Improved scaling law for the prediction of deuterium retention in beryllium co-deposits

Hydrogen isotope co-deposition with Be eroded from the first wall is expected to be the main fusion fuel retention mechanism in ITER. Since good fuel accounting is crucial for economic and safety reasons, reliable predictions of hydrogen isotope retention are needed. This study builds upon the well-established empirical De Temmerman scaling law [1] that predicts D/Be ratios in co-deposited layers based on deposition temperature, deposition rate, and deuterium particle energy. Expanding the data used in the original development of the scaling law with an additional dataset obtained with more recent measurements using a different technique to the original De Temmerman approach, allows us to obtain new values for free parameters and improve the prediction capabilities of the new scaling law. In an effort to improve the model even further, scaling with D2 background pressure was included and a new two-term model derived, describing D retention in low- and high-energy traps separately.

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.

Thermonuclear Fusion Reactor Plasma-Facing Materials under Conditions of Ion Irradiation and Plasma Flux

A review of experimental studies carried out at the NRC “Kurchatov Institute” on plasma-facing thermonuclear fusion reactor materials is presented in the paper. An experimental method was developed to produce high-level radiation damage in materials simulating the neutron effect by surrogate irradiation with high-energy ions. Plasma-surface interaction is investigated on materials irradiated to high levels of radiation damage in high-flux deuterium plasma. The total fluence of accelerated ions (3–30 MeV, 4He2+, 12C3+, 14N3+, protons) on the samples was 1021–1023 m−2. Experiments were carried out on graphite materials, tungsten, and silicon carbide. Samples have been obtained with a primary defect concentration from 0.1 to 100 displacements per atom, which covers the predicted damage for the ITER and DEMO projects. Erosion dynamics of the irradiated materials in steady-state deuterium plasma, changes of the surface microstructure, and deuterium retention were studied using SEM, TEM, ERDA, TDS, and nuclear backscattering techniques. The surface layer of the materials (3 to hundreds µm) was investigated, and it was shown that the changes in the crystal structure, the loss of their symmetry, and diffusion of defects to grain boundaries play an important role. The most significant results are presented in the paper as an overview of our previous work for many years (carbon and tungsten materials) as well as the relatively recent results (silicon carbide).