The synergies between displacement damage creation and hydrogen presence: the effect of D ion energy and flux

In this work the synergism between displacement damage creation and presence of hydrogen isotopes was studied. Tungsten samples were irradiated by 10.8 MeV W ions with or without the presence of D ions with two different energies of 300 eV/D and 1000 eV/D and different temperatures. In order to compare the results obtained with different exposure parameters the samples were afterwards additionally exposed to D ions at 450 K to populate the created defects. By increasing the W irradiation time, ion flux and energy, the increase of D concentration and D retention was observed as measured by nuclear reaction analysis and thermal desorption spectroscopy. By fitting the D depth profiles and D desorption spectra by the rate equation code MHIMS-R we could see that additional fill-levels were populated with higher flux and ion energy which ends up in higher final D concentration and retention as compared to experiments with lower D flux and energy.

Microstructural Evolution and Behavior of Deuterium in a Ferritic ODS 12 Cr Steel Annealed at Different Temperatures

In the present work, an ODS 12 Cr steel was characterized using Electron Microscopy techniques, in an as-received condition and after annealing treatments between 773 K and 1573 K. Results show a complex microstructure, with the presence of fine Y–Ti–O nanoparticles dispersed in the matrix. After annealing at 1573 K, the average diameter of Y–Ti–O nanoparticles increases from ~ 4 to ~ 7 nm and partial recrystallization occurs. The trapping behavior of deuterium in the steel in its as-received state and annealed at 1573 K was investigated. Samples were exposed to low-energy deuterium plasma and analyzed with thermal desorption spectroscopy, after waiting times of 1 day and 25 days. The samples measured 1 day after exposure released a higher total amount of deuterium than the ones measured after 25 days. The effect of waiting time is explained by the release of deuterium, at 300 K, from sites with low activation energy for detrapping, Ed. In the as-received condition, part of the deuterium detrapped at 300 K was re-trapped by high-Ed sites. For the samples in the annealed condition, the redistribution of deuterium from low-Ed to high-Ed sites was not observed, but the total amount of deuterium released was higher.

Reduced defect recovery in self-ion damaged W due to simultaneous deuterium exposure during annealing

Deuterium (D) plasma exposure during annealing of self-ion damaged tungsten (W) is shown to exhibit reduced defect recovery when compared to annealing without D plasma exposure. In these experiments, samples were first damaged with 20 MeV W ions. Next, samples were annealed either with or without simultaneous D2 plasma exposure. The simultaneous annealed samples were first decorated by D2 plasma at 383 K prior to ramping up to an annealing temperature of 473, 573, 673, or 773 K and held for 1 hour with concurrent plasma exposure. The vacuum annealed samples each had a corresponding temperature history but without D$_2$ plasma treatment. Finally, all samples were exposed to D2 plasma at 383 K to decorate any remaining defects. Nuclear reaction analysis (NRA) and thermal desorption spectroscopy (TDS) shows that the simultaneous plasma-exposed and annealed samples exhibited virtually no defect recovery at annealing temperatures of up to 673 K, and had higher D retention than found in the vacuum annealed samples. TDS results indicate that only the lowest detrapping energy defects recover at an 773~K anneal for the simultaneous plasma annealed samples, while the vacuum annealed samples showed defect recovery at all anneal temperatures. This experiment clearly demonstrates that D occupied defects can significantly reduce or eliminate defect annealing in W, and is consistent with the existence of synergistic plasma exposure/displacement damage effects in fusion-energy relevant plasma facing materials.

In-situ measurements of low energy D plasma-driven permeation through He pre-damaged W

Experiments concerning the effect of helium (He) plasma exposure on deuterium (D) plasma-driven permeation (PDP) through tungsten (W) foils in a linear plasma facility has been performed. 0.05 mm thick W foils were exposed to ~2×1020 m-2s-1 He plasma with various fluences at 883 K. After He irradiating, D permeation tests were performed for the samples and retention was also measured by high-resolution thermal desorption spectroscopy (TDS). It was observed that He pre-irradiation resulted in a significant reduction of D permeation and retention in W. Microstructure observation indicated that the surfaces of samples after He irradiation turned rough and He nanobubbles were formed near the surface. The defective structure including He nanobubbles very likely enhances D reemission and accordingly reduces the permeation and retention in He pre-irradiated W.

Crystal structure change in multilayer GeH flakes by hydrogen desorption under ultrahigh vacuum environments

To confirm the feasibility of the theoretically proposed method of forming free-standing germanene [Araidai et al., J. Appl. Phys. 128, 125301 (2020).], we have experimentally investigated hydrogen desorption properties from the hydrogen-terminated germanane (GeH) flakes. Thermal desorption spectroscopy analysis revealed that hydrogen desorption occurred during the heating under an ultrahigh vacuum environment, corresponding to mass loss of 1.0 wt%. Moreover, we have found that using an ultrahigh vacuum ambient and short-time annealing for hydrogen desorption is a key to sustain the crystal structures.

Deuterium retention in W-Ta alloys after low-temperature plasma irradiation and gas exposure

In this study plates of W, W-xTa alloys (x = 1; 3; 5 concentration in at.%) with a large grain size were used as experimental samples. All the samples were polished to a mirror surface and outgassed in vacuum at 1100 K during 2 hours. Sets of W, W-1Ta, W-3Ta, W-5Ta samples were irradiated with low-temperature D plasma up to fluences of 2e25 D/m2. Other sets of W, W-1Ta, W-3Ta, W-5Ta samples were exposed in D2 gas in a temperature range of 425625 K and pressure 103 Pa. The D retention in W and W-Ta alloys was measured by thermal desorption spectroscopy (TDS). An influence of Ta dopant on deuterium retention in W was observed. The dopant of tantalum slightly reduces the accumulation of deuterium in tungsten during gas exposure. Increasing temperature of samples during D-plasma irradiation from 415 K to 615 K reduces deuterium retention up to 2 orders of magnitude.

Deuterium Retention and Release Behavior from Beryllium Co-Deposited Layers at Distinct Ar/D Ratio

Beryllium-deuterium co-deposited layers were obtained using DC magnetron sputtering technique by varying the Ar/D2 gas mixture composition (10/1; 5/1; 2/1 and 1:1) at a constant deposition rate of 0.06 nm/s, 343 K substrate temperature and 2 Pa gas pressure. The surface morphology of the layers was analyzed using Scanning Electron Microscopy and the layer crystalline structure was analyzed by X-ray diffraction. Rutherford backscattering spectrometry was employed to determine the chemical composition of the layers. D trapping states and inventory quantification were performed using thermal desorption spectroscopy. The morphology of the layers is not influenced by the Ar/D2 gas mixture composition but by the substrate type and roughness. The increase of the D2 content during the deposition leads to the deposition of Be-D amorphous layers and also reduces the layer thickness by decreasing the sputtering yield due to the poisoning of the Be target. The D retention in the layers is dominated by the D trapping in low activation binding states and the increase of D2 flow during deposition leads to a significant build-up of deuterium in these states. Increase of deuterium flow during deposition consequently leads to an increase of D retention in the beryllium layers up to 300%. The resulted Be-D layers release the majority of their D (above 99.99%) at temperatures lower than 700 K.

Study on the hydrogen-induced delayed fracture behavior of Q-P980 and MS980

In this paper, the hydrogen diffusion behavior and hydrogen induced delayed fracture (HIDF) of Q-P980 (Q-P: Quenching and Partitioning) and MS980 (MS: Martensitic steel) steels were investigated using hydrogen penetration, slow strain rate tensile (SSRT) tests, thermal desorption spectroscopy (TDS) tests, fracture analysis, and microstructural examination in this paper. The austenite in Q-P980 is massive retained-austenite (RA) with low stability. The TRIP (Transformation Induced Plasticity) effect will occur in the process of strain and change into high carbon martensite. HIDF is caused by a substantial amount of surplus hydrogen being enriched at the border and flaws. The fracture has a broad cleavage surface and is a typical quasi-cleavage fracture. MS980 has been sufficiently tempered, resulting in a substantial quantity of distributed spherical cementite (150nm) precipitating around the lath martensite. This size and form of cementite may successfully trap hydrogen while maintaining the material's mechanical characteristics. And tempering can effectively reduce the local stress level of steel, so MS980 has a very low HE susceptibility. HIDF is related to local stress and hydrogen accumulation. We suppose that Z is a constant and ZC is a critical value which associated to σ and CH (the local stress and local hydrogen concentration), rising as σ and CH rises. The atomic bonds at the crack tip, lattice position and the phase interface will fracture when ZC reaches a particular value Z. Tempering to minimize local stress and carbide precipitation to capture hydrogen are two strategies for reducing hydrogen embrittlement (HE) susceptibility, particularly for dislocation strengthened steel. Microalloying elements can generate precipitates that function as hydrogen traps and obstruct the HELP (Hydrogen Enhanced Localized Plasticity) process, lowering local stress and hydrogen accumulation.