Within the framework of the EUROfusion Consortium, the Characterization of armour, heat sinks materials and joints sub-project of the Work Package Material (WP-MAT) has been dedicated to the development of different tungsten (W) monoblock mock-ups equipped with advanced materials for divertor target applications in the EUDEMO fusion reactor. Assessing the status of the relevant joining interfaces of these mock-ups, not only after fabrication but throughout the whole component lifetime, plays a key role in the qualification process. At the ENEA Special Technologies Laboratory (TES), a number of facilities have been built to perform non-destructive inspections of plasma-facing components for fusion applications by ultrasonic testing (UT). The present work reports on the results of the UT inspections assessing the structural integrity of the relevant joining interfaces of three small-scale mock-ups provided with advanced W armour materials, specifically W-matrix with W2C inclusions consolidated by Spark Plasma Sintering (SPS), K-doped rolled W and K-doped laminated W. The UT examinations are carried out after fabrication and after the high heat flux tests (HHFT) at the neutral beam facility GLADIS. All results confirm the high-quality joining achieved by HIP and HRP. During the HHF tests of mock-ups, after a few hundred HHFT cycles defects are detected at the joining interfaces, due to debonding, delamination and W material cracks mainly affecting the loaded zone. The ultrasonic pulse-echo technique provides not only the size and position of the defects in the plane orthogonal to the ultrasonic beam, but also their depth in the material. During the analysis, the probe is inserted inside the pipe and the mock-up is examined in a cylindrical configuration. The coupling medium (demineralized water) is poured only inside the pipe. The main inspection parameters and the piezoelectric probes are chosen to obtain the maximum resolution in accordance with the thickness and joining interfaces to be analyzed.

The stress state in a plasma-facing component (PFC) under high heat-flux (HHF) loads is the most important factor to determine the lifetime of the component. Stresses in a typical tungsten monoblock type PFC are produced by fabrication process and reactor operation where the component is subjected to HHF loads. In this study, both stress contributions were determined non-destructively by means of neutron diffraction technique. To this end, operational HHF loads were simulated using a high-power neutral hydrogen beam facility (GLADIS) to impose cyclic surface heating at 20 MW/m2. A dedicated small-scale mock-up was fabricated applying hot radial pressing technique to join four tungsten blocks to a CuCrZr alloy cooling pipe via a soft 0.1 mm thick soft copper interlayer. This thin copper interlayer was used to simplify the residual stress profile for this preliminary test. The neutron diffraction measurements were carried out, at room temperature, at two different high-flux reactors: FRM II and the HANARO. Separate stress-relieved tungsten and CuCrZr samples were examined as reference state. The 3D stress tensor was determined in the same external block of the mock-up for both measurements, scanning it from the front face of the tungsten block towards the inner wall of the CuCrZr pipe. The results obtained at these two neutron sources are in good quantitative agreement. Comparing them with the stress profiles before thermal loading, it appears that after the HHF test at GLADIS compressive stresses up to -800 MPa developed in the tungsten block near the interlayer, while the CuCrZr pipe was scarcely affected, probably since the tungsten block accommodated most of the thermal impact. While stress measurements very close to the interlayer might have been affected by spatial resolution issues and error in the reference lattice parameter, the results of these experiments clearly indicate the significant impact of HHF loads on the stress profiles in the tungsten blocks.

Understanding the behavior of tungsten (W) surface damage under the synergistic effects of high heat flux (HHF) loading and helium (He) irradiation is essential for predicting material performance during off-normal operations in ITER. In this study, surface modifications occurring at high temperatures (>2200 K) up to the melting point were investigated by conducting experiments involving two campaigns of vertical displacement events like HHF He neutral beam pulse irradiation on polycrystalline W samples at the test facility Garching LArge DIvertor Sample. As the surface temperature of W increased due to irradiation (2253-3683 K), pinholes appeared on the surface, showing a trend of increasing size and decreasing number density, indicating severe lattice damage. Accordingly, we proposed a model for pinhole growth under high-temperature He irradiation based on thermal activation diffusion of He. The calculated activation energy for He diffusion in this process was found to be 0.51 eV, which is considerably higher than the results obtained from previous simulations (0.021-0.157 eV) (Zhou et al 2010 Nucl. Fusion 50 115010; Becquart and Domain 2006 Phys. Rev. Lett. 97 1-4; Shu et al 2013 Nucl. Instrum. Methods Phys. Res. B 303 84-6; Fu et al 2021 J. Nucl. Mater. 543 152599). This suggests that extensive defects in the matrix have a significant impact on the diffusion of He in high-temperature environments, which is distinct from diffusion behavior at lower temperatures. However, as the surface temperature further increased beyond the melting point, the melting and re-solidification process nearly completely repaired almost all defects induced by He ion irradiation. The re-solidified grains were characterized by being intact, damage-free, and having lower residual stress. This study establishes a foundation for the quantitative analysis of helium migration mechanisms under high-temperature helium irradiation, which lays the foundation for understanding material structural damage behavior under off-normal operations for ITER.

Coating Inconel tiles by tungsten is a necessary step towards the full tungsten first wall coverage of the COMPASS-U tokamak. Thin tungsten coatings on Inconel based on physical vapor deposition were successfully produced consequently to a programme of R&D within the COMPASS Upgrade project by three different suppliers. This contribution presents the qualification phase of these tungsten coatings under COMPASS-U relevant high heat fluxes (50 cycles at 10 MW/m2, 100 cycles at 30 MW/m2) in the neutral beam test facility GLADIS. The behavior of the different tungsten coatings during exposures and the main conclusions from the post-mortem analysis are presented. The most important result is that under COMPASS-U relevant heat fluxes, no damage was observed on the front face of all samples, proving the practicability of such coatings for fusion application.

Cold spray could provide a simple and cost-effective method for coating large areas of plasma facing components regardless of their shape. Low local power deposition and the possibility to perform the process under ambient pressure are favorable features of this technology. In initial cold spray coating tests with pure tungsten, no pore-free coatings could be achieved, but trials with a mixture of W and Ta showed rather good results, resulting in dense coatings with a thickness of up to 2 mm on steel with a W content of up 70 %. In this contribution the initial high heat flux tests were expanded towards systematic studies of the coating behaviour under slow transient loads of up to 20 MW/m(2) and surface temperatures up to 1200 degrees C. As a result, it could be shown that the coatings survive cyclic loading up to surface temperatures of 1000 degrees C even under the highest heat loads of 20 MW/m(2) without any damage or surface modification.

The divertor target of a nuclear fusion reactor is a key in-vessel component with critical operational functions to exhaust particles at associated thermal power. For the European demonstration reactor (EU-DEMO, the maximum heat flux that peaks at the strike point is expected to reach 10 MWm-2 during a long pulse normal operation and up to 20-40 MWm-2 for slow transient events. The reliability and longevity of divertor targets compatible with the harsh and complex loading environment of a fusion power plant is the most crucial requirement to ensure power exhaust and thermal management. Therefore, the mechanical stability of the materials and the structural integrity of the components remain crucial requirements. Since 2014, the design and technology R&D activities for the DEMO divertor have been conducted in the framework of the Work Package Divertor of the EUROfusion Consortium. In late 2020, the preconcept design phase of the DEMO fusion reactor reported that uncertainty of stress states of divertor target components brought by fabrication processes and after high heat flux loads is one critical issue that needs to be addressed soon. Such residual stresses are often unknown from both an experimental and a modelling point of view. This work aims to shed light on this issue by Bragg edge neutron imaging of divertor mock-ups to determine residual strain distributions before and after high-heat flux tests. Moreover, non-destructive evaluations of the structural integrity of such divertor mock-ups have been performed by neutron tomography measurements.

Melt ejection and melt motion of tungsten (W) as a plasma-facing material due to accidental thermal overload events are primary concerns for ITER and DEMO. Previous experiments have revealed that W-1 wt.% La2O3 is an effective material for suppressing cavitation under a single pulsed heat load (Yuan et al 2014 Nucl. Fusion 54 083026). To further investigate this effect, both W and W-1 wt.% La2O3 (WL10) were subjected to multiple melt exposures in the high heat flux facility GLADIS and the tokamak ASDEX Upgrade (AUG). In GLADIS, H/He high-power neutral particle beams were used, with a total energy density of 41-50 MJ m-2 per pulse, adjusted by the pulse duration. In AUG, eight ELMing H-mode deuterium (D) plasma discharges (#37680-87) were employed with samples exposed by the divertor manipulator DIM-II. The results revealed that the surface and cross-sectional morphologies of the resolidified melt layers are similar in both GLADIS and AUG. The pure W samples exhibited pronounced cavitation, with numerous spherical voids in the resolidified layer. In contrast, the WL10 samples developed an undulating surface morphology with a dense resolidified layer at the top, free of spherical voids and lanthanum particles, indicating sustained cavitation suppression even after La2O3 had vaporized from the top surface during previous melting. This suppression effect is likely primarily due to the formation of the undulating surface, which exposes the still La2O3-rich deeper material regions. The La2O3 particles participated in subsequent melting cycles, continuously preventing cavitation. Additionally, we compared D retention in W and WL10 samples subjected to the eight D plasma discharges in AUG. D retention was significantly reduced from 2.5 x 1021 to 4.4 x 1019 D m-2 for W and slightly decreased from 1.5 x 1021 to 1.2 x 1021 D m-2 for WL10 when comparing unmolten and molten samples. Overall, multiple melting experiments in both GLADIS and AUG confirm that W-La2O3 is a good cavitation-resistant material, effectively suppressing cavitation in W melt.

KSTAR (Korea Superconducting Tokamak Advanced Research) plans to upgrade the external heating power to 24 MW by improving heating systems such as NBI, ECH, ICRF, and so on. The upgrade divertor system should guarantee resistance to high heat flux and cooling capacity for the exhausting power in the scrape-off layer domain. The activity for the upgrade of the KSTAR divertor was initiated in 2019, and the upgrade divertor was successfully manufactured and installed in October 2023.The upgraded KSTAR divertor system employs the water-cooled tungsten monoblock, ITER-like divertor type. Tungsten is the most robust and promising plasmafacing material under high heat flux plasma circumstances, and the combination of CuCrZr heat sink and pressurized water coolant is the effective cooling method. The upgraded KSTAR divertor system has a single null configuration and 64 cassette divertor modules placed at the bottom of the vacuum vessel. A divertor module consists of the inner target, the central target, the outer target, and the cassette body, with supports to connect each part. CFD analysis was carried out in the previous study to confirm the thermal stability of a whole divertor module. The result showed the design could be operated within a thermal allowable range in 10 MW/m2 heat flux. The temperature distribution from CFD analysis is applied to the thermo-mechanical analysis. Based on the ASME code, the upgrade KSTAR divertor was estimated for plastic collapse, ratcheting, fatigue, and buckling. The result showed the upgraded KSTAR divertor is reliable from the thermal and mechanical points of view. In the KSTAR divertor, tungsten monoblocks were used only in the straight section due to space constraints. However, in anticipation of a new Korean fusion device following KSTAR, we also explored the production technology for a divertor target that includes a curved section. While the hot isostatic pressing (HIP) process was utilized for dissimilar metal bonding in the KSTAR divertor, we applied the hot radial pressing (HRP) process for manufacturing the curved section target to diversify our production methods. A small mock-up sample of the curved section, manufactured using the HIP and HRP, underwent testing for 20 MW/m2 during 1,000 high-heat flux tests. Both the HIP and HRP samples were completed without any issues. The result confirms the design and quality of the KSTAR divertor target were reliable enough to withstand the heat load, although the recrystallization of tungsten occurred.

An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170 GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, X-divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

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 basic principle of electron beam melting (EBM) technology is the additive generation of structures by the selective melting of metal powder layer by layer with an electron beam under vacuum conditions. The cooling rate of the EBM process can be reduced drastically by increasing the temperature of the powder bed to avoid the formation of solidification cracks by brittle materials such as tungsten (W). This refractory metal is a promising candidate as plasma facing material for future fusion reactors. The selection of tungsten is owing to its physical properties such as the melting point of 3420 °C, the high strength and high thermal conductivity, the low thermal expansion and low erosion rate. Disadvantages are the low ductility, and fracture toughness at room temperature. Furthermore, the manufacturing by mechanical machining, such as milling and turning, is extremely cost and time consuming. An interesting alternative process route to conventional manufacturing technologies is EBM. It allows the near-net shape fabrication of prototype structures with geometrical freedom and has proven its capability for mass production by the manufacturing of hip prostheses made of titanium.This manuscript describes the fabrication of tungsten parts via electron beam melting, with application to the manufacturing of divertor armour. The investigation comprises the microstructure examination, crystallographic texture, as well as mechanical characterization via tensile and Charpy impact testing. This is followed by the presentation of process routes to fabricate mock-ups with different designs and copper cooling structures.Furthermore, the different mock-ups were exposed to high heat flux (HHF) applying transient thermal loads to assess thermal shock and thermal fatigue performance of EBM tungsten.Post mortem analyses were performed quantifying the occurring damage with respect to reference tungsten grades by microscopical means.The achieved results demonstrate the high potential to process tungsten via electron beam melting.

In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements-new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors-as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D-T operations is presented in (King et al Nucl. Fusion submitted).

Tungsten is the primary candidate armour material for the divertor target of the European demonstration fusion power plant. During operation at high temperature, pure tungsten is subject to fracture and recrystallization which results in a loss of strength and worsening of the thermal properties. Additionally, loss-of-coolant accidents with simultaneous air ingress can generate volatile and radioactive tungsten oxides. Advanced W-alloys were developed as alternative and upgrading armour materials of pure tungsten, such as potassium-doped tungsten laminates and self-passivating tungsten alloys. Three mock-ups were manufactured using potassium-doped tungsten laminates, W-10Cr-0.5Y and W-10Cr-0.5Y-0.5Zr as armour materials, each of them consisting of n°4 blocks. The fabrication required optimization and upscaling of the ITER-like process which foresees oxygen-free high conductivity copper as interlayer joined to W-alloy armour block and CuCrZr ITER grade pipe welded to the Cu/W-alloy blocks by hot radial pressing. For quality control of the fabrication steps, non-destructive examination by ultrasonic testing was done on the monoblocks as received, after casting, after hot radial pressing and after high heat flux testing. The results demonstrated that these W-alloys can be used as armour materials of the European demonstration fusion power plant divertor target.

Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm’s law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98( p,y ) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios (‘hybrid’ and ‘elevated’ q -profile) were established on AUG and characterised concerning their plasma performance.

The Divertor Tokamak Test (DTT) facility is a fusion device under construction in Italy. The mission of DTT is to test alternative divertor concepts under integrated physics and technological conditions that can reliably be extrapolated to DEMO. Due to the plasma core characteristics with relevant edge and scrape-off layer (SOL) parameters and a wall entirely in tungsten (W), DTT will provide an extensive set of information useful to select the most appropriate strategy for the power exhaust in DEMO. Several divertors, which may differ in design or/and technologies or/and poloidal profile, will be tested during the life of the machine. The first divertor to be installed will have to accommodate a multitude of strike points, located at various positions according to the different magnetic configurations, which will be tested in the first operational phases of the machine with the aim to identify the most promising. The first divertor will not test innovative technological solutions but will mainly take advantage of the technologies already qualified for the ITER divertor production. Thus, the entire divertor plasma-facing surface is designed to be used as targets: it will be made of W monoblocks joined on CuCrZr pipes (plasma-facing units, PFUs) similar to the ITER targets. With the purpose to increase the flexibility in operational scenarios by maximizing the allowable thermal load for the PFUs, the possibility of using monoblocks with a plasma side reduced thickness was investigated. By reducing the thickness of the armor, it is possible to limit plastic deformation of the monoblock and to preserve the characteristics of the plasma-facing surface during the component lifetime. A thickness between 3 and 4 mm is compatible both the erosion estimates in the DTT divertor area and the manufacturing constraints and therefore proposed for the DTT PFUs. Several mock-ups based on monoblock design were in the past tested under thermal fatigue, confirming the reliability of the monoblock design and the manufacturing processes, but with larger armor thicknesses (6–8 mm). The experimental verification of the monoblock performance with the proposed reduced thickness has been verified in the GLADIS facility at IPP Garching with a thermal load of 20 MW/m $^{2}$ applied for 1000 cycles of 10 s. The results showed the absence of plastic deformation and negligible increase in surface roughness.

The EU-DEMOnstration fusion power plant (DEMO) first wall protection strategy relies on limiter components to face both normal and off-normal plasma transient events. The heat loads during these events are likely to damage the breeding blanket’s first wall otherwise. Since W is the preferred plasma-facing material for EU-DEMO, the plasma-facing component design of the limiters follows considerations based on heat transfer in solids undergoing phase transition. The understanding of this problem has paved the way for a 1-D thermal modeling in MATLAB Thermal Analysis foR Tracking InterFaces under meLting&vaporizaTion-induced plasma Transient Events (TARTIFL&TTE), which has then been improved and extended to 3-D geometries within a Multiphysics environment. Hence, the 3-D TARTIFL&TTE implementation in COMSOL Multiphysics. Although the validation has already started against some data available in the literature and described in the companion paper, dedicated experiments are performed in the Garching LArge DIvertor Sample Test Facility (GLADIS) for melting studies. Carried out as a joint activity between EUROfusion and U.K. Atomic Energy Authority (UKAEA), the aim of these experiments is generating a traceable and controlled experimental database in support of heat transfer studies in solid components undergoing phase transition. The data are here used in support of the 3-D TARTIFL&TTE validation benchmark. To broaden the database, three different materials are chosen, i.e., TZM, W, and SS-316 grade. The requirements defining the experiments comply with the hypotheses behind 3-D TARTIFL&TTE, for it to be able to reproduce the experiments. Therefore, a uniform heat flux on the loaded surface is provided by the H neutral beam on the footprint, and loading time and heat flux magnitude are chosen such that only melting is reached. This allows the liquid metal to stay in place once formed. No attempts to reach vaporization are made, since the vertical position of the target promotes the molten layer sliding under gravity effects. Measured and modeled results (temperature, absorbed energy, and melt layer depth) show good agreement during the melting phase. As a stepwise benchmark, validation will be also sought under vaporization events. Future work is focused on addressing this last point.

Currently, tungsten fibre-reinforced (Wf) composites are regarded as promising materials for plasma-facing components of future magnetic confinement fusion devices. In this context, tungsten fibre-reinforced tungsten (Wf/W) is being investigated as a pseudo-ductile composite material overcoming the intrinsic brittleness of bulk tungsten while tungsten fibre-reinforced copper (Wf/Cu) is being developed as a high-strength composite heat sink material. In this contribution, we discuss the current development status and the progress that has been achieved recently with respect to characterization and upscaling of the aforementioned materials.In cooperation with industry, upscaling of multifilamentary W yarn fabrication was demonstrated. Multilayered W fibre braids were made from such yarns and used for the manufacturing of 400mm long medium-scale tungsten fibre-reinforced copper heat sink tubes. The maturity of short tungsten fibre-reinforced tungsten composites produced by powder metallurgy allowed the fabrication of flat tile mock-ups. Test procedure and first results of high heat flux tests are shown. Finally, we discuss the challenges and the benefits of these composites for the use in high heat flux components.

Cold gas spraying for the production of thick tungsten (W) coatings has been investigated for use at plasma facing components in fusion devices. Since the brittle nature of W strongly impedes its deposition, a systematic study was performed using mixtures of tungsten and tantalum (Ta) powders. Whereas the use of 100% W powder was not successful yet, 2 mm thick coatings on steel were produced by using a W/Ta powder mixture with 90 vol% W, yielding a W content in the coating of 70 vol%. The coatings show negligible porosity and very good adhesion to the substrate. High heat flux experiments on samples with the size 80 × 80 mm2 were performed in order to investigate the behaviour under low (≤4 MW/m2) steady state loads and high power (∼40 MW/m2) transients. During the pulses with low power density, being typical for applications at the main chamber first wall, no defects were observed and a thermal conductivity close to that of the bulk materials was found. During the high power transients lasting for 200 ms cracks parallel to the surface appeared inside the coating.

Within the research along the European Fusion Roadmap, water-cooled divertor PFCs are foreseen in the design of a first fusion demonstration power plant (DEMO) in order to provide reliable heat removal capability. In the frame of this concerted attempt, the Max Planck Institute for Plasma Physics is concentrating on the development and testing of composite materials based on tungsten (W, preferred armour material) and copper (Cu, preferred heat sink material). W fibres (W _f ) as monofilaments and yarns as reinforcement play a central role in these investigations due to their extraordinary properties concerning ductility already at room temperature and high tensile strength. Recent investigations on the impact of radiation damage suggest that the fibres retain their ductility upon irradiation. W reinforced with W fibres (W _f /W) allows to overcome the intrinsic brittleness of W. Quantitative mechanical fracture tests of W _f /W confirm the basic mechanisms of fibre reinforcement and the increased resistance to mechanical fatigue. The good wettability of W with liquid Cu and the absence of any metallurgical solubility make up an ideal material pairing for composite production. W fibre-reinforced Cu (W _f /Cu) cooling tubes provide a rather high thermal conductivity (> 250 W mK ^−1 ) and at least twice the strength of CuCrZr in hoop direction in the temperature range up to at least 500 °C. Very recent neutron irradiation experiments confirm the sustainment of ductility of the W _f /Cu composite. Numerical simulations suggest that thermal stresses in W-Cu PFCs could be strongly reduced by tailoring the local W and Cu volume fraction. This ‘freely’ distributed material composition can be achieved by means of additively manufactured W skeletons consecutively infiltrated by Cu. Investigations with W preforms produced by Laser Beam Powder Bed Fusion and infiltrated by Cu demonstrate the feasibility of this approach while testing of specifically prepared specimen is ongoing.

Using liquid metals confined in capillary porous structures (CPSs) as a plasma-facing component (PFC) could prolong the lifetime of the divertor in the high heat flux area. However, the high atomic number of tin (Sn) limits its acceptable fraction in the main plasma. Therefore, a crucial step in developing this concept is to test it in a tokamak environment, particularly in the diverted plasma region, e.g. ASDEX Upgrade (AUG). In this paper, the design of liquid tin module (LTM) is explained, and the testing in the high heat flux device GLADIS before its use in AUG is presented. The LTM was additively manufactured using selective laser melting, consisting of a 1.5mm porous layer tungsten (W) directly attached to a solid W bulk. The LTM has a plasma-facing area of 16×40mm2 and was filled with 1.54g of Sn. In GLADIS, the module was exposed to power loads between 2 and 8MWm−2 for 1 up to 10s, first unfilled and later filled with Sn. The surface temperature was monitored with infrared imaging and pyrometry. The thermal response was used to compare with simulations in Ansys Mechanical, enabling a determination of the module's effective thermal properties. Sn droplets could be observed on the infrared camera, until a surface temperature of about a 1000°C was reached. The enhanced wetting of tin on the plasma-facing surface, which was observed by a visible camera, suggests that there is a conditioning of the surface, possibly due to the removal of impurities and oxides. Subsequent examinations of the adjacent tile revealed minor Sn leakages emanating from the module's edge. Furthermore, the module showed no indication of mechanical failure. Therefore, these results indicated that the LTM qualifies for the heat fluxes expected in ASDEX Upgrade.