Review of Irradiation Damage Behavior of Tungsten Exposed to Plasma in Nuclear Fusion

doi:10.11900/0412.1961.2023.00078

Acta Metall Sin

2023, Vol. 59

Issue (8): 986-1000 DOI: 10.11900/0412.1961.2023.00078

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Review of Irradiation Damage Behavior of Tungsten Exposed to Plasma in Nuclear Fusion

LIU Wei¹(

), CHEN Wanqi², MA Menghan¹, LI Kailun³

¹School of Materials Science and Engineering, Tsinghua University, Beijing 100084, China
²CNNC China Nuclear Power Engineering Co., Ltd., Beijing 100840, China
³Institute of Engineering Thermophysics, Chinese Academy of Sciences, Beijing 100190, China

Cite this article:

LIU Wei, CHEN Wanqi, MA Menghan, LI Kailun. Review of Irradiation Damage Behavior of Tungsten Exposed to Plasma in Nuclear Fusion. Acta Metall Sin, 2023, 59(8): 986-1000.

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Abstract

Tungsten is the most promising candidate as plasma facing material in nuclear fusion reactors because of its high melting point, high thermal conductivity, low sputtering rate, and low tritium retention. However, when exposed to low-energy high flux plasma, tungsten undergoes micro/nanoscale damage, such as surface blistering and surface nanostructure, on its surface. These damage structures can degrade thermal and mechanical properties, thereby adversely affecting the reservice performance of tungsten. In this paper, the current research status of the damage behavior of tungsten when exposed to H/D plasma was focused. The research progress of the mechanisms of surface blistering nucleation and growth, as well as the effects of irradiation defects on thermal conductivity, mechanics, and service performance was summarized. These data can provide a theoretical basis for optimizing the microstructure of tungsten materials, thus improving its service performance and extending its service life.

Key words: nuclear fusion tungsten plasma irradiation damage service performance

Received: 27 February 2023

ZTFLH:

TG146.4

Fund: National Natural Science Foundation of China(12105314)

Corresponding Authors: LIU Wei, professor, Tel:13910677301, E-mail: liuw@mail.tsinghua.edu.cn

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Fig.1 Surface (a, c, e) and crosss-ection (b, d, f) morphologies of surface blistering of D plasma exposed W with surface normal directions close to [111] (a, b), [110] (c, d), and [001] (e, f) directions, respectively^[16] (Red arrows in Fig.1d show two blisters caused by the gas pressure inside the cavities beneath the surface; A blue arrow in Fig.1d shows a large cavity along the grain boundary; Blue arrows in Fig.1f show that there were also cavities beneath the surface, but these cavities did not induce any obvious blisters on the surface at all)

Fig.2 Surface morphologies of W samples exposed to D plasma (943 K) at irradiation doses of 1.55 × 10²⁶ m^-2 (a), 4.21 × 10²⁶ m^-2 (b), and 7.05 × 10²⁶ m^-2 (c); and surface bubble size and density distribution statistics (d)^[21]

Fig.3 Surface blistering morphologies of surface bubbles on [111] planeswith different grain orientations (a-c)^[16] (The specific grain orientations are marked in the [110] direction)

Fig.4 Surface blistering model based on plastic deformation mechanism^[16]

Fig.5 Surface bubbles and nanobubbles of W after D plasma irradiation (38 eV, 10²⁴ m^-2·s^-1, 423 K, 7 × 10²⁶ m^-2)^[27] (a, b) surface bubbles viewed vertically (a) and tilted at 45° (b) (c, d) nanobubbles in SEM conventional mode (c) and Inlens mode (d) (Arrows in Fig.5d show the morphologies of broken nanobubbles)

Fig.6 TEM images of recrystallized W exposed to D plasma^[34]
(a) original morphology of recrystallized W
(b) intra-granular bubbles (Arrows show bubbles)
(c-f) bubbles and surrounding dislocations

Fig.7 Schematics of loop punching mechanism^[28]
(a) H atoms accumulate at a nucleation site
(b) H bubble grows by continuous absorption of H atoms and outward release of dislocation loops ( b₁—Burgers vector 1, b₂—Burgers vector 2, p—gas pressure, μ—shear modulus)

Fig.8 Typical morphologies of dislocations loops distributed around the blister in recrystallized W after exposed to H plasma^[41]
(a) four dislocation loop arrays distributed near the intra-granular H blister
(b) enlarge area 1 in Fig.8a, prismatic dislocation loops and “coffee-bean” loops distributed along [11

1 ¯

] direction (Rectangularareas show the same group dislocations which observed under different g vector)

Fig.9 Morphologies of shear dislocation loops arrayed at the tip of the intra-granular blisters in recrystallized W after exposed to H plasma (Rectangular areas show the same group dislocations which observed under different g vector)^[41]
(a) g₁ = 200 (b) g₂ = 01

1 ¯

1 ¯

(d) g₄ = 020 (e) g₅ = 200 (f) g₇ =

1 ¯ 1 ¯

Fig.10 TEM morphologies of intra-granular blisters of recrystallized tungsten, H blisters respectively located in(100) plane (area 1) (a), (001) plane (area 2) (b), and (010) plane (area 3) (c); the locations of areas 1-3 in W near-surface (d)^[59]

Fig.11 H blister nucleation process at <100> edge dislocation^[59]
(a) time t = 1 ns, the edge dislocation with a <100> Burgers vector is not filled by H atoms (The inset enlarged figure shows the dislocation core structure)
(b) t = 5 ns, the dislocation core opens towards the <011> direction and accommodates more H atoms. The H-rich phase-transformation region is also formed at this stage (Inset shows the high magnified image of rectangular area)
(c) t = 10 ns, the dislocation core extends further
(d) t = 15 ns, with increased blister size, the phase-transformation region increases. The inset enlarged figure shows the crystal structure of the phase-transition region, which is fcc W with H filling all of its octahedral sites
(e) t = 20 ns, the H dislocation diffusion path is visible
(f) t = 25 ns, the final configuration is apparent; the phase-transition region grows continuously, following the blister tip (Inset shows the high magnified image of rectangular area)

Fig.12 Thermal diffusivity versus irradiation dose and temperature for four samples (Thermal diffusivity of unexposed W > 6.5 × 10^-5 m²/s; HT—high temperature; LT—low temperature; HD—high dose sample; LD—low dose; LT: about 450 K; HT: about 650 K; LD: about 5 × 10²⁵ m^-2 (70 s); HD: about 1 × 10²⁷ m^-2 (1400 s))^[70]

Fig.13 Indentation hardness of W at different irradiation temperatures^[44]

Fig.14 Morphologies of W after different exposures^[84]
(a) only steady state plasma exposure
(b) only transient state high heat loads laser
(c) consecutive exposure, laser induced transient heat loads on pre-exposed plasma
(d) simultaneous exposure

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