• Energy Research

A micro-structuring of the tungsten plasma-facing surface can strongly reduce near surface thermal stresses induced by ELM heat fluxes. This approach has been confirmed by numerical simulations with the help of ANSYS software. For experimental tests, two 10 × 10 mm2 samples of micro-structured tungsten were manufactured. These consisted of 2000 and 5000 vertically packed tungsten fibres with dimensions of Ø240 µm × 2.4 mm and Ø150 µm × 2.4 mm, respectively. The 1.2 mm bottom parts of the fibres are embedded in a copper matrix. The top parts of the fibres have gaps about of 10 µm so they are not touching each others. The top of all tungsten fibres was electro-polished. A Nd:YAG laser with a pulse duration 1 ms and a pulse repetition frequency of 25 Hz was used to simulate up to 105 ELM-like heat pulses. No damage on either of the micro-structured tungsten samples were observed. Neon plasma erosion rate and fuel retention of the micro-structured tungsten samples were almost identical to bulk tungsten samples. Keywords: Micro-structured tungsten, PFC, PFM, High heat load, Thermal cycling, Retention, Erosion, Emissivity

In this paper the impact of steady state pure D plasma on WCrY smart alloys at ion energies of 120 and 220 eV is reported. For this purpose a comparison with simultaneously exposed pure W samples is drawn. Different analysis techniques employed for pre- and post-plasma sample analysis hint at a significant depletion of Cr and enrichment of W for lower ion energies. Preferential sputtering leads to enhanced volumetric loss at 220 eV. Analysis of redeposited material indicated local redeposition of Cr. Modelling the ion irradiation with SDTrimSP is used to further interpret experimental results. Depending on the sample temperature during plasma exposure and the magnitude of the ion flux, diffusion of Cr towards the surface is a determining factor for erosion of smart alloys for higher ion energies. Keywords: Smart alloys, Plasma-wall-interaction, Plasma ion irradiation, DEMO, W-based alloys

Deuterium retention and detrapping behavior in the ferritic-martensitic steels EUROFER’97 and P92 after exposure to plasma at high fluences ≥ 1026D/m2 was studied using thermal desorption spectroscopy (TDS), supported by nuclear reaction analysis. Low-temperature irradiation at 450K and fluences ≥ 1026D/m2 with low impact energy D+ / D++He+ ions of 40eV at PSI-2 resulted in a deuterium inventory of 7–18 ×1019D/m2 predominantly at depths ≥8.6µm. Helium admixture led to a reduction of total D retention in both steels, irrespective of surface erosion and composition. The deuterium spectra of both steels displayed one D2 desorption peak at ∼ 540–570K and HD maxima at 540–590, 700–730 and 900–930K. It is suggested that deuterium is mostly retained in the bulk of steel material on interfaces of carbide precipitates and on grain boundaries. Keywords: EUROFER, RAFM, Deuterium, Retention, High fluence, Desorption

Materials are the most urgent issue in nuclear fusion research. Besides tungsten, steels are considered for unifying functional and structural materials due to their cost and mechanical advantages over tungsten. However, the fusion neutrons impose a strong constraint on the ingredients of the steel in order to avoid long lasting activation, while the material has to pertain sputtering resistance, low hydrogen retention, and long-term mechanical stability. In this proof-of-principle, we demonstrate the interesting properties of the new material HiperFer (High performance Ferrite) as a material suitable for fusion applications.The investigation covers neutron activation modelled by FISPACT-II, plasma sputtering and deuterium retention experiments in PSI-2, thermo-mechanical properties and component modelling. The material was found to feature a low nuclear inventory. Its sputtering yield reduces due to preferential sputtering by a factor 4 over the PSI-2 D2 plasma exposure with possible reductions of up to 70 indicated by SD.Trim.SP5 modelling. The exposure temperature shows a strong influence on this reduction due to metal diffusion, affecting layers of 1 µm in PSI-2 at 1150 K exposure for 4 h. Deuterium retention in the ppm range was found under all conditions, together with ∼10 ppm C and N solubility of the ferritic material. The creep and cyclic fatigue resistance exceed the values of Eu-97 steel. As an all HiperFer component, heat loads in the order of 1.5 MW/m² could be tolerated using water-cooled monoblocks. In conclusion, the material solves several contradictions present with alternative reduced-activation steels, but its applications temperatures >820 K also introduce new engineering challenges. Keywords: Nuclear material, Steel, Activation, Plasma-facing material, Structural material, Nuclear fusion

Plasma–wall interactions play a crucial role for the performance of fusion devices and the lifetime of plasma-facing components. In this work the results of some plasma–wall interaction processes, namely the erosion/deposition pattern and the fuel content of divertor baffle tiles exposed in Wendelstein 7-X in the initial phase of island divertor, are presented. Net-deposition of carbon with co-deposited oxygen and hydrogen is determined on the graphite tiles after about 1 hour of plasma operation in hydrogen and helium in operation phase OP 1.2a. C is predominately a result of net-erosion of the graphite target plates and oxygen is the strongest intrinsic impurity in Wendelstein 7-X in OP 1.2a. The hydrogen content distribution on a set of tiles exposed in equivalent positions in three of the five modules of the stellarator was quantified for the first time. Ex-situ performed laser-induced breakdown spectroscopy measurements show the depth-resolved fuel content in deposited layers as well as implantation and diffusion in the base material. Complementary, gas analysis after laser-induced ablation offers quantitative hydrogen content determination in the deposited layers up to 1018 hydrogen atoms/cm2that is non-uniformly distributed over the 95mm×125 mm sized tiles. The results show a toroidal asymmetry with 60% more hydrogen in the stellarator module 2 with respect to module 1 as well as a slight top-down asymmetry for the baffle tiles in upper and lower half modules. A clear dependence of the hydrogen content on the surface temperature during plasma exposition in stellarator discharges in standard magnetic divertor configuration with edge transformation 5/5 was not observed.

The development and application of robust tritium permeation barrier coatings is crucial for a safe and economic fusion reactor operation. Three different tungsten and tungsten nitride layers on Eurofer97 substrates were investigated by deuterium permeation measurements and compared. The microstructure and crystal structure was characterized before and after permeation measurements. The layer permeability is independent of the layer thickness and substrate of the sample. For a reliable comparison of different tritium permeation barrier coatings, the layer permeability of each layer was calculated. With this layer permeability, the permeation flux through potential fusion device components can be estimated. As examples, the permeation flux through a 0.5 cm thick steel component can be reduced by two orders of magnitude by a 2 μm thick WN layer and nearly four orders of magnitude by a 2 μm non-cracked tungsten layer.

Nuclear materials and energy 41, 101812 - (2024). doi:10.1016/j.nme.2024.101812 Published by Elsevier, Amsterdam [u.a.]