Microstructure and mechanical properties of neutron-irradiated ultra-fine-grained SUS316L stainless steels and electrodeposited nanocrystalline Ni and Ni–W alloys

Matsuoka, Hiroki; Yamasaki, Tohru; Zheng, Yubin; Mitamura, T.; Terasawa, Michitaka; Fukami, Tatsuki

doi:10.1016/j.msea.2006.02.409

Cited by 27 publications

(12 citation statements)

References 5 publications

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“…So, as expected, the enhanced elimination of point defects at grain boundaries appears to limit the formation of point defect clusters within the grains. This result is consistent with what was observed by Matsuoka et al [7] in neutron irradiated 316-SS samples in which the grain size ranges from 0.1 to 50 mm. Radiation-induced defects clusters were not found for grain size less than 300 nm.…”

Section: Resultssupporting

confidence: 93%

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Comparison of radiation-induced segregation in ultrafine-grained and conventional 316 austenitic stainless steels

Etienne¹,

Radiguet²,

Cunningham³

et al. 2011

Ultramicroscopy

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Section: Resultssupporting

confidence: 93%

“…Several studies have shown that radiation-induced damage is reduced in ultrafinegrained materials [4][5][6][7] due to the high grain boundary density. Since grain boundaries act as point defect sinks, the increase of grain boundary surfaces enhances the annihilation of radiationinduced point defects.…”

Section: Introductionmentioning

confidence: 99%

Comparison of radiation-induced segregation in ultrafine-grained and conventional 316 austenitic stainless steels

Etienne¹,

Radiguet²,

Cunningham³

et al. 2011

Ultramicroscopy

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“…Thus, in order to change the possible effects of point defects fluxes in CW 316SS, an ultrafine grained (UFG) CW 316SS was elaborated by severe plastic deformation. The benefit of UFG 316SS, with grain size about 1-0.1 lm, has already been reported in the literature [27] when exposed to neutron irradiation. In our case, grain size is of the order of 40 nm.…”

Section: Introductionsupporting

confidence: 60%

Experimental atomic scale investigation of irradiation effects in CW 316SS and UFG-CW 316SS

Pareige

Etienne

Radiguet

2009

Journal of Nuclear Materials

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a b s t r a c tMaterials of the core internals of pressurized water reactor (austenitic stainless steels) are subject to neutron irradiation. To understand the ageing mechanisms associated with irradiation and propose life predictions of components or develop new materials, irradiation damage needs to be experimentally investigated. Atomic scale investigation of a neutron-irradiated CW316 SS with the laser pulsed atom probe gives a detailed description of the solute segregation in the austenitic grains. In order to understand the mechanism of solute segregation detected in the neutron-irradiated materials, ion irradiations were performed. These latest irradiations were realized on a CW 316SS as well as on a nanostructured CW 316SS. The study of irradiation effects in a nanograin material allows first, to easily analyse grain boundary segregation and second, to test the behaviour under irradiation of a new nanostructured material. The three aspects of this atomic scale investigation (neutron irradiation effect, model ion irradiation, new nanostructured CW 316 SS) are tackled in this paper.

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“…It is important to note that the grains of the outer UFG zone are equiaxed with predominant high-angle boundaries and large crystallographic disorientation of the lattice. Such a structure should provide an increase in operational properties, in particular, radiation resistance [ 12 , 13 , 14 , 15 ]. The resulting microstructure is in good agreement with the research data [ 6 , 7 , 18 , 19 , 20 , 23 , 24 ].…”

Section: Discussionmentioning

confidence: 99%

“…In particular, a decrease in the proportion of amorphization, radiation hardening and swelling was recorded for nanostructured alloys obtained by equal-channel angular pressing [ 10 , 11 ] the most common method for obtaining UFG structures. The study of radiation resistance of this class of materials is most actively conducted in the nuclear centers of the United States [ 12 ], France [ 13 ], Japan [ 14 ] and other countries [ 15 , 16 ].…”

Section: Introductionmentioning

confidence: 99%

Using of Radial-Shear Rolling to Improve the Structure and Radiation Resistance of Zirconium-Based Alloys

et al. 2020

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An overview of the prospects for the development of nuclear technologies and the conclusion of the relevant requirements for advanced structural materials, their classification and features were performed. In order to obtain a bar with a modified radiation-resistant outer layer, an experiment of radial-shear rolling under the most stringent conditions was carried out. For the same conditions, a FEM-simulation of sequential rolling in eight passes with a total compression of 70.7% (from a diameter of 37 mm to 20 mm) was conducted. For adequate simulation results a new material database for Zr-1%Nb alloy using plastometry investigations was generated. An experimental obtaining of a gradient-modified structure with an ultrafine-grained (UFG) periphery and an elongated rolling texture in the center of the bar was performed.

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Microstructure and mechanical properties of neutron-irradiated ultra-fine-grained SUS316L stainless steels and electrodeposited nanocrystalline Ni and Ni–W alloys

Cited by 27 publications

References 5 publications

Comparison of radiation-induced segregation in ultrafine-grained and conventional 316 austenitic stainless steels

Comparison of radiation-induced segregation in ultrafine-grained and conventional 316 austenitic stainless steels

Experimental atomic scale investigation of irradiation effects in CW 316SS and UFG-CW 316SS

Using of Radial-Shear Rolling to Improve the Structure and Radiation Resistance of Zirconium-Based Alloys

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