Dissociative Electron Attachment Cross Sections for H₂ and D₂

New measurements of the absolute cross sections for dissociative electron attachment (DEA) in molecular hydrogen and deuterium are presented which resolve previous ambiguities and provide a test bed for theory. The experimental methodology is based upon a momentum imaging time-of-flight spectrometer that allowed us to eliminate any contributions due to electronically excited metastable neutrals and ultraviolet light while ensuring detection of all the ions. The isotope effect in the DEA process in the two molecules is found to be considerably larger than previously observed. More importantly, it is found to manifest in the polar dissociation process (also known as ion pair production) as well

Dynamic fuel retention in tokamak wall materials: An in situ laboratory study of deuterium release from polycrystalline tungsten at room temperature

International audienceRetention of deuterium ion implanted in polycrystalline tungsten samples is studied in situ in an ultra-high vacuum apparatus equipped with a low-flux ion source and a high sensitivity thermo-desorption setup. Retention as a function of ion fluence was measured in the 10^17 -10^21 D+/m^2 range. By combining this new fluence range with the literature in situ experimental data, we evidence the existence of a retention = fluence^ 0.645±0.025 relationship which describes deuterium retention behavior on polycrystalline tungsten on 8 orders of magnitude of fluence. Evolution of deuterium retention as a function of the sample storage time in vacuum at room temperature was followed. A loss of 50% of the retained deuterium is observed when the storage time is increased from 2 h to 135 h. The role of the surface and of natural bulk defects on the deuterium retention/release in polycrystalline tungsten is discussed in light of the behavior of the single desorption peak obtained with Temperature Programmed Desorption

Deuterium retention and transport in ion-irradiated tungsten exposed to deuterium atoms: role of grain boundaries

The influence of grain boundaries on deuterium (D) retention and transport was investigated in nanocrystalline tungsten (W) by exposing the samples to sub eV D atoms. Thin tungsten films with nanometer-sized grains were produced by pulsed laser deposition on tungsten substrates. Their grain size was increased up to one micrometer by thermal annealing in vacuum up to 1223 K. Irradiation damage was created by 20 MeV W ions at 290 K. The transmission electron microscopy analysis showed one order of magnitude larger dislocation density in nanometer-grained samples compared with the larger-grained samples. The samples were after W irradiation exposed to 0.3 eV D atoms at 600 K. D retention and D depth profiles were measured by nuclear reaction analysis. In the as-deposited nanometer-grained samples, D populated the damaged region more than three times faster than in the samples with larger grains, indicating that grain-boundaries increase D transport through the material. The concentration of defects was assessed by the final D concentration in the samples. The sample with the smallest grain size showed slightly larger D concentration in the irradiated area, but the difference in the D concentration was not substantial between different-grained samples. A large D concentration in the non-irradiated nanometer-grained sample was measured which is an indication for a high defect density in the initial material. From our observations, it can be postulated that the nanocrystalline microstructure did not substantially influence the generation of irradiation-induced defects by defect annihilation at grain boundaries

Influence of surface morphology on erosion of plasma-facing components in H-mode plasmas of ASDEX Upgrade

Net erosion of plasma-facing materials was investigated at the low-field-side (outer) strike-point area of the ASDEX Upgrade (AUG) divertor during H-mode discharges with small and frequent ELMs. To this end, Au and Mo marker samples with different surface morphologies and geometries were exposed to plasmas using the DIMII divertor manipulator. The results were compared to existing erosion and deposition patterns from various Land H-mode experiments, in the latter case the main difference was the size and frequency of the ELMs. We noticed that increasing surface roughness reduces net erosion but less than what is the case in L-mode. On the other hand, net-erosion rates in H-mode are generally 2–5 times higher than the corresponding L-mode values, in addition to which exposure in H-mode conditions results in strong local variations in the poloidal and toroidal erosion/deposition profiles. The latter observation we associate with the large migration length, on the order of several cm, of the eroded material, resulting in strong competition between erosion and re-deposition processes especially at poloidal distances > 50 mm from the strike point. Considerable net erosion was measured throughout the analysed poloidal region unlike in L-mode where the main erosion peak occurs in the vicinity of the strike point. We attribute this qualitative difference to the slow decay lengths of the plasma flux and electron temperature in the applied H-mode scenario. Both erosion and deposition require detailed analyses at the microscopic scale and the deposition patterns may be drastically different for heavy and light impurities. Generally, the rougher the surface the more material will accumulate on locally shadowed regions behind protruding surface features. However, rough surfaces also exhibit more non-uniformities in the quality or even integrity of marker coatings produced on them, thus complicating the analyses of the experimental data. We conclude that local plasma parameters have a huge impact on the PFC erosion rates and, besides incident plasma flux, surface morphology and its temporal evolution have to be taken into account for quantitative estimates of erosion rates and PFC lifetime under reactor-relevant conditions

Tritium retention in W plasma-facing materials : Impact of the material structure and helium irradiation

This article has an erratum: DOI 10.1016/j.nme.2020.100729Plasma-facing materials for next generation fusion devices, like ITER and DEMO, will be submitted to intense fluxes of light elements, notably He and H isotopes (HI). Our study focuses on tritium (T) retention on a wide range of W samples: first, different types of W materials were investigated to distinguish the impact of the pristine original structure on the retention, from W-coated samples to ITER-grade pure W samples submitted to various annealing and manufacturing procedures, along with monocrystalline W for reference. Then, He and He-D irradiated W samples were studied to investigate the impact on He-damages such as nano-bubbles (exposures in LHD or PSI-2) on T retention. We exposed all the samples to tritium gas-loading using a gentle technique preventing any introduction of new damage in the material. Tritium desorption is measured by Liquid Scintillation counting (LSC) at ambient and high temperatures (800 degrees C). The remaining T inventory is then measured by sample full dissolution and LSC. Results on T inventory on He exposed samples highlighted that in all cases, tritium desorption as a gas (HT) increases significantly due to the formation of He damages. Up to 1.8 times more T can be trapped in the material through a competition of various mechanisms, but the major part of the inventory desorbs at room temperature, and so will most likely not take part to the long-term trapped inventory for safety and operational perspectives. Unfortunately, investigation of "as received" industrial W (used for the making of plasma-facing materials) highlighted a strong impact of the pre existing defects on T retention: up to 2.5 times more T is trapped in "as received W" compared to annealed and polish W, and desorbs only at 800 degrees C, meaning ideal W material studies may underestimate T inventory for tokamak relevant conditions.Peer reviewe

Ion beam analysis of fusion plasma-facing materials and components : facilities and research challenges

Following the IAEA Technical Meeting on ‘Advanced Methodologies for the Analysis of Materials in Energy Applications Using Ion Beam Accelerators’, this paper reviews the current status of ion beam analysis (IBA) techniques and some aspects of ion-induced radiation damage in materials for the field of materials relevant to fusion. Available facilities, apparatus development, future research options and challenges are presented and discussed. The analysis of beryllium and radioactivity-containing samples from future experiments in JET or ITER represents not only an analytical but also a technical challenge. A comprehensive list of the facilities, their current status, and analytical capabilities comes alongside detailed descriptions of the labs. A discussion of future issues of sample handling and the current status of facilities at JET complete the technical section. To prepare the international IBA community for these challenges, the IAEA technical meeting concludes the necessity for determining new nuclear reaction cross-sections and improving the inter-laboratory comparability by defining international standards and testing these via a round-robin test.Peer reviewe