Speaker
Description
Tritium (T) transport through the first wall into the coolant is a major concern in fusion reactor studies. When irradiated by plasmas, hydrogen permeation flux through in-vessel components would be significantly higher than that of gas-exposure cases. To support reactor design studies, low energy plasma-driven hydrogen isotope permeation through the first wall has been extensively investigated. This review will introduce our recent research progress in three relevant topics:
(1) Surface damage effects on hydrogen isotope permeation [1,2]. In-situ measurements of low energy deuterium (D) through helium pre-damaged tungsten (W) has been done. With the increases of helium (He) pre-irradiation fluence, the D permeation flux was found to reduce effectively.
(2) The role of W-structural materials interface [3]. The transport behavior of D in W-Cu joining sample was explored using gas driven permeation and thermal desorption spectroscopy. A large number of D atoms were found to be trapped by dislocations and impurities at the intermediate layer.
(3) Isotope effects [4,5]. The co-permeation experiments of H and D isotopes through∼mm thickness materials. The H/D ratio of the steady state permeation fluxes was found to be close to the classical theoretical. Hydrogen isotope exchange is an effective method for T removal in fusion reactor materials.
References
[1] Lu Wang et al., Nucl. Fusion 62 (2022) 086006.
[2] Xue-Chun Li et al., Nucl. Fusion 62 (2022) 064001.
[3] Xue-Chun Li et al., Nucl. Mater. Energy 38 (2024) 101598.
[4] Cai-Bin Liu et al., Nucl. Fusion 62 (2022) 126017.
[5] Fei Sun et al., Nucl. Fusion 64 (2024) 046011.
