• Energy Research

Samples of EUFROFER, a reduced activation ferritic martensitic steel, were exposed in the linear plasma device Pilot-PSI to a deuterium (D) plasma with incident ion energy of ∼40 eV and incident D flux of 2–6 × 1023 D/m2s to fluences up to 1027 D/m2 at surface temperatures ranging from 400 K to 950 K. The main focus of the study lays on the surface morphology changes dependent on the surface temperature and the surface composition evolution, e.g., the enrichment in tungsten; but also the erosion and the D retention are studied. The created surface morphology varies strongly with surface temperature from needle-like to corral-like structures. The visible lateral length scale of the formed structures is in the range of tens of nanometres to above 1 µm and exhibits two thermal activated regimes below and above ∼770 K with activation energies of 0.2 eV and 1.3 eV, respectively. The lateral variation of the enrichment of heavy elements on the surface is correlated to this surface morphology at least in the high temperature regime, independent of the origin of the enrichment (intrinsic from the sample or deposited by the plasma). Also the erosion exhibits temperature dependence at least above ∼770 K as well as a fluence dependence. The amount of deuterium retained in the top 500 nm is almost independent of the exposure temperature and is of the order of 1018 D/m2, which would correspond to a sub-monolayer D coverage on the surface. The retained D in the volume summing up over the complete samples exceeds the D retained close to the surface by one order of magnitude.

Divertor designs involving liquid lithium have been proposed as an alternative to solid designs and wall conditioning techniques. However, Li affinity with tritium poses a risk for the fuel cycle. This study investigates deuterium retention in pre-lithiated samples and Li-D co-deposits in the DIII-D tokamak, making for the first time a direct comparison between Li-D co-deposits and pre-deposited Li films. Samples were exposed to H-mode plasmas in the far scrape-off layer (SOL), and Li powder was injected in-situ with the impurity powder dropper to study the uniformity of Li coatings, and the dependence of fuel retention on Li thickness. The results show that at temperatures below the melting point of lithium, deuterium retention is independent of the thickness of pre-deposited Li layers, with Li-D co-deposits being the primary factor for fuel retention. Both pre-deposited and in-situ deposited Li showed lower erosion than predicted by sputtering yield calculations. These results suggest that fuel retention in fusion reactors using lithium in the divertor will likely be dominated by co-deposits rather than in the divertor itself. If one desires to use Li to achieve flatter temperature profiles, operando Li injection is advantageous over pre-deposited Li films, at least at temperatures below the melting point of lithium.

Liquid metal infused capillary porous structures (CPSs) are considered as a potential divertor solution for DEMO due to their potential power handling capability and resilience to long term damage. In this work the power handling and performance of such Sn-based CPS systems is assessed both experimentally and via modelling. A Sn-CPS target was exposed to heat fluxes of up to 18.1 MW m−2 in He plasma in the Pilot-PSI linear device. Post-mortem the target showed no damage to nor any surface exposure of the underlying W-CPS felt. The small pore size (∼40 µm) employed resulted in no droplet formation from the target in agreement with calculated Rayleigh-Taylor and Kelvin-Helmoholtz instability thresholds. The temperature response of the Sn-target was used to determine the thermal conductivity of the mixed Sn-CPS material using COMSOL modelling. These values were then used via further finite element analysis to extrapolate to DEMO relevant monoblock designs and estimate the maximum power handling achievable based on estimated temperature windows for all component elements of the design. For an optimized design a heat-load of up to 20 MW m−2 may be received while the use of CPS also offers other potential design advantages such as the removal of interlayer requirements.

Facing extreme plasma loads, the structural integrity of the tungsten divertor is crucial for ITER, an engineering marvel in nuclear fusion reactors, to achieve its fusion performance targets. Induced by repeated transient heating from plasmas, the thermal fatigue damage of tungsten–typically accompanied by the formation of surface relief–has been identified as a critical issue but an in-depth understanding is still lacking. Here, we report the formation of amorphous and anisotropic surface relief on ITER-grade tungsten surfaces under ITER-relevant hydrogen plasma loads. Measured by both electron backscatter diffraction over large fields of view and transmission Kikuchi diffraction of site-specific lamellae, such surface relief preferentially forms on grains with {110} planes parallel to the surface. This cannot be explained by the orientation-dependent resolved shear stress according to the Schmid law, threshold displacement energy anisotropy, or oxidation anisotropy. Furthermore, surface relief amorphization is revealed by high-resolution transmission electron microscopy imaging and selected area electron diffraction analysis, and is explained by a novel vacancy-induced amorphization mechanism. The results provide new insights into the thermal fatigue behavior of tungsten for fusion applications.

The retention of deuterium (D) and helium (He) is studied in pure tungsten after high flux mono-plasma exposure. The recrystallized and plastically deformed tungsten samples are studied to clarify the impact of the material microstructure, in particular dislocation density, on the trapping and release of D and He. Thermal Desorption Spectroscopy (TDS) measurements are performed to reveal the release stages and quantify the retention. Preliminary transmission electron microscopy study was applied to clarify the microstructural modifications induced by the plasma exposure to support the discussion and conclusions. It has been demonstrated that plastic deformation causes considerable suppression of He release within the explored limit of the TDS temperature-1300 K. This is opposite to what is found for the pure D exposure, where the plastic deformation evidently enhances the D retention, given equivalent exposure conditions in terms of surface temperature and ion fluence.

Remobilization is one of the most prominent unresolved fusion dust-relevant issues, strongly related to the lifetime of dust in plasma-wetted regions, the survivability of dust on hot plasma-facing surfaces and the formation of dust accumulation sites. A systematic cross-machine study has been initiated to investigate the remobilization of tungsten micron-size dust from tungsten surfaces implementing a newly developed technique based on controlled pre-adhesion by gas dynamics methods. It has been utilized in a number of devices and has provided new insights on remobilization under steady-state and transient conditions. The experiments are interpreted with contact mechanics theory and heat conduction models.