The mechanism of tungsten (W) blistering under deuterium (D) plasma exposure is still under investigation. To clearly demonstrate the microstructure and nucleation mechanism of blistering on W exposed to D plasma, special recrystallized W disc samples were prepared and electropolished with back-thinned method for TEM observation. D plasma exposure (2.0 × 1026 D m−2, 573 K) brought it numerous intra-granular blisters and protrusions on W with typical orientation dependence. TEM observation revealed the intra-granular blister microstructure that substantial dislocations were generated on the center and edge of blister via severe plastic deformation of blister cap. Dislocation tangles formed by dislocations from blisters were revealed and supposed to be the nucleation of intra-granular blister. By using the g · b = 0 and g · b × u = 0 criterion, 〈0 0 1〉 dislocations with edge component were identified which were generated by 〈1 1 1〉 edge dislocation interaction in dislocation tangles. A {0 0 1}〈0 0 1〉 edge dislocation nucleating and blistering mechanism based on previous works is proposed, and by applying which blistering in recrystallized W could be well explained. Dedicated experiments demonstrated that intra-granular blister formation depends on local dislocation density which validated the mechanism.
In ITER and future tokamaks, recrystallization has been identified as an important issue which may reduce the strength of tungsten plasma-facing component and deteriorate its thermal shock resistance. In this study, isothermal annealing of un-exposed and helium-exposed tungsten was performed to investigate the effect of helium plasma exposure on the recrystallization kinetics. Rolled tungsten samples with a helium plasma fluence of 1.4 × 10 26 m −2 were annealed at temperatures ranging from 1273 K to 1973 K for 1 h. It was found that helium plasma exposure influence both the recrystallization stage (for 1423 K < T < 1573 K) and the grain growth stage (T > 1523 K). The results suggested that the retarding effect is caused by the impediment of high-angle grain boundaries migration by helium clusters and bubbles. Retarded recrystallization was observed at a depth up to a few micrometers beneath the surface. Present results demonstrate that helium plasma exposure plays an important role when qualifying the tungsten divertor performance under heat loading conditions.
To investigate the effect of blistering on hydrogen isotope (HI) retention, a series of deuterium plasma exposures were performed using recrystallized tungsten samples at 500 K with high fluences up to 1.0 × 10 28 ions m −2 in the linear plasma device STEP. An increase of blister density and deuterium retention was observed with increasing plasma fluence. Based on the simulation of the thermal desorption spectra using TMAP, defects with different detrapping energies are found to be located at a depth of tens of microns, which coincides with the depth of the grain boundaries (GBs) close to the surface. The defect characterizations using transmission electron microscopy and positron annihilation Doppler broadening identified the defects as dislocation type and vacancy type, which were created by blistering. It is suggested that these defects can diffuse deep into the material, and the interaction between the diffusion of the defects and GBs causes a peculiar deuterium desorption spectrum over plasma fluences. Additionally, these blister-induced defects are the main source of deuterium retention. Regarding the effect of the blister-induced defects on deuterium retention, a blister-dominated retention mechanism is proposed to describe HI retention in conditions when blistering is severe as in this study. This investigation provides a new insight into the effect of blistering on retention and the modelling of retention in a tokamak edge plasma environment.
This contribution summarized the recent studies of tungsten-based plasma-facing materials in the linear plasma device like the simulator for tokamak edge plasma (STEP), focusing on the examination of newly developed tungsten (W)-based materials and plasma-induced defects in pure W. Pure W, W-V, W-Y 2 O 3 and W-ZrC samples were exposed to a high-flux plasma of ~ 10 21 -10 22 m −2 s −1 with a fluence up to 10 26 m −2 at a surface temperature below 500 K. The investigation of fundamental evolution of plasma-induced defects in pure W indicated a critical role of hydrogen-dislocation interactions. Suppressed surface blistering was observed in all W-based materials, but deuterium desorption behavior and retention were distinct with respect to different materials. The studies showed that the linear plasma device like the STEP was indispensable in the understanding of plasma-material interactions and the qualification of new materials for future fusion reactors.
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