Speaker
Description
A critical challenge in the fusion materials science lies in understanding the influence of radiation damage on hydrogen isotopes (HIs) retention, particularly concerning safety and tritium self-sufficiency. Reduced activation ferritic/martensitic (RAFM) steels are leading candidates for use as structural material in the first wall such as the breeding blanket modules. In deuterium-tritium (D-T) plasmas, structural materials will be exposed to 14 MeV neutrons, leading to atomic displacements and helium (He) production as a result of nuclear reactions. The resulting defects and He bubbles act as trapping sites for HIs, significantly influencing their transport and retention behavior.
This study investigates deuterium (D) retention in displacement-damaged and He-implanted EUROFER97. The objective is to characterize He-related trap sites for D and study their thermal evolution. EUROFER97 samples are irradiated with 20 MeV tungsten ions at temperatures ranging from 290 to 770 K, inducing a primary peak damage dose of 0.6 dpa. Selected samples are additionally implanted by 0.5 MeV He ions to a peak concentration of up to 0.6 at.% to form He bubbles. These samples are then exposed to a low-temperature and low-energy D plasma or to D$_2$ gas to fill the radiation-induced defects with D. Nuclear Reaction Analysis ($^3$He-NRA) is applied to determine D depth profiles. Thermal Desorption Spectroscopy (TDS) is employed to investigate D desorption.
Transmission electron microscopy (TEM) confirmed the formation of nanoscale He bubbles, which were found to significantly enhance local D concentration. Although the density of He-related trapping sites decreases with increasing annealing temperature, they are not completely recovered even after annealing at 720 K, unlike displacement-damage without the presence of He that recovers at 620 K. Irradiation at reactor-operating temperatures leads to lower D retention compared to annealing at the same temperatures after irradiation at 290 K. D is de-trapped from He-related defects and desorbed from EUROFER97 at about 520 K. Notably, TDS spectra differ significantly between plasma-exposed and gas-exposed samples, with the latter showing two additional high-temperature desorption peaks.
Finally, rate-equation simulations employing the TESSIM-X code were conducted to determine trapping and de-trapping energies for both He-related and intrinsic defects, enabling the interpretation of the experimental TDS spectra. This work significantly improves the ability to predict the tritium loss in a DEMO first wall by providing experimentally determined defect parameters that account the impact of He-related traps. This enhances the accuracy of tritium inventory models under fusion-relevant irradiation and thermal cycling conditions.