TY - JOUR
T1 - Deciphering the role of second phase precipitates on early-stage surface morphology development of dispersion-strengthened W alloys under low energy He irradiation
AU - Lang, Eric
AU - Madden, Nathan
AU - Smith, Charles
AU - Krogstad, Jessica
AU - Allain, J. P.
N1 - Funding Information:
The characterization was carried out in part in the Frederick Seitz Materials Research Laboratory Central Research Facilities, University of Illinois. This work is supported by US DOE contract no. DE-SC0014267 .
Publisher Copyright:
© 2019
PY - 2019/5
Y1 - 2019/5
N2 - Tungsten is the material of choice for plasma-facing components in the divertor region of future plasma-burning tokamak fusion reactors. However, under low energy helium irradiation at elevated temperatures, significant surface morphology changes are expected, including pores, blisters, and fuzz. Dispersion-strengthened tungsten materials with small additions of transition metal carbide dispersoids have been proposed as an alternative to pure tungsten, as they have shown enhanced thermomechanical properties and possible radiation damage tolerance. However, their response to low energy helium irradiation has yet to be fully elucidated. In this work, dispersion-strengthened tungsten alloys containing 1–10 wt.% tantalum carbide, titanium carbide, or zirconium carbide were exposed to 250 eV helium ions at 600 and 800 °C to a 1 × 10 20 cm −2 fluence to understand the early-stage irradiation response to helium bombardment. At 600 °C, nanostructuring is only observed on titanium carbide particles. As the temperature is raised to 800 °C, pores and ripples were developed on tungsten grains for all samples; fiber-form structures and isolated tendril growth is observed only on titanium carbide particles. Minimal surface morphology changes were observed on tantalum carbide and zirconium carbide particles. X-ray photoelectron spectroscopy of the alloyed specimens post-irradiation at 800 °C indicates the formation of zirconium and titanium oxides on the surface. Potential thermodynamic, sputtering, and composition-based formation mechanisms behind the novel nanostructuring and chemistry changes of the complex materials are discussed.
AB - Tungsten is the material of choice for plasma-facing components in the divertor region of future plasma-burning tokamak fusion reactors. However, under low energy helium irradiation at elevated temperatures, significant surface morphology changes are expected, including pores, blisters, and fuzz. Dispersion-strengthened tungsten materials with small additions of transition metal carbide dispersoids have been proposed as an alternative to pure tungsten, as they have shown enhanced thermomechanical properties and possible radiation damage tolerance. However, their response to low energy helium irradiation has yet to be fully elucidated. In this work, dispersion-strengthened tungsten alloys containing 1–10 wt.% tantalum carbide, titanium carbide, or zirconium carbide were exposed to 250 eV helium ions at 600 and 800 °C to a 1 × 10 20 cm −2 fluence to understand the early-stage irradiation response to helium bombardment. At 600 °C, nanostructuring is only observed on titanium carbide particles. As the temperature is raised to 800 °C, pores and ripples were developed on tungsten grains for all samples; fiber-form structures and isolated tendril growth is observed only on titanium carbide particles. Minimal surface morphology changes were observed on tantalum carbide and zirconium carbide particles. X-ray photoelectron spectroscopy of the alloyed specimens post-irradiation at 800 °C indicates the formation of zirconium and titanium oxides on the surface. Potential thermodynamic, sputtering, and composition-based formation mechanisms behind the novel nanostructuring and chemistry changes of the complex materials are discussed.
KW - Dispersion-strengthened tungsten
KW - Helium damage
KW - Tungsten
KW - Tungsten fuzz
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U2 - 10.1016/j.nme.2019.01.016
DO - 10.1016/j.nme.2019.01.016
M3 - Article
AN - SCOPUS:85061435827
SN - 2352-1791
VL - 19
SP - 47
EP - 54
JO - Nuclear Materials and Energy
JF - Nuclear Materials and Energy
ER -