TY - JOUR
T1 - The effect of annealing treatment on plasma induced surface blistering and deuterium retention in tungsten
AU - Jiang, Zhenyu
AU - Zhang, Wenjie
AU - Zhang, Honghui
AU - Song, Tianyi
AU - Liu, Yong
AU - Li, Xiaochun
AU - Zhu, Kaigui
N1 - Publisher Copyright:
© 2026 Elsevier B.V.
PY - 2026/8
Y1 - 2026/8
N2 - Hydrogen isotope retention and surface blistering in tungsten, which is regarded as a primary candidate for plasma-facing materials in fusion reactors, can degrade its performance. Annealing treatment is commonly used to relieve residual stress but can also alter the defect density and influence the blistering and deuterium retention behaviors. In this work, the blistering and deuterium retention behaviors of samples unannealed and annealed at 1273 K were investigated. To further clarify the effects of annealing temperature, samples were annealed at 473 K, 673 K, 873 K, 1073 K and 1273 K, respectively, with an unannealed sample and a sample recrystallized at 1973 K serving as references. Nanoindentation tests revealed that the bulk equivalent hardness remains relatively stable until it begins to decrease when the annealing temperature exceeds 1073 K. The maximum shear stress, calculated based on pop-in events, increases with annealing temperature, indicating a decreasing trend in dislocation density. Then the samples were exposed to deuterium plasma with a flux of 5.2 × 1018 D/(m2∙s) and a fluence of 2.0 × 1023 D/m2. Blisters were observed on the surface of samples except the recrystallized one. The average sizes of blisters on each sample are similar, while the densities of blisters exhibit an abrupt drop when annealed at 1073 K. The deuterium retention of samples was analyzed by thermal desorption spectroscopy (TDS) and it decreases as annealing temperature increases. The desorption curves were fitted by TMAP7 code and the fitted results suggest a decreasing concentration of traps. All the results indicate that annealing reduces defect density, thereby mitigating blistering and decreasing deuterium retention.
AB - Hydrogen isotope retention and surface blistering in tungsten, which is regarded as a primary candidate for plasma-facing materials in fusion reactors, can degrade its performance. Annealing treatment is commonly used to relieve residual stress but can also alter the defect density and influence the blistering and deuterium retention behaviors. In this work, the blistering and deuterium retention behaviors of samples unannealed and annealed at 1273 K were investigated. To further clarify the effects of annealing temperature, samples were annealed at 473 K, 673 K, 873 K, 1073 K and 1273 K, respectively, with an unannealed sample and a sample recrystallized at 1973 K serving as references. Nanoindentation tests revealed that the bulk equivalent hardness remains relatively stable until it begins to decrease when the annealing temperature exceeds 1073 K. The maximum shear stress, calculated based on pop-in events, increases with annealing temperature, indicating a decreasing trend in dislocation density. Then the samples were exposed to deuterium plasma with a flux of 5.2 × 1018 D/(m2∙s) and a fluence of 2.0 × 1023 D/m2. Blisters were observed on the surface of samples except the recrystallized one. The average sizes of blisters on each sample are similar, while the densities of blisters exhibit an abrupt drop when annealed at 1073 K. The deuterium retention of samples was analyzed by thermal desorption spectroscopy (TDS) and it decreases as annealing temperature increases. The desorption curves were fitted by TMAP7 code and the fitted results suggest a decreasing concentration of traps. All the results indicate that annealing reduces defect density, thereby mitigating blistering and decreasing deuterium retention.
KW - Annealing
KW - Blister
KW - Deuterium plasma exposure
KW - Deuterium retention
KW - Tungsten
UR - https://www.scopus.com/pages/publications/105040065486
U2 - 10.1016/j.jnucmat.2026.156758
DO - 10.1016/j.jnucmat.2026.156758
M3 - 文章
AN - SCOPUS:105040065486
SN - 0022-3115
VL - 630
JO - Journal of Nuclear Materials
JF - Journal of Nuclear Materials
M1 - 156758
ER -