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Superpermeation allows for hydrogen fluxes through metal foil membranes at rates orders of magnitude higher than pressure driven permeation. This process occurs only for hydrogen isotopes, meaning it is hydrogen-selective, and it can work against a pressure gradient, implying pumping capabilities. These characteristics allow for using superpermeation as the base process for a very efficient, selective separation of hydrogen from other gases. However, the efficacy of superpermeation needs further research both experimentally and theoretically. Its efficiency relies on a surface energetic barrier that hinders both adsorption of molecular hydrogen on the downstream side and desorption on the upstream side, while leaving atomic hydrogen absorption unaffected. Such a barrier can be created by a monolayer of non-metallic impurities (usually oxygen) that naturally develops at group 5 metal surfaces. The physics explaining why such a monolayer drastically affects atomic hydrogen reactions are being explored in this work via density functional theory (DFT) calculations for the implementation of which we use the Vienna ab-initio Simulations Package (VASP). By performing structural relaxations and saddle point-searching calculations deploying a dimer method using VASP, energy diagrams for atomic hydrogen absorption are obtained for two representative materials, namely niobium and vanadium. The differences in these diagrams are analyzed and compared in order to determine which material is optimal for superpermeation. To that end, slabs with (1 0 0) surface orientation are compared for the case with and without an O monolayer coverage. The characteristic energies involved according to the diagrams and the types of absorption sites will be key parameters to understand and, eventually, optimize for the emerging phenomena. It was found that the presence of an oxygen monolayer is necessary of superpermeation to occur, and that for the 100 orientation, the vanadium system provides better characteristics.

Erosion, transport and deposition of wall impurities are major concerns in future magnetic fusion devices, both from the perspective of the fusion plasma and the machine wall. An extensive study on molybdenum transport and deposition performed in the TEXTOR tokamak yielded a detailed deposition map that is ideal for benchmark deposition studies. A qualitative benchmark is attempted in this article with the ASCOT code. We set up a full 3D model of the TEXTOR tokamak and studied the influence of different physical mechanisms and their strengths on molybdenum deposition patterns on the simulated plasma-facing components: atomic processes, Coulomb collisions, scrape-off layer (SOL) profiles, source distribution, marker starting energy, radial electric field strength, SOL flow and toroidal plasma rotation. The outcome comprises 13 simulations, each with 100,000 markers. The findings are: • Toroidal plasma movement, either within the LCFS or as SOL flow, is negligible. • SOL profile and marker starting energy have modest impact on deposition. • Source distribution has a large impact in combination with radial electric field profiles. • The E⇀×B⇀ drift has the highest impact on the deposition profiles.

The edge of the RFX-mod (R=2m, a=0.46m) Reversed Field Pinch device is characterized by weak magnetic chaos affecting ion and electron diffusion. Edge particle transport is strongly influenced by a toroidal asymmetry caused by magnetic islands and an ambipolar radial electric field ensures local neutrality, in a way similar to the stochastic edge of tokamaks when resonant magnetic perturbations (RMPs) are applied. The H? emission and floating potential Vf measured in different poloidal and toroidal positions shows a helical shape of the Plasma Wall Interaction, fitting the spatial periodicity of the innermost resonant tearing mode (m/n=1/7) [1]. However, detailed measurements, along the poloidal (parallel) direction, of the electron density and temperature with the Thermal Helium Beam, and of the floating potential Vf with electrostatic probes, show that the response of the edge plasma depends on the poloidal angle, in a more complicated way than a pure 1/7 harmonic. In particular, multiple poloidal harmonics can be recognized in the measurements. The results are robust, because data analysis has been performed with different techniques: in terms of correlations between Vf signals and the corresponding local flux-surface displacement, by the conditional average technique applied at Vf signals, and finally also in terms of a travelling helical angle frame as reference of the measurements. The interpretation of the results is not obvious, but it highlights the fact that the correlation between magnetic islands and kinetic properties of the edge plasma is not a simple one-to-one causal relationship, as it is often assumed in RMP studies in tokamaks.

AbstractFive samples of recrystallized pure tungsten were exposed to transient heat loads using the electron beam of the JUDITH 1 and JUDITH 2 installations of Forschungszentrum Jülich. The heat flux and base temperature were the same for all samples; only the number of pulses and exposure device differed. Transmission electron microscopy was applied to determine the first defects that are introduced during exposure and to compare the effects of the two machines. With increasing number of pulses, first dislocations are formed near the grain boundaries, and then line dislocations and clusters of dislocations appear within the grains. Upon prolonged exposure, the dislocations migrate and cluster in dislocation pile-ups. Comparing exposure in JUDITH 1 to JUDITH 2, the amount of defects is much higher in the samples exposed in JUDITH 1.

Since tungsten (W) was considered as the most promising plasma facing materials (PFMs) in fusion reactors, there has been extensive research on the physical performance of W-PFMs. It is found that under the extreme conditions in a fusion reactor, W-PFMs should be in a nonequilibrium state of high electronic temperature and low ionic temperature. This leads to the possibility of non-thermal phase transitions, where the crystal structure of the tungsten material may change from body-centered cubic (bcc) phase to hexagonal close-packed (hcp) phase or face-centered cubic (fcc) phase. Consequently, it is necessary to investigate the relevant physical properties of hcp-W and fcc-W under the electron-excited state. In this work, the fundamental physical properties, including atomic structures, electronic structures, elastic constants, and vacancy formation energies, of bcc-W, hcp-W and fcc-W, were theoretically calculated at various electronic temperatures. The mechanical stability of these three phases was also systematically analyzed under varying electronic temperatures. The results of this research are expected to provide a certain guidance in the optimization of W-PFMs in future fusion reactors.

For the development of the tritium monitoring system in ITER the hydrogen isotope release by Laser-Induced Desorption (LID) from Be layers is studied to determine the laser parameters for a high desorption efficiency while minimising dust production and surface modifications is also pursued. Be layers of 1 µm thickness with 25–30 at% D and 3 × 1022 D/m2 comparable to JET-ILW areal concentrations [1] have been produced by High Power Impulse Magnetron Sputtering (HiPIMS) on ITER grade W. Laser pulses of 1, 5 and 10 ms duration heat the layer in vacuum in the Fuel REtention DIagnostic Setup (FREDIS) and release the retained D thermally. By mass spectrometry in FREDIS and subsequent Nuclear Reaction Analysis (NRA) inside the laser spot the desorbed and remaining D is quantified. While a pulse duration of 1 ms cannot fully desorb the deuterium, it is found that a single 5 or 10 ms laser pulse with an absorbed energy density of ca. 1.5 MJ/m2 corresponding to a heat flux factor around 20 MW√s/m2 leads to nearly complete desorption of the retained D. This encourages the development of a useful tritium monitoring system, although the present layers produce some dust due to local delamination of the layer on at least 11% of the heated surface (at 1.4 MJ/m2 absorbed energy within 5 ms) and lead to unavoidable crack formation. Keywords: Fuel retention, Beryllium, Tritium monitoring, Laser, Desorption, FREDIS

Oxide dispersion strengthened tungsten (ODS-W) is a potential candidate for plasma-facing materials (PFMs) in future fusion reactors. In this work, deuterium (D) retention and surface blistering in W-1 wt% La2O3 (W-La2O3) have been investigated after exposure to low-energy (40 eV) D plasma with various exposure temperatures (400–600 K) and fluences (3.6 × 1024–1.4 × 1025 D/m2). Surface blistering and D retention exhibit a strong dependence on the exposure temperature and fluence. The most pronounced effect is found at 500 K. The blister-induced defects including dislocations and vacancies are considered to dominate the D retention. At 400 K and 600 K, the D retained in W-La2O3 is governed by unique intrinsic defects including interfaces, micro-pores, and unoxidized La particles. Regarding the exposure fluence, as expected, surface blistering and D retention are positively correlated with it, in which two dominant stages of nucleation and growth for blistering are identified from the changes in area density and size of blisters. Based on the results obtained from W-La2O3, comparisons with W are performed with the exposure condition (500 K, 1.4 × 1025 D/m2) where the blistering and D retention is most pronounced. Although the area density of blisters is similar between the two materials, the average size of blisters is larger in W-La2O3. Notably, an additional high-temperature D desorption shoulder appears in the release spectra of W-La2O3, which is probably due to the particular defects such as interfaces, micro-pores and La particles, and finally resulting in a higher D retention in W-La2O3 than that in W.