In-pile creep rupture properties of ODS ferritic steel claddings

Kaito, Takeji; Ohtsuka, Satoshi; Inoue, Masakí; Asayama, Tai; Uwaba, Tomoyuki; Mizuta, S; Ukai, Shigeharu; Furukawa, Tomohiro; Ito, Chikako; Kagota, Eiichi; Kitamura, Ryoichi; Aoyama, Takafumi; Inoue, Toshihiko

doi:10.1016/j.jnucmat.2008.12.117

Cited by 36 publications

(14 citation statements)

References 10 publications

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“…One possible explanation for enhanced irradiation creep at temperatures of about 600°C could be that at this temperature thermal creep already becomes the important mechanism for this material. This behavior which was observed also for PM2000 and TiAl [61,62] would explain also why no irradiation creep effects were detected for in-pile creep of advanced ODS materials at temperatures of 700°C and higher [63]. These results seem to demonstrate that the presence of dispersoids does not have a significant influence on the irradiation creep behavior.…”

Section: Creep Under Irradiationsupporting

confidence: 62%

Damage assessment in structural metallic materials for advanced nuclear plants

Hoffelner

2010

J Mater Sci

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Future advanced nuclear plants are considered to operate as cogeneration plants for electricity and heat. Metals and alloys will be the main portion of structural materials employed (including fuel claddings). Due to the operating conditions these materials are exposed to damaging conditions like creep, fatigue, irradiation and its combinations. The paper uses the most important alloys: ferritic-martensitic steels, superalloys, oxide dispersion strengthened steels and to some extent titanium aluminides to discuss its responses to these exposure conditions. Extrapolation of stress rupture data, creep strain, swelling, irradiation creep and creep-fatigue interactions are considered. Although the stress rupture-and the creep behavior seem to meet expectations, the long design lives of 60 years are really challenging for extrapolations and particularly questions like negligible creep or occurrence of diffusion creep need special attention. Ferritic matrices (including oxide dispersion strengthened (ODS), steels) have better irradiation swelling behavior than austenites. Presence and size of dispersoids having a strong influence on high-temperature strength bring only insignificant improvements in irradiation creep. A strain-range-separation based approach for creep-fatigue interactions is presented which allows a real prediction of creep-fatigue lives. An assessment of capabilities and limitations of advanced materials modeling tools with respect to damage development is given.

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Section: Creep Under Irradiationsupporting

confidence: 62%

Damage assessment in structural metallic materials for advanced nuclear plants

Hoffelner

2010

J Mater Sci

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“…It can be noted that the excess oxygen content (E x ), calculated from equation proposed in Ref. [11] does not exceed 0.1 %.…”

Section: Methodsmentioning

confidence: 88%

“…This observation can be explained by the interactions of the grain boundaries in ferritic matrix with the nanoparticles present in the material [19]. Especially, Y-Ti-O nanoparticles are expected to suppress grain growth imparting a very good thermal stability of the ODS ferritic steels [11,12].…”

Section: Coarsening Mechanism Of Oxide Nanoparticlesmentioning

confidence: 99%

Microstructural changes upon annealing in ODS-strengthened ultrafine grained ferritic steel

et al. 2013

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In this study, the stability of grain size and oxide nanoparticles in the ODS steel upon annealing at high temperature (650-1350°C) has been evaluated. The ODS Fe-Cr-W-Ti-Y 2 O 3 steel has been manufactured by powder metallurgy, consolidated by hot isostatic pressing and processed by hydrostatic extrusion. Such a processing brings about ultrafine grain structure reinforced with oxide nanoparticles (few nm in diameter) and results in superior mechanical properties. The stability of nano-oxides has been analyzed by small angle X-ray scattering together with transmission electron microscopy. The results obtained revealed excellent thermal stability of ultrafine grained ODS ferritic steel, which was attributed to the resistance of oxides against coarsening.

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“…• C higher, the rupture life of 9Cr-ODS steel is shortened by 1/30 [20]. Then, the high-temperature irradiation in the second irradiation test might provide the 9Cr-ODS steel cladding tubes with considerable creep damage, thus playing a role in the fuel pin rupture.…”

Section: Discussionmentioning

confidence: 99%

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Ohtsuka

Kaito

Yano

et al. 2013

Journal of Nuclear Science and Technology

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Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60 to demonstrate the in-reactor performance of 9Cr-ODS steel for use as fuel cladding tubes. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors, such as microstructure instability and fuel pin rupture, occurred. Investigations of the cause of these peculiar irradiation behaviors were carried out. The detection sensitivity in an ultrasonic inspection test was shown to be low for the metallic Cr and metallic Fe inclusions. The peculiar microstructure change reappeared with high-temperature thermal-aging of the 9Cr-ODS steel containing metallic Cr inclusions. The strength and ductility of the defective part containing metallic Cr inclusions were appreciably lower than those of a standard part without the inclusions. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change in 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.

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In-pile creep rupture properties of ODS ferritic steel claddings

Cited by 36 publications

References 10 publications

Damage assessment in structural metallic materials for advanced nuclear plants

Damage assessment in structural metallic materials for advanced nuclear plants

Microstructural changes upon annealing in ODS-strengthened ultrafine grained ferritic steel

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

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