced mostly homogeneously distributed dislocation loops (6-7nm on average), and with a greater fraction of 〈100〉 loops (~40%) than was seen in the neutron-irradiated alloy, and at a density of 6.8±0.8 x10 22 m -3 . In contrast to the loops produced by neutron irradiation, the self-ion irradiated Fe9Cr contained only vacancy-type loops. Chromium also remained in solution. Proton-irradiated Fe9Cr contained interstitial dislocation loops close to helical-dislocation segments, similar to the neutron-irradiated sample. Chromium α'-phases were also identified in the proton-irradiated sample at a density of 2.5±0.3x10 23 m -3 , and large voids (up to 7nm) were found at a density over 10 22 m -3 . Like the neutron-irradiated sample, the density of dislocation loops was also heterogeneously distributed; far from grain boundaries and dislocation lines the density was 2.5±0.4 x10 22 m -3 , while close to helical dislocation lines the density was 8.1±1.3 x10 22 m -3 .

show abstract

Section: -On the Formation Of Dislocation Loopsmentioning

confidence: 99%

Microstructural examination of neutron, proton and self-ion irradiation damage in a model Fe9Cr alloy

Haley

Shubeita

Wady

et al. 2020

Journal of Nuclear Materials

View full text Add to dashboard Cite

show abstract

“…Neutron and ion irradiation in the temperature range of 334°C to 384°C up to 2.5 dpa have been shown to produce a mixed population of dislocation loops with either a Burgers vector of 2 ⁄ 〈111〉 which can be either interstitial or vacancy in nature or 〈100〉 with interstitial nature [19,96,97]. Determination of the Burgers vectors of the dislocation loops is of high importance for irradiated FeCrAl alloys.…”

Section: Dislocation Loop Morphologiesmentioning

confidence: 99%

“…Composition also plays a role on the evolution of dislocation loops under irradiation. In-situ ion irradiations at 320°C have shown that dislocation loop densities are higher in low-Cr (10-12 wt.%) FeCrAl alloys compared to higher Cr variants [97]. Additionally, the ratio of 2 ⁄ 〈111〉 to 〈100〉 dislocation loops also increased with increasing Cr content.…”

Section: Dislocation Loop Morphologiesmentioning

confidence: 99%

Handbook of the Materials Properties of FeCrAl Alloys For Nuclear Power Production Applications

Yamamoto

Snead

Field

et al. 2017

Self Cite

View full text Add to dashboard Cite

alloying additions to increase specific performance factors [13]. Herein, both generations of alloys are grouped together as "ORNL model" alloys unless directly specified.As will be shown later, FeCrAl alloys as a class of materials can demonstrate a wide range of properties depending on their structure, chemistry, and processing. Through combining the databases produced on the GE and ORNL model alloys, as well as other significant commercial and model alloy systems, we have developed the first edition of a comprehensive handbook on FeCrAl alloys for nuclear power applications. This handbook serves to draw from nearly half a century of scientific research but it is known that more studies are forthcoming. As such, the following serves as only the first edition; it is expected that as FeCrAl alloys become a more mature concept, the handbook will be updated and become more substantial.

show abstract

“…To enhance the safety and reliability of nuclear power, different strategies of accident-tolerant fuel (ATF) claddings are proposed [1,5,6]. Compared with the discovery of new materials such as SiC f /SiC [7,8], high entropy alloys [9], and FeCrAl alloys [10,11] to completely replace the Zr-based alloys, making a protective coating on the surface of Zr-based alloys not only significantly improves the high-temperature steam corrosion resistance [12][13][14], but also retains the advantages of the Zr component [15][16][17][18], and thus is considered as a more direct and feasible approach to ATF claddings [19].…”

Section: Introductionmentioning

confidence: 99%

Investigation of Microstructure and Nanoindentation Hardness of C+ & He+ Irradiated Nanocrystal SiC Coatings during Annealing and Corrosion

Liu

et al. 2020

Materials

View full text Add to dashboard Cite

The microstructure and nanoindentation hardness of unirradiated, irradiated, annealed and corroded SiC coatings were characterized. Irradiation of 400 keV C+ and 200 keV He+ with approximately 10 dpa did not cause obvious amorphous transformation to nanocrystal SiC coatings and induced helium bubbles with 2–3 nm dimension distributed uniformly in the SiC matrix. High temperature annealing resulted in the transformation of SiC nanocrystals into columnar crystals in the irradiated region. Line-shaped bubble bands formed at the columnar crystal boundaries and their stacking fault planes and made the formation of microcracks of hundreds of nanometers in length. Meanwhile, some isolated helium bubbles distributed in SiC grains still maintained a size of 2–3 nm, despite annealing at 1200 °C for 5 h. The SiC coating showed excellent corrosion resistance under high-temperature, high-pressure water. The weight of the sample decreased with the increase of corrosion time. The nanoindentation hardness and the elastic modulus increased significantly with C+ and He+ irradiation, while their values decreased with high-temperature annealing. An increase in the annealing temperature led to an increased reduction in the values. Corrosion caused the decrease of nanoindentation hardness and the elastic modulus in the whole test depth range, whether the samples were irradiated or unirradiated.

show abstract

Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys

Cited by 105 publications

References 33 publications

Microstructural examination of neutron, proton and self-ion irradiation damage in a model Fe9Cr alloy

Microstructural examination of neutron, proton and self-ion irradiation damage in a model Fe9Cr alloy

Handbook of the Materials Properties of FeCrAl Alloys For Nuclear Power Production Applications

Investigation of Microstructure and Nanoindentation Hardness of C+ & He+ Irradiated Nanocrystal SiC Coatings during Annealing and Corrosion

Contact Info

Product

Resources

About