B. Tyburska-Püschel

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

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.

Deuterium trapping at defects created with neutron and ion irradiations in tungsten

The effects of neutron and ion irradiations on deuterium (D) retention in tungsten (W) were investigated. Specimens of pure W were irradiated with neutrons to 0.3 dpa at around 323 K and then exposed to high-flux D plasma at 473 and 773 K. The concentration of D significantly increased by neutron irradiation and reached 0.8 at% at 473 K and 0.4 at% at 773 K. Annealing tests for the specimens irradiated with 20 MeV W ions showed that the defects which play a dominant role in the trapping at high temperature were stable at least up to 973 K, while the density decreased at temperatures equal to or above 1123 K. These observations of the thermal stability of traps and the activation energy for D detrapping examined in a previous study (≈1.8 eV) indicated that the defects which contribute predominantly to trapping at 773 K were small voids. The higher concentration of trapped D at 473 K was explained by additional contributions of weaker traps. The release of trapped D was clearly enhanced by the exposure to atomic hydrogen at 473 K, though higher temperatures are more effective for using this effect for tritium removal in fusion reactors.

Sub-surface structures of ITER-grade W (Japan) and re-crystallized W after ITER-similar low-energy and high-flux D plasma loadings

Tungsten is a promising candidate for plasma-facing materials in fusion reactors. In this work, two types of W materials were investigated: (i) sintered and forged tungsten (ITER-grade Japan, grain sizes 2–100 μm, elongated normal to the surface) and (ii) the same W grade, but after additional re-crystallization (at 2073 K, grain size ~50 μm). The samples were exposed to deuterium with an ion energy of 38 eV D−1, a fluence of 1027 D m−2 and a flux of ~1022 D m−2 s−1 in a plasma generator at elevated temperatures (320–700 K). The D retention (determined by thermal desorption spectroscopy and nuclear reaction analysis) of both sample types is compared. The samples were analysed with scanning electron microscopy combined with a focused ion beam for iterative cross-sectioning to obtain three-dimensional (3D) data of the sub-surface. Electron backscattered diffraction was applied to determine the grain orientation and deformation. First nano-secondary ion mass spectroscopy investigations were performed on a D-loaded sample to analyse the lateral accumulation of H/D on the surface.

Hydrogen isotope exchange in tungsten irradiated sequentially with low-energy deuterium and protium ions

Hydrogen isotope exchange in tungsten was investigated at various temperatures both after sequential exposure to low-energy deuterium (D) and protium (H) plasmas and after sequential irradiation with low-energy D and H ions. The methods used were thermal desorption spectroscopy, and the D(3He,p)4He nuclear reaction at 3He energies varied from 0.69 to 4.0 MeV, allowing the determination of the D concentration at depths of up to 6 μm. It was found that a major portion of the deuterium initially accumulated in the D-implanted W is released on subsequent exposure to H plasma or irradiation with H ions. Depth profiling of D without and with subsequent H implantation shows strong replacement close to the surface near room temperatures, but extending to all analyzable depths at elevated temperatures.

On the reduction of deuterium retention in damaged Re-doped W

Modelling tritium retention in radiation-damaged pure-tungsten samples, the concentration of deuterium retained in the tungsten-ion-induced damage zone decreases with increasing exposure temperature; whereas for a similarly treated tungsten–rhenium alloy, it can drop significantly faster than in pure tungsten, given sufficiently high temperatures. In contrast, similar changes in retention behaviour were not observed in undamaged samples. Based on these findings, as well as on corresponding previous TEM results, it is concluded that the concentration of high-energy radiation-induced defects responsible for trapping of deuterium is lower in the alloy than in pure tungsten. Therefore, a small amount of transmutation rhenium in damaged tungsten should be able to keep tritium retention to a low level in ITER.