Void swelling in high dose ion-irradiated reduced activation ferritic–martensitic steels

Wang, Xu; Monterrosa, Anthony M.; Zhang, Feifei; Huang, Hung Chang; Yan, Qingzhi; Jiao, Zhijie; Was, Gary S.; Wang, Lumin

doi:10.1016/j.jnucmat.2015.03.050

Cited by 53 publications

(16 citation statements)

References 32 publications

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“…1a, the much smaller voids in NiFe and NiCoFeCr represent much higher swelling resistance than in nickel and NiCo. The swelling values represent the local volume changes due to the void formation, which is calculated by a common equation23. While nickel shows a significant overall swelling of ∼1.8% and NiCo has a lower void swelling of about 0.42%, NiFe and NiCoFeCr exhibit substantially lower void swelling with similar values of ∼0.02%, more than two orders of magnitude lower than in nickel.…”

Section: Resultsmentioning

confidence: 99%

Enhancing radiation tolerance by controlling defect mobility and migration pathways in multicomponent single-phase alloys

et al. 2016

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A grand challenge in material science is to understand the correlation between intrinsic properties and defect dynamics. Radiation tolerant materials are in great demand for safe operation and advancement of nuclear and aerospace systems. Unlike traditional approaches that rely on microstructural and nanoscale features to mitigate radiation damage, this study demonstrates enhancement of radiation tolerance with the suppression of void formation by two orders magnitude at elevated temperatures in equiatomic single-phase concentrated solid solution alloys, and more importantly, reveals its controlling mechanism through a detailed analysis of the depth distribution of defect clusters and an atomistic computer simulation. The enhanced swelling resistance is attributed to the tailored interstitial defect cluster motion in the alloys from a long-range one-dimensional mode to a short-range three-dimensional mode, which leads to enhanced point defect recombination. The results suggest design criteria for next generation radiation tolerant structural alloys.

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Section: Resultsmentioning

confidence: 99%

Enhancing radiation tolerance by controlling defect mobility and migration pathways in multicomponent single-phase alloys

et al. 2016

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“…In order to avoid the influence of injected interstitial self-ions, the depth range of 450 nm ~750 nm from the surface with a dose of 188 dpa was used for analyzing void swelling in this study. Helium ions were pre-implanted with the following energies, 80, 140, 220, 310, and 420 keV, respectively, to obtain a relatively uniform concentration through the depth range of 300~1000 nm from the sample surface10. Bright-field (BF) TEM images of typical area in fine grains and coarse grains of COS-2 after irradiation are shown in Fig.…”

Section: Resultsmentioning

confidence: 99%

“…A 9Cr-F/M steel with pre-implanted helium was also irradiated in the same condition as a benchmark steel. The swelling results of the 9Cr-F/M steel have been published in another paper10, and the swelling value will be compared in this study. The objective of this study is to further understand the relationship between swelling resistance and oxide stability in the irradiated ODS steel.…”

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confidence: 99%

Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation

Wang

et al. 2017

Sci Rep

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The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density.

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“…M 2 X have been characterized as either coherent [47,48] or semi-coherent [15,35] and are treated here as a variable-biased sink using formalism described by Brailsford and Bullough [45,46,49]:…”

Section: Precipitatesmentioning

confidence: 99%

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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Void swelling in high dose ion-irradiated reduced activation ferritic–martensitic steels

Cited by 53 publications

References 32 publications

Enhancing radiation tolerance by controlling defect mobility and migration pathways in multicomponent single-phase alloys

Enhancing radiation tolerance by controlling defect mobility and migration pathways in multicomponent single-phase alloys

Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation

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

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