Stability of oxide particles under electron irradiation in a 9Cr ODS steel at 400 °C

Li, Feng; Abe, Hiroaki; Ishizaki, Takahiro; Li, Yanfen; Nagasaka, Tetsuya; Muroga, T.; Nagase, Takeshi; Yasuda, Hidehiro

doi:10.1016/j.jnucmat.2014.09.011

Cited by 18 publications

(6 citation statements)

References 11 publications

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“…Because oxide particles act as point-defect sinks, any changes to the nano-dispersoid microstructure will affect the overall radiation tolerance of the alloy in the entire operating temperature range for FW/B structures. The oxide particle stability, recently reviewed in [216], has been extensively researched using neutrons [194,196,199,204,205,[217][218][219][220][221][222][223][224][225][226][227][228][229][230], ions [219, and electron irradiations [261,262] in a variety of ODS steels. Extensive ion irradiation literature exists on a wide range of ODS steels where nano-dispersoids are reported to be stable over the irradiation temperature and dose ranges of RT-835 • C and ion doses from ∼2 to 200 dpa [205, 233, 234, 237, 240, 243, 247-249, 251, 259, 263].…”

Section: Potential Causes Of Lthementioning

confidence: 99%

“…Extensive ion irradiation literature exists on a wide range of ODS steels where nano-dispersoids are reported to be stable over the irradiation temperature and dose ranges of RT-835 • C and ion doses from ∼2 to 200 dpa [205, 233, 234, 237, 240, 243, 247-249, 251, 259, 263]. However, substantial literature also exists over a similar ion irradiation temperature and dose range, where deterioration of the oxide particles is reported [205,219,222,226,231,235,238,239,241,242,245,250,257,261,262,264,265]. The majority of these reports typically show changes in the size or number density of the nano-dispersoids, while irradiations in the temperature range of RT-400 • C have also shown radiation-induced amorphisation (RIA) of the oxide particles in many ODS steels [196,229,235,243,251,261,264].…”

Section: Potential Causes Of Lthementioning

confidence: 99%

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Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

et al. 2022

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Reduced activation ferritic-martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. Performance of these steels will deteriorate in-service due to neutron damage and transmutation-induced gases like helium/hydrogen at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the major irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (i) low temperature hardening-embrittlement (LTHE) - including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile to brittle transition temperature and fracture toughness, (ii) inter-mediate temperature cavity swelling, (iii) the effect of helium on LTHE and cavity swelling, (iv) irradiation creep and (v) tritium management issues are reviewed. The potential causes of LTHE are discussed that highlights need for advanced characterization techniques. Mechanical properties including tensile/Charpy impact of RAFM and ODS steels are compared to show that current generation of ODS steels also suffer from LTHE, and they can show irradiation hardening up to high temperatures, ~400-500 °C. To minimize this, future ODS steel development for FW/B specific application should target materials with lower Cr concentration (to minimize α’) and minimize other elements that could form embrittling phases under irradiation. RAFM steel designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand synergistic effects of helium on thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex including stresses due to magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multi-scale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility

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Section: Potential Causes Of Lthementioning

confidence: 99%

Section: Potential Causes Of Lthementioning

confidence: 99%

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

et al. 2022

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“…The favorable binding of vacancy cluster-Cr-O complex defects [54] further supports the tendency for ballistically-ejected species to remain in solution (rather than reforming into clusters) at lower temperatures. Yet, there are also numerous observations of recoil resolution of oxide NPs at irradiation temperatures in excess of 400°C [8], [14]- [18], [20], [27], [29], [32], [35], [41], [44], [55], suggesting that temperature may not be the only factor determining whether atoms remain in solution.…”

Section: Ballistic Dissolutionmentioning

confidence: 99%

A review of the irradiation evolution of dispersed oxide nanoparticles in the b.c.c. Fe-Cr system: Current understanding and future directions

Wharry

Swenson

Yano

2017

Journal of Nuclear Materials

104

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Thus far, a number of studies have investigated the irradiation evolution of oxide nanoparticles in b.c.c. Fe-Cr based oxide dispersion strengthened (ODS) alloys. But given the inconsistent experimental conditions, results have been widely variable and inconclusive. Crystal structure and chemistry changes differ from experiment to experiment, and the total nanoparticle volume fraction has been observed to both increase and decrease. Furthermore, there has not yet been a comprehensive review of the archival literature. In this paper, we summarize the existing studies on nanoparticle irradiation evolution. We note significant observations with respect to oxide nanoparticle crystallinity, composition, size, and number density. We discuss three possible contributing mechanisms for nanoparticle evolution: ballistic dissolution, Ostwald ripening, and irradiation-enhanced diffusion. Finally, we propose future directions to achieve a more comprehensive understanding of irradiation effects on oxide nanoparticles in ODS alloys.

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“…If this phenomenon is normally expected at low temperature, some authors also reported dissolution at higher temperature. Li et al [5] observed nanoparticle shrinking under electron irradiation at 400 • C in a Fe-9Cr ODS. Swenson and Wharry [6] observed the dissolution of the nano-oxides in Fe-9Cr ODS steels after neutron irradiation up to 3 dpa at the even higher temperature of 500 • C. -Ostwald ripening under irradiation is similar to the well-known thermal process where small particles shrink to the benefit of large ones to minimize interfacial energy; the main difference is the irradiation cascades, which help in increasing interfacial solute concentration and enhance solute transport [3].…”

Section: Introductionmentioning

confidence: 99%

Nano-Structured Materials under Irradiation: Oxide Dispersion-Strengthened Steels

et al. 2021

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Oxide dispersion-strengthened materials are reinforced by a (Y, Ti, O) nano-oxide dispersion and thus can be considered as nanostructured materials. In this alloy, most of the nanoprecipitates are (Y, Ti, O) nano-oxides exhibiting a Y2Ti2O7 pyrochlore-like structure. However, the lattice structure of the smallest oxides is difficult to determine, but it is likely to be close to the atomic structure of the host matrix. Designed to serve in extreme environments—i.e., a nuclear power plant—the challenge for ODS steels is to preserve the nano-oxide dispersion under irradiation in order to maintain the excellent creep properties of the alloy in the reactor. Under irradiation, the nano-oxides exhibit different behaviour as a function of the temperature. At low temperature, the nano-oxides tend to dissolve owing to the frequent ballistic ejection of the solute atoms. At medium temperature, the thermal diffusion balances the ballistic dissolution, and the nano-oxides display an apparent stability. At high temperature, the nano-oxides start to coarsen, resulting in an increase in their size and a decrease in their number density. If the small nano-oxides coarsen through a radiation-enhanced Ostwald ripening mechanism, some large oxides disappear to the benefit of the small ones through a radiation-induced inverse Ostwald ripening. In conclusion, it is suggested that, under irradiation, the nano-oxide dispersion prevails over dislocations, grain boundaries and free surfaces to remove the point defects created by irradiation.

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Stability of oxide particles under electron irradiation in a 9Cr ODS steel at 400 °C

Cited by 18 publications

References 11 publications

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

A review of the irradiation evolution of dispersed oxide nanoparticles in the b.c.c. Fe-Cr system: Current understanding and future directions

Nano-Structured Materials under Irradiation: Oxide Dispersion-Strengthened Steels

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