Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

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Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM. / Tunes, M. A.; Harrison, R. W.; Greaves, G. et al.
In: Journal of nuclear materials, Vol. 493.2017, No. September, 01.09.2017, p. 230-238.

Research output: Contribution to journal › Article › Research › peer-review

Harvard

Tunes, MA, Harrison, RW, Greaves, G, Hinks, JA & Donnelly, SE 2017, 'Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM', Journal of nuclear materials, vol. 493.2017, no. September, pp. 230-238. https://doi.org/10.1016/j.jnucmat.2017.06.012

APA

Tunes, M. A., Harrison, R. W., Greaves, G., Hinks, J. A., & Donnelly, S. E. (2017). Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM. Journal of nuclear materials, 493.2017(September), 230-238. https://doi.org/10.1016/j.jnucmat.2017.06.012

Vancouver

Tunes MA, Harrison RW, Greaves G, Hinks JA, Donnelly SE. Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM. Journal of nuclear materials. 2017 Sept 1;493.2017(September):230-238. Epub 2017 Jun 13. doi: 10.1016/j.jnucmat.2017.06.012

Author

Tunes, M. A. ; Harrison, R. W. ; Greaves, G. et al. / Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM. In: Journal of nuclear materials. 2017 ; Vol. 493.2017, No. September. pp. 230-238.

Bibtex - Download

@article{c8141206efdc48a48ebcb1879d04492d,
title = "Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM",
abstract = "Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7×1017ions·cm−2 in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7×1017ions·cm−2 in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.",
keywords = "Bubble lattices, Hydrides, In situ ion irradiation, Radiation damage, Zirconium alloys",
author = "Tunes, {M. A.} and Harrison, {R. W.} and G. Greaves and Hinks, {J. A.} and Donnelly, {S. E.}",
year = "2017",
month = sep,
day = "1",
doi = "10.1016/j.jnucmat.2017.06.012",
language = "English",
volume = "493.2017",
pages = "230--238",
journal = "Journal of nuclear materials",
issn = "0022-3115",
publisher = "Elsevier",
number = "September",

}

RIS (suitable for import to EndNote) - Download

TY - JOUR

T1 - Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

AU - Tunes, M. A.

AU - Harrison, R. W.

AU - Greaves, G.

AU - Hinks, J. A.

AU - Donnelly, S. E.

PY - 2017/9/1

Y1 - 2017/9/1

N2 - Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7×1017ions·cm−2 in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7×1017ions·cm−2 in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.

AB - Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7×1017ions·cm−2 in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7×1017ions·cm−2 in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.

KW - Bubble lattices

KW - Hydrides

KW - In situ ion irradiation

KW - Radiation damage

KW - Zirconium alloys

UR - http://www.scopus.com/inward/record.url?scp=85021097371&partnerID=8YFLogxK

U2 - 10.1016/j.jnucmat.2017.06.012

DO - 10.1016/j.jnucmat.2017.06.012

M3 - Article

AN - SCOPUS:85021097371

VL - 493.2017

SP - 230

EP - 238

JO - Journal of nuclear materials

JF - Journal of nuclear materials

SN - 0022-3115

IS - September

ER -

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