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Self-passivating W-11.4Cr-0.6Y (in wt.%) alloy is a plasma-facing candidate armour material in fusion power plants. In the present work, the as-sintered material, fabricated via ball milling and field-assisted sintering, was annealed at 1000 °C for varying durations to induce phase decomposition. This process leads to the transformation of the initially homogeneous microstructure into two distinct phases: the W-rich phase (αW, Cr) and the Cr-rich phase (αCr, W). Cr-rich phases preferentially form at grain boundaries, where yttrium oxides are also located, and gradually coarsen to the submicron range with increasing annealing time. The chemical compositions of both phases remain relatively stable after 75 h of annealing. The Cr content in (αW, Cr) is 18.6 at.% at 75 h and 17.8 at.% at 100 h. Compared to the as-sintered state, the 100 h-annealed material exhibits significant softening at room temperature and demonstrates increased flexural strength across all tested temperatures, but lower fracture toughness at elevated temperatures. The oxidation behavior of the 100 h-annealed material under humid air at 1000 °C reveals two stages in its TGA curve: inital growth of the inner oxide layer followed by subsequent development of the protecting chromia layer. In contrast, the as-sintered material exhibits a continuous, linear mass increase throughout the oxidation process. These findings present promising prospects of the decomposed microstructure for first wall applications.

The self-passivating yttrium-containing WCr alloy has been developed and researched as a potential plasma-facing armour material for fusion power plants. This study explores the use of yttria (Y2O3) powders instead of yttrium elemental powders in the mechanical alloying process to assess their applicability for this material. Fabricated through field-assisted sintering, WCr-Y2O3 ingots show Y2O3 and Cr-containing oxides (Cr-O and Y-Cr-O) dispersed at grain boundaries (GBs), while WCrY ingots contain Y-O particles at grain boundaries, both resulting from unavoidable oxidation during fabrication. WCr-Y2O3 demonstrates higher flexural strength than WCrY across all temperature ranges, ranging from 850 to 1050 MPa, but lower fracture toughness, between 3 and 4 MPa·√m. Enhanced oxidation resistance is observed in WCr-Y2O3, with lower mass gain as compared to WCrY during the 20-hour oxidation test. This study confirms the effectiveness of both yttria and yttrium in the reactive element effect (REE) for the passivation of WCr alloy, suggesting the potential of Y2O3-doped WCr for first wall applications in a fusion power plant.

Wendelstein 7-X, the world’s largest superconducting stellarator in Greifswald (Germany), started plasma experiments with a water-cooled plasma-facing wall in 2022, allowing for long pulse operation. In parallel, a project was launched in 2021 to develop a W based divertor, replacing the current CFC divertor, to demonstrate plasma performance of a stellarator with a reactor relevant plasma facing materials with low tritium retention. The project consists of two tasks: Based on experience from the previous experimental campaigns and improved physics modelling, the geometry of the plasma-facing surface of the divertor and baffles is optimized to prevent overloads and to improve exhaust. In parallel, the manufacturing technology for a W based target module is qualified.This paper gives a status update of project. It focusses on the conceptual design of a W based target module, the manufacturing technology and its qualification, which is conducted in the framework of the EUROfusion funded WPDIV program. A flat tile design in which a target module is made of a single target element is pursued. The technology must allow for moderate curvatures of the plasma-facing surface to follow the magnetic field lines. The target element is designed for steady state heat loads of 10 MW/m2 (as for the CFC divertor). Target modules of a similar size and weight as for the CFC divertor are assumed (approx.

The self-passivating tungsten-based alloy W-11.4Cr-0.6Y (in wt.%) is a potential plasma-facing material for the first wall of future fusion reactors, which has been shown to suppress oxidation of tungsten and withstand temperatures of up to 1000 °C. In this study, the effect of Y addition on the microstructure and oxidation behavior of W-11.4Cr alloy at 1000 °C is analyzed by comparing it with W-11.4Cr-0.6Y, both prepared using identical synthesis routes. While the binary W-Cr alloy already exhibits improved oxidation resistance over pure W due to the formation of an outer Cr2WO6 layer, it still shows a tendency for spallation and, hence, is not protective. A continuous passivating chromia layer is only obtained with the addition of Y, and we demonstrate that it results in a 50-fold decrease in the oxide growth rate and eliminates the preferred growth of the oxide at edges seen in the binary alloy. Although a porous, complex oxide scale containing mixed oxide layers and WO3 is formed in both cases, the addition of Y results in lower porosity, which makes the oxide scale more adherent. [ABSTRACT FROM AUTHOR]

After a long device enhancement phase, scientific operation resumed in 2022. The main new device components are the water cooling of all plasma facing components and the new water-cooled high heat flux divertor units. Water cooling allowed for the first long-pulse operation campaign. A maximum discharge length of 8 min was achieved with a total heating energy of 1.3 GJ. Safe divertor operation was demonstrated in attached and detached mode. Stable detachment is readily achieved in some magnetic configurations but requires impurity seeding in configurations with small magnetic pitch angle within the edge islands. Progress was made in the characterization of transport mechanisms across edge magnetic islands: Measurement of the potential distribution and flow pattern reveals that the islands are associated with a strong poloidal drift, which leads to rapid convection of energy and particles from the last closed flux surface into the scrape-off layer. Using the upgraded plasma heating systems, advanced heating scenarios were developed, which provide improved energy confinement comparable to the scenario, in which the record triple product for stellarators was achieved in the previous operation campaign. However, a magnetic configuration-dependent critical heating power limit of the electron cyclotron resonance heating was observed. Exceeding the respective power limit leads to a degradation of the confinement.

The self-passivating yttrium-containing WCr alloy has been developed and researched as a potential plasma-facing armour material for fusion power plants. This study explores the use of yttria (Y2O3) powders instead of yttrium elemental powders in the mechanical alloying process to assess their applicability for this material. Fabricated through field-assisted sintering, WCr-Y2O3 ingots show Y2O3 and Cr-containing oxides (Cr-O and Y-Cr-O) dispersed at grain boundaries (GBs), while WCrY ingots contain Y-O particles at grain boundaries, both resulting from unavoidable oxidation during fabrication. WCr-Y2O3 demonstrates higher flexural strength than WCrY across all temperature ranges, ranging from 850 to 1050 MPa, but lower fracture toughness, between 3 and 4 MPa·√m. Enhanced oxidation resistance is observed in WCr-Y2O3, with lower mass gain as compared to WCrY during the 20-hour oxidation test. This study confirms the effectiveness of both yttria and yttrium in the reactive element effect (REE) for the passivation of WCr alloy, suggesting the potential of Y2O3-doped WCr for first wall applications in a fusion power plant. [ABSTRACT FROM AUTHOR]

Tungsten is foreseen presently as the plasma-facing material for divertors in fusion power plants. In order to achieve durable operation of divertors of current fusion reactors, an efficient way of maintaining the divertor functionality is needed. A system capable of in situ tungsten coating of the divertor via low-pressure plasma spraying was proposed to maintain the divertor integrity. In this work, tungsten was deposited on NB31 carbon fibre composite substrates using the low-pressure plasma spraying technology to evaluate the feasibility of this technique. The thickness, porosity, composition, adhesion, and microstructure of the coatings were investigated by scanning electron microscopy image analysis and energy dispersive spectroscopy. Based on the initial results, the spray parameters were iteratively improved in a campaign-based study. The coatings exhibited improving properties through an adjusting of the carrier gas flow, the scanning speed, and the spray distance. By lowering the carrier gas flow, the porosity of the coatings was reduced, resulting in coatings of 98% bulk density. Adjusting the carrier gas flow reduced the amount of semi-molten particles in the coatings significantly. A decrease in both scanning speed and spray distance increased the substrate’s temperature, which led to better adhesion and porosity.

During the spark plasma sintering (SPS) consolidation process, the pressure affects the densification and microstructure evolution of the sintered body. In this paper, the W-Cr-Y-Zr alloy powder was heated to 1000 °C under different applied pressure conditions using spark plasma sintering process, and the effect of pressure on the densification process and microstructure was analyzed. Due to the low sintering temperature, the crystalline size of all the produced W-Cr-Y-Zr alloy is less than 10 nm, which is close to that of the original powders. Cr-rich phase can be detected in the sintered samples due to spinodal decomposition. It is found in this work that the external pressure will increase the contact area between the powder particles, resulting in a higher local pressure at the particle contact, which promotes densification by sliding between the particles under the condition of softening of the particle surface. Additionally, according to the viscous flow theory, the viscous flow activation energy decreases with the increase of pressure. This is because the pressure provides additional driving force to the powder viscous flow process and accelerates the powder shrinkage.

Self-passivating Metal Alloys with Reduced Thermo-oxidation (SMART) are under development for the primary application as plasma-facing materials for the first wall in a fusion DEMOnstration power plant (DEMO). SMART materials must combine suppressed oxidation in case of an accident and an acceptable plasma performance during the regular operation of the future power plant. Modern SMART materials contain chromium as a passivating element, yttrium as an active element and a tungsten base matrix. An overview of the research and development program on SMART materials is presented and all major areas of the structured R&D are explained. Attaining desired performance under accident and regular plasma conditions are vital elements of an R&D program addressing the viability of the entire concept. An impressive more than 104-fold suppression of oxidation, accompanied with more than 40-fold suppression of sublimation of tungsten oxide, was attained during an experimentally reproduced accident event with a duration of 10 days. The sputtering resistance under DEMO-relevant plasma conditions of SMART materials and pure tungsten was identical for conditions corresponding to nearly 20 days of continuous DEMO operation. Fundamental understanding of physics processes undergone in the SMART material is gained via fundamental studies comprising dedicated modeling and experiments. The important role of yttrium, stabilizing the SMART alloy microstructure and improving self-passivating behavior, is under investigation. Activities toward industrial up-scale have begun, comprising the first mechanical alloying with an industrial partner and the sintering of a bulk SMART alloy sample with dimensions of 100 mm × 100 mm × 7 mm using an industrial facility. These achievements open the way to further expansion of the SMART technology toward its application in fusion and potentially in other renewable energy sources such as concentrated solar power stations.

The tungsten (W) foil laminate is an advanced material concept developed as a solution for the low temperature brittleness of W. However, the deformed W foils inevitably undergo microstructure deterioration (crystallization) during the joining process at a high temperature. In this work, joining of the W foil laminate was carried out in a field-assisted sintering technology (FAST) apparatus. The joining temperature was optimized by varying the temperature from 600 to 1400 °C. The critical current for mitigating the microstructure deterioration of the deformed W foil was evaluated by changing the sample size. It is found that the optimal joining temperature is 1200 °C and the critical current density is below 418 A/cm2. According to an optimized FAST joining process, the W foil laminate with a low microstructure deterioration and good interfacial bonding can be obtained. After analyzing these current profiles, it was evident that the high current density (sharp peak current) is the reason for the significant microstructure deterioration. An effective approach of using an artificial operation mode was proposed to avoid the sharp peak current. This study provides the fundamental knowledge of FAST principal parameters for producing advanced materials.

W-Cr-Y smart alloys are potential material candidates for plasma facing components due to their protective behaviour during the loss-of-coolant accident (LOCA), while maintaining beneficial properties of W during the normal operation of the fusion power plant. During plasma exposure, the lighter alloying elements are preferentially sputtered at the surface, but in case of a LOCA, the plasma quenches and sputtering stops and diffusion of the alloying elements to the surface becomes intensive. The diffusion of Cr to the surface due to alloying elements (Y, Ti) yields a protective oxide layer stopping the sublimation of WO3. The phase stability and short-range ordering of ternary alloys in W-Cr-Y(Ti) systems has been investigated, using combination of Density Functional Theory (DFT) and Cluster Expansion (CE) methods with Monte-Carlo (MC) simulations. It has been found out from the DFT calculations, that all pairs in the W-Cr-Y system have positive values of the enthalpy of mixing, while most of the Cr-Ti and Ti-W binary structures have negative enthalpies of mixing. The shift in the short-range order as a function of temperature between Cr and W has been predicted as a result of Y addition in W-Cr-Y alloys compared to W70Cr30, by around 400 K towards lower temperatures. A strong tendency towards clustering of Y has been observed even at elevated temperatures (1800 K). The decrease of the order–disorder transition temperature (ODTT) as a result of the Y addition has been observed, while the addition of Ti has not shown any significant changes in the ordering of W-Cr-Ti alloys compared to W-Cr alloy. Our MC simulations showed that for the W70Cr29Y1 alloy the enthalpy of mixing (Hmix) value is positive in the whole analysed temperature range. Free energy of mixing above 1000 K has been calculated from the first nearest neighbours approximation for W70Cr29Y1 and W70Cr29Ti1 alloys. The results of the present investigations provide an insight enabling for optimizing chemical composition of materials for future plasma facing components.