Modelling of deuterium diffusion and thermal desorption in tungsten exposed to high-flux deuterium plasma

Abstract

In nuclear fusion, hydrogen isotopes transportation in tungsten has been identified as a crucial process which could affect various key processes in the operation of a fusion device, such as tritium self-sustaining and operational safety. To enhance the understanding of deuterium diffusion and trapping in millimeter-thick tungsten under high-flux plasma conditions, we carried out a systematic simulation investigation on the deuterium plasma exposure and subsequent thermal desorption using the TMAP7 code based on experimental results. The simulation reveals that the defect concentration with a low de-trapping energy , as estimated solely from simulating thermal desorption, is lower than its actual density due to the temperature history of plasma exposure and subsequent cooling processes. Moreover, the growth rate of deuterium mobile concentration exhibits nonmonotonic behavior over time during plasma exposure, which strongly depends on the ratio of defect concentration with a high de-trapping energy to that with a low de-trapping energy. Additionally, we observed that the maximum deuterium diffusion depth does not follow a proportional relationship with the square root of exposure time, which is attributed to the growing number of plasma-induced defects and the reducing injection coefficient with increasing fluences. This work provides valuable insights into the understanding of deuterium transportation in tungsten during high-flux plasma exposure which could contribute to understanding and modelling of hydrogen isotopes recycling in future fusion devices.

Deuterium retention in cyclic transient heat loaded tungsten with increasing cycle numbers

“…(the minimum concentration limit set in TMAP) with depth reaching 35 µm. Sun et al [52] calculated that under conditions of D flux at 4.40 × 10 22 D m −2 s −1 and temperature at 500 K, after 1.5 h of D diffusion, the concentration drastically decreased to a very low level (∼1.60 × 10 −8 at.fr.) by reaching 40 µm.…”

Focus on plasma-facing materials in nuclear fusion reactors

“…While the materials science community has a long history of conducting fundamental and applied research on plasma-materials interactions [6][7][8], this collection focuses on the same type of physics with respect to a nuclear fusion environment. The topics covered are unique to 'Plasma-Facing Materials in Nuclear Fusion Reactors' such as: deuterium and tritium retention in PFCs [9][10][11][12][13]; fundamental processes at the plasma-surface interface [10,[14][15][16][17][18][19][20]; evolution of structure and properties under fusion-reactor-relevant heat loads [21]; material degradation under ion exposure [15,16,19]; material degradation under neutron irradiation [9,21]; material erosion, migration, and deposition [14,15,18,20,22]; plasma fueling [12]; and diagnostics for plasmamaterials interactions [23]. Although the details of the underlying mechanisms that govern the above phenomena remain largely unresolved, the results presented here will drive the emergence of engineering solutions to the amelioration of plasma-facing materials degradation.…”