Ion irradiation stability of oxide nano-particles in ODS alloys: TEM studies

Santra, S.; Amirthapandian, S.; Balaji, S.; Panigrahi, B. K.; Serruys, Y.; Robertson, C.

doi:10.1016/j.jnucmat.2019.151861

Cited by 12 publications

(8 citation statements)

References 36 publications

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“…Model alloys ODS-1, 2, and 3 were fabricated using powder metallurgy technique, as described in Refs. 19 and 20 and briefly recalled hereafter. The three selected chemical compositions are ODS-1: 0.3%Y 2 O 3 and Fe (balance).…”

Section: Materials Fabrication Routementioning

confidence: 99%

“…43 For this reason, the mean positron lifetime MLT evolutions, including tau-1, tau-2, and tau-3 contributions, are mainly attributed to open volume variations taking place at particle/matrix interfaces, where cavitation has been directly evidenced. 19,21…”

Section: Journal Of Applied Physicsmentioning

confidence: 99%

“…[10][11][12][13][14][15][16][17][18] The particular issue of ODS particle stability in post-irradiated Fe-0.3Y 2 O 3 , Fe-0.2Ti-0.3Y 2 O 3 , and Fe-14Cr-0.2Ti-0.3Y 2 O 3 model alloys (called ODS-1, 2, and 3 for simplicity) has been addressed elsewhere, thanks to TEM and atom probe investigations. [19][20][21] Complementary SANS (small angle neutron scattering) and PLEPS (pulsed low-energy positron system) measurements are nonetheless presented in Sec. IV and further interpreted in Sec.…”

Section: Introductionmentioning

confidence: 99%

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Indentation response of model oxide dispersion strengthening alloys after ion irradiation up to 700 °C

Robertson

Mathon

Panigrahi

et al. 2022

Journal of Applied Physics

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This paper presents an experimental investigation of irradiation-induced evolutions in three different oxide dispersion strengthening (ODS) alloys. High-dose, dual beam Ni–He ion irradiations are carried out up to 700 °C. The significant dose-dependent changes in the ODS particle size and number density are documented and interpreted in terms of specific point defect transport mechanisms, from small angle neutron scattering, TEM, and pulsed low-energy positron system measurements combined. The corresponding micro-mechanical changes in the alloys are evaluated based on the indentation response, which is, in turn, interpreted in terms of related, sub-grain plasticity mechanisms. The room temperature tests (without dwell time) reveal that the microscale work-hardening rate increases with decreasing the particle number density and pronounced strain localization effect. The elevated temperature tests (up to 600 °C, with dwell time) show that the indentation creep compliance is mostly temperature-independent after irradiation up to 25 dpa at Tirr = 500 °C and markedly temperature-dependent, after irradiation beyond 40 dpa at Tirr = 600 °C. This effect is ascribed to particular creep mechanisms associated with indent-induced plasticity, i.e., high stress and high dislocation density conditions.

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Section: Materials Fabrication Routementioning

confidence: 99%

Section: Journal Of Applied Physicsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

Indentation response of model oxide dispersion strengthening alloys after ion irradiation up to 700 °C

Robertson

Mathon

Panigrahi

et al. 2022

Journal of Applied Physics

Self Cite

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“…These nanoparticles can not only hinder the movement of dislocations, but also absorb the point defects induced by irradiation, enhancing the pore expansion resistance and preventing void swelling of the materials under neutron irradiation. [16][17][18][19][20][21] For these reasons, a higher number density, a finer size, and a better distribution of the oxides are always pursued.Mechanical alloying (MA) is one of the most widely used methods to prepare ODS steels, where Y 2 O 3 powder is often added to the steel powder for long-term high-energy ball-milling. During MA, a large amount of lattice defects (e.g., vacancies and dislocations), along with the supersaturated dissolved Y and O atoms that are decomposed from the Y 2 O 3 powder, can be gradually restored in the powders.…”

mentioning

confidence: 99%

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confidence: 99%

Fabrication of Oxide‐Dispersion‐Strengthened Ferritic Alloys by Mechanical Alloying Using Pre‐Alloyed Powder

Zhang

Zhao

et al. 2022

steel research int.

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Progress toward controllable nuclear fusion operations largely relies on the high performance of advanced nuclear structural materials. [1][2][3][4][5][6] Attributed to the extremely harsh environments where fusion reactors work, the components suffer from both high level of heat flux and high dose of neutron radiation. Oxide-dispersion-strengthened (ODS) steel is one of the most promising candidate structural materials for the next-generation fusion reactors due to its excellent resistance to high-temperature creep, corrosion, and neutron radiation. [7][8][9][10][11][12] ODS steels are mainly based on the reduced activated alloys such as Fe-Cr-W system. These alloys can be divided as reduced activated ferritic martensitic (RAFM) steels with a lower Cr (9-12 wt%) content and reduced activated ferritic (RAF) ones with a higher Cr (>12 wt%) content. Unlike RAFM steels, the RAF steels do not experience a γ-α transformation at elevated temperatures. Therefore, they can be operated at a higher temperature and have a better resistance to radiation swelling. [13][14][15] ODS steels possess a huge number of nano-oxides that are thermally stabilized in the matrix. These nanoparticles can not only hinder the movement of dislocations, but also absorb the point defects induced by irradiation, enhancing the pore expansion resistance and preventing void swelling of the materials under neutron irradiation. [16][17][18][19][20][21] For these reasons, a higher number density, a finer size, and a better distribution of the oxides are always pursued.Mechanical alloying (MA) is one of the most widely used methods to prepare ODS steels, where Y 2 O 3 powder is often added to the steel powder for long-term high-energy ball-milling. During MA, a large amount of lattice defects (e.g., vacancies and dislocations), along with the supersaturated dissolved Y and O atoms that are decomposed from the Y 2 O 3 powder, can be gradually restored in the powders. Then, powder metallurgy is conducted to this non-equilibrious MA powders to build the components with proper sizes and shapes. [22] In this step, one of the following densification strategies, that is, hot extrusion (HE), hot isostatic pressing (HIP), or spark plasma sintering (SPS), is usually carried out. [23][24][25] Assisted by the subsequent heat treatment, a huge number of Y 2 O 3 nanoparticles win the chance to simultaneously precipitate from the supersaturated matrix. In the earlier procedures, MA is definitely a timeconsuming stage. It takes an extremely long period, usually longer than 50 h, for the complete decomposition and dissolution of Y 2 O 3 through ball-milling since it has the highest bonding energy among most of the common oxides. [24][25][26] If MA process can be accelerated by a faster decomposition of the oxide

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Development of Advanced Nuclear Structural Materials for Sustainable Energy Development

Srivastava¹,

Kapoor²,

Amarendra

2022

J Indian Inst Sci

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Ion irradiation stability of oxide nano-particles in ODS alloys: TEM studies

Cited by 12 publications

References 36 publications

Indentation response of model oxide dispersion strengthening alloys after ion irradiation up to 700 °C

Indentation response of model oxide dispersion strengthening alloys after ion irradiation up to 700 °C

Fabrication of Oxide‐Dispersion‐Strengthened Ferritic Alloys by Mechanical Alloying Using Pre‐Alloyed Powder

Development of Advanced Nuclear Structural Materials for Sustainable Energy Development

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