We have investigated the dissolution, segregation and diffusion of hydrogen (H) in a tungsten (W) grain boundary (GB) using a first-principles method in order to understand the GB trapping mechanism of H. Optimal charge density plays an essential role in such a GB trapping mechanism. Dissolution and segregation of H are directly associated with the optimal charge density, which can be reflected by the H solution and segregation energy sequence for the different interstitial sites. To occupy the optimal-charge-density site, H can be easily trapped by the W GB with the solution and segregation energy of −0.23 eV and −1.11 eV, respectively. Kinetically, such a trapping is easier to realize due to the much lower diffusion barrier of 0.13–0.16 eV from the bulk to the GB in comparison with the segregation energy, suggesting that it is quite difficult for the trapped H to escape out of the GB. However, the GB can hold no more than 2 H atoms because the isosurface of optimal charge density almost disappears with the second H atom in, leading to the conclusion that H2 molecule and thus H bubble cannot form in the W GB. Taking into account the lower vacancy formation energy in the GB as compared with the bulk, we propose that the experimentally observed H bubble formation in the W GB should be via a vacancy trapping mechanism.

https://iopscience.iop.org/article/10.1088/0029-5515/50/2/025016/pdf

Investigating behaviours of hydrogen in a tungsten grain boundary by first principles: from dissolution and diffusion to a trapping mechanism

Although extensive studies on hydrogen (H) retention in tungsten (W) loaded by ion implantation, plasma exposure and gas charging have been done, reported amounts of H retained in W, and their dependences on the incident flux, fluence and temperature are very inconsistent. The present work aims at the clarification of the H retention mechanisms with critical review of reported data and presentation of recent results. H diffusion coefficient for dissolved H in W is large enough to allow for a deep penetration of H into W bulk. Energetic H loading results in H saturation in the subsurface layers. The thickness of the saturated layers could grow under prolonged loading and the H in the layers remain long or hard to remove. Both the growth of the thickness of the saturated layers and the deep penetration of H into the bulk could result in significantly large H retention in W and dissipate the advantage of W as plasma facing wall.

https://iopscience.iop.org/article/10.1088/0031-8949/2014/T159/014044/pdf

Review of hydrogen retention in tungsten

Blister formation and deuterium retention on tungsten exposed to low energy and high flux deuterium plasma

Polycrystalline, annealed tungsten targets were bombarded with 12.3 MeV W4+ ions to various damage levels. Deuterium was implanted by high-flux plasmas in Pilot-PSI (>1024 m−2 s−1) at a surface temperature below 525 K. Deuterium retention has been studied by nuclear reaction analysis and by thermal desorption spectroscopy. We found that deuterium retention is strongly enhanced by the tungsten bombardment and that saturation occurs at a W4+ fluence of about 3 × 1017 m−2. The maximum deuterium concentration in the damaged region was measured to be 1.4 at.%. This is in accordance with other experiments that were carried out at much lower fluxes. We therefore conclude that the saturation behaviour and the maximum retention are not affected by the high fluxes used in our experiments.A simple geometric model is presented that assumes that the saturation solely originates in the tungsten irradiation and that explains it in terms of overlapping saturated volumes. The saturated volume per incident MeV ion amounts to 3 × 104 nm3. From our results, we are able to obtain an approximate value for the average occupation number of the vacancies.

/pdf/saturation-of-deuterium-retention-in-self-damaged-tungsten-5cjwj7p5ia.pdf

Saturation of deuterium retention in self-damaged tungsten exposed to high-flux plasmas

Incident energy dependence of blistering at tungsten irradiated by low energy high flux deuterium plasma beams

Modification of tungsten coated carbon by low energy and high flux deuterium irradiation

Deuterium ion irradiation has been carried out with incident energies of 100 eV and flux of 1 /spl times/ 10/sup 22/ D/sup +/m/sup -2/s/sup -1/ at a temperature in range between 333 K and 1130 K up to a dose of 1 /spl times/ 10/sup 26/ D/sup +/m/sup -2/. Three kinds of tungsten used are pure tungsten made by powder metallurgy tungsten (PM-W), vacuum plasma spray tungsten (VPS-W) and single crystal tungsten (SC-W). Surface morphology before and after the irradiation is observed with a SEM. Retention property of deuterium after the irradiation is also examined with a TDS. Blisters with a diameter of about 0.2-2.0 /spl mu/m are formed on PM-W by deuterium irradiation at 333 K and 673 K but size and shape of those blisters change depending on the irradiation temperatures. However, no blisters are found on PM-W at elevated temperatures between 723 and 1130 K. On the other hand, blisters with a diameter of about 2 /spl mu/m are formed on SCW irradiated at 343 K but no blisters are observed on SCW irradiated at 383, 623 and 1123 K. In the case of VPS-W irradiated at 333, 723 and 1123 K, modification like blister with a diameter of about 0.2 /spl mu/m occurs. TDS measurement also shows that deuterium is not retained in sample, which the blisters are not formed.

T. Kato

Papers

Modification of tungsten coated carbon by low energy and high flux deuterium irradiation

Blister formation and deuterium retention on tungsten irradiated by low energy and high flux deuterium plasma