Bulk Tungsten Research Articles

Experimental characterization of leading edge cracking on bulk tungsten divertor components during 2017–2019 WEST operation

During the 2017–2019 WEST operation, a set of ITER-like plasma-facing units consisting of tungsten monoblocks was installed on the lower divertor. Despite modest heating power, 27 % of these monoblocks had cracks on their poloidal leading edges, which were not protected by a toroidal bevel. A comprehensive experimental characterization of the 133 cracked monoblocks indicates that damage occurs near the plasma strike point areas where neat cracks separated by 0.15–0.5 mm in the poloidal direction were observed. Cracks are up to 1 mm long and 1 mm deep and did not affect the lower divertor ability to sustain long pulses. The most important parameter affecting crack formation is the relative vertical misalignment between components.

Normal high velocity solid dust impacts on tiles of tokamak-relevant temperature

Runaway electron incidence on plasma facing components triggers explosive events that are accompanied by the expulsion of fast solid debris. Subsequent dust-wall high speed impacts constitute a mechanism of wall damage and dust destruction. Empirical damage laws that can be employed for erosion estimates are based on room-temperature impact experiments. We use light-gas gun shooting systems to accelerate solid tungsten dust to near-supersonic speeds towards bulk tungsten targets that are maintained at different temperatures. This concerns targets cooled down to −100°C with liquid nitrogen and targets resistively heated up to 400 °C. Post-mortem surface analysis reveals that the three erosion regimes (plastic deformation, bonding, partial disintegration) weakly depend on the target temperature within the investigated range. It is concluded that empirical damage laws based on room-temperature measurements can be safely employed for predictions.

Overview of damage to beryllium limiters by unmitigated disruptions and runaway electrons in the JET tokamak with metal walls

Abstract The Joint European Torus (JET) fusion reactor was upgraded to the metallic wall configuration in 2011 which consisted of bulk beryllium (Be) tiles in the main chamber and bulk tungsten (W) and W-coated CFC tiles in the divertor [1]. During each campaign, a series of wall damages were observed; on the upper dump plates (UDP) positioned to the top part of the vessel walls and on the inner wall - mainly affecting the inner wall guard limiters (IWGL). In both cases, it was concluded that the causes of these damages were unmitigated plasma disruptions. In the case of JET with the metallic wall configuration, most of these plasma disruptions were intentionally provoked. The overall objective was to study the behaviour of these phenomena, in order to assess their impact on the wall, improve understanding of morphological material changes, and – based on that - to develop, implement and test mitigation techniques for their prospective use on ITER. The current results bring additional information on the effects of the unmitigated plasma disruptions on the UDPs and are a significant extension of work presented in [2] where the scale of the damage after three operational campaigns on the Be top limiters of JET was highlighted. In addition, new data is presented on the damaging effect that the high energetic runaway electrons had on the Be IWGL in JET.

Tungsten Wire—From Lamp Filaments to Reinforcement Fibers for Composites in Fusion Reactors

The use of drawn tungsten wire marked a breakthrough in electric lighting which led to a significant advancement in human society and technology. Nowadays, thermal light sources using tungsten filaments are being more and more replaced by semiconductor‐based light‐emitting diodes (LEDs). Despite the lower visibility of incandescent lamps in everyday life, new markets and applications for drawn tungsten wires are opening up, which makes this complex material a highly‐studied subject. One of these applications is the use of heavily drawn tungsten wires as high‐performance reinforcement fibers for new composites that can withstand the extreme environment present in a nuclear fusion reactor. The main advantage of tungsten wires for this class of composites is their ductility and high strength. It can be used to increase the high‐temperature strength of copper‐based heat sink materials as well as the fracture toughness of bulk tungsten considered for the use of highly loaded reactor wall components. These new applications for tungsten wire has also given new impetus to the study of their fundamental properties.

Fabrication of Bulk Tungsten Microstructure Arrays for Hydrophobic Metallic Surfaces Using Inductively Coupled Plasma Deep Etching.

Hydrophobic surfaces have attracted great attention due to their ability to repel water, and metallic surfaces are particularly significant as they have several benefits, for example they self-clean and do not corrode in marine environments, but also have several applications in the aircraft, building and automobile industries. Tungsten is an ideal material for metallic surfaces due to its remarkable mechanical properties. However, conventional micromachining methods of micro- or nanostructures, including mechanical fabrication and laser and wet etching are incapable of balancing functionality, consistency and cost. Inspired by the etching process of silicon, deep etching of bulk tungsten has been developed to achieve versatile microstructures with the advantages of high efficiency, large scale and low cost. In this article, fabrication methods of tungsten-based hydrophobic surfaces using an ICP deep etching process were proposed. Micro- or hierarchical structure arrays with controllable sidewall profiles were fabricated by optimizing etching parameters, which then exhibited hydrophobicity with contact angles of up to 131.8°.

Co-deposited layers on gap surfaces of bulk tungsten divertor tiles in JET ITER-like wall: Directional effects and nanostructures

Co-deposited layers on surfaces of bulk tungsten divertor tiles (W lamellae) from the first campaign of JET with the ITER-Like Wall (JET-ILW, 2011–2012) were examined by means of a cross-sectional transmission electron microscope observation. The focus was on geometrical effects in impurity deposition, mainly beryllium (Be), on surfaces located in the poloidal gap separating adjacent lamellae. The study was carried out on the sides (gap surfaces) of two W lamellae from Stack C (located on the outboard part of the horizontal section of the lower divertor), from the region most exposed to the plasma (lamella C23) and in the magnetic shadow of the upstream divertor module (lamella C3). The tile manufacturing process (cold rolling) left shallow grooves, i.e. structures classified as convex (hill) and concave (valley) regions. These regions are decisive for the deposition structure. The main results were: (i) two kinds of impurity deposition features, “homogeneous” and “directional”, have been distinguished; (ii) the directional ones were characterized by nanoscale inclined vertical stripes in the deposition layer; (iii) homogeneous deposition without directional features were in the valley region. The results clearly indicate the impact of the surface finish, even in the tile gaps, on the qualitative and quantitative aspects of deposition.

Crystallographic faceting of bulk tungsten surfaces observed during in situ heating in an environmental scanning electron microscope

The surface faceting of metals has been studied for several decades, typically using experimental methods such as low-energy electron diffraction. In the current paper, surface faceting of tungsten pellets was studied during in situ heating inside an environmental scanning electron microscope (ESEM), which allowed for direct, high-fidelity observation of morphological changes in response to high-temperature annealing under a moderate vacuum (0.8 Torr). Additionally, a second set of tungsten samples was annealed in an ultrahigh vacuum chamber (10−8 Torr), albeit without direct observation via ESEM. This study revealed that oxygen plays a crucial role in tungsten surface faceting and in a morphological transition to vertex rounding. Furthermore, this paper discusses relevant techniques for identifying the crystallographic indices of surface facets. It is demonstrated that a combination of electron backscatter diffraction, tilted-sample imaging from multiple angles, and serial sectioning in a focused ion beam system can be effectively deployed to determine the geometry of faceting, which in turn can be verified with the aid of software for generating Wulff shapes. The influence of processing environment on tungsten surface faceting enables the possibility of designing and controlling the morphology and crystallographic facets of tungsten surfaces.

Hydrogen interaction with vacancy defects in tungsten: Unraveling the influence on diffusion mechanisms and mechanical properties

This study investigates hydrogen diffusion in tungsten at high temperatures using molecular dynamics simulations. It explores the effect of different hydrogen concentrations on diffusion mechanisms, the interaction of hydrogen atoms with vacancy defects, and their impact on the mechanical properties of tungsten. The activation energy for hydrogen diffusion in perfect bulk tungsten with 0.1 % hydrogen concentration is calculated as 0.235 eV, with a diffusion pre-exponential factor of 5.336 × 10-8 m2/s at temperatures ranging from 1000 to 3000 K. The investigation reveals two factors contributing to the increase in activation energy with temperature: migration paths involving TIS-TIS and TIS-OIS-TIS pathways, and an increase in migration energy. Three diffusion regimes are identified, with trapping effects dominating at temperatures below 1500 K, TIS-TIS pathway dominating at intermediate temperatures, and TIS-OIS-TIS pathway becoming more prominent at higher temperatures. Increasing hydrogen concentration up to 2 % (H/W concentrations in fusion environments) has minimal effect on diffusion in perfect tungsten. However, in a W-H-V system with a 1 % vacancy defect, increasing hydrogen concentration increases the effective diffusion coefficient and activation energy. The presence of hydrogen atoms and vacancy defects reduces the tensile strength and elastic modulus of tungsten, with hydrogen atoms having a greater impact on mechanical properties compared to vacancies. The degree of reduction in mechanical properties is proportional to the concentration of hydrogen atoms and vacancy defects.

Tungsten Material Behavior under H2, D2, and He Plasma Interaction Conditions in the Framework of Fusion-Relevant Studies.

In the current study, bulk tungsten material surfaces are exposed to hydrogen, deuterium, and helium plasmas in the radiofrequency domain (13.56 MHz) at an input power of 250 W using the hollow-cathode configuration. The ejected material is collected on titanium substrates at various distances (from 6 mm up to 40 mm). Therefore, the exposed tungsten materials are investigated for surface changes (blister occurrence, dust formation, or nano-structuration), along with the crystallinity, depending on the plasma's exposure times (from 30 min up to 120 min for each plasma type). Also, the collected materials are analyzed (morphological, structural, and statistical investigations) for dust and dust film-like appearance. Plasma discharges are analyzed using two methods: optical emission spectroscopy, and single Langmuir probes, to emphasize the nature of the used plasmas (cold discharges, ~2 eV), along with the presence of tungsten emission (e.g., WI 406.31 nm, WI 421.31 nm) during the plasma lifetime. By using a dedicated protocol, a method was established for obtaining fusion-relevant tungsten surfaces in the hydrogen and deuterium plasma discharges. By using the implemented method, the current paper introduces the possibility of obtaining a new tungsten morphology, i.e., the dandelion-like shape, by using helium plasma, in which the W18O49 compound can be found.

Dynamic formation of preferentially lattice oriented, self trapped hydrogen clusters

A series of MD and DFT simulations were performed to investigate hydrogen self-clustering and retention in tungsten. Using a newly develop machine learned interatomic potential, spontaneous formation of hydrogen platelets was observed after implanting low-energy hydrogen into tungsten at high fluxes and temperatures. The platelets formed along low miller index orientations and neighboring tetrahedral and octahedral sites and could grow to over 50 atoms in size. High temperatures above 600 K and high hydrogen concentrations were needed to observe significant platelet formation. A critical platelet size of six hydrogen atoms was needed for long term stability. Platelets smaller than this were found to be thermally unstable within a few nanoseconds. To verify these observations, characteristic platelets from the MD simulations were simulated using large-scale DFT. DFT corroborated the MD results in that large platelets were also found to be dynamically stable for five or more hydrogen atoms. The LDOS from the DFT simulated platelets indicated that hydrogen atoms, particularly at the periphery of the platelet, were found to be at least as stable as hydrogen atoms in bulk tungsten. In addition, electrons were found to be localized around hydrogen atoms in the platelet itself and that hydrogen atoms up to 4.2 Å away within the platelet were found to share charge suggesting that the hydrogen atoms are interacting across longer distances than previously suggested. These results reveal a self-clustering mechanisms for hydrogen within tungsten in the absence of radiation induced or microstructural defects that could be a precursor to blistering and potentially explain the experimentally observed high hydrogen retention particularly in the near surface region.

Performance of tungsten plasma facing components in the stellarator experiment W7-X: Recent results from the first OP2 campaign

The transition to reactor-relevant materials for the plasma facing components (PFCs) is an important and necessary step to provide a proof of principle that the stellarator concept can meet the requirements of a future fusion reactor by demonstrating high performance steady-state operation. As a first step to gain experience with tungsten as plasma-facing material in the Wendelstein 7-X (W7-X) stellarator, graphite tiles coated with an approximately 10 µm MedC tungsten layer (NILPRP Bucharest) were installed to complete the ECRH beam dump area in two of the five plasma vessel modules over an area of approximately one square meter each. In addition, tungsten baffle tiles are installed (40 tiles in total) with either bulk tungsten as part of NBI shine-through target or with a tungsten heavy alloy (W95-Ni3.5-Cu1.5) to replace the graphite tiles that were previously thermally overloaded. Based on an advanced diffusive field line tracing method and EMC3-Eirene simulations, the overloaded baffle tiles were redesigned by making the tiles thinner (i.e. moving the plasma-facing surface (PFS) away from the hot plasma region) and by reducing the local angle of incidence through toroidal displacement of the watershed. Significant erosion of the tungsten tiles can only be expected if sputtering by impurity ions such as carbon or oxygen ions contributes. However, the resulting central concentration of tungsten and the corresponding radiation losses are expected to be marginal. The expected deposition of carbon on the tungsten surfaces in the baffle regions mitigates further the possible tungsten enrichment in the core plasma. In OP2.1, no adverse effects of the installed tungsten PFCs on the plasma performance were observed during normal plasma operation. With the design changes made in the baffle area, the predicted heat load reductions could be experimentally confirmed.

Wall cratering upon high velocity normal dust impact

Dust-wall high speed impacts, triggered by the termination of runaway electrons on plasma facing components, constitute a source of erosion. Normal high velocity mechanical impacts of tungsten dust on bulk tungsten plates are reproduced in a controlled manner by light gas gun shooting systems. Post-mortem surface analysis revealed that three erosion regimes are realized; plastic deformation, bonding and partial disintegration. The large impact statistics allowed the extraction of reliable empirical damage laws in the latter regime, which can be employed for erosion estimates in future reactors.

A deep learning interatomic potential suitable for simulating radiation damage in bulk tungsten

A deep learning interatomic potential suitable for simulating radiation damage in bulk tungsten

Intrinsic strong light-matter coupling with self-hybridized bound states in the continuum in van der Waals metasurfaces.

Photonic bound states in the continuum (BICs) provide a standout platform for strong light-matter coupling with transition metal dichalcogenides (TMDCs) but have so far mostly been implemented as traditional all-dielectric metasurfaces with adjacent TMDC layers, incurring limitations related to strain, mode overlap and material integration. Here, we demonstrate intrinsic strong coupling in BIC-driven metasurfaces composed of nanostructured bulk tungsten disulfide (WS2) and exhibiting resonances with sharp, tailored linewidths and selective enhancement of light-matter interactions. Tuning of the BIC resonances across the exciton resonance in bulk WS2 is achieved by varying the metasurface unit cells, enabling strong coupling with an anticrossing pattern and a Rabi splitting of 116 meV. Crucially, the coupling strength itself can be controlled and is shown to be independent of material-intrinsic losses. Our self-hybridized metasurface platform can readily incorporate other TMDCs or excitonic materials to deliver fundamental insights and practical device concepts for polaritonic applications.

Investigation of the Influence of Nanoscale Porosity in the Interfacial Layers on the Mechanical Properties of Helium Plasma-Exposed Tungsten

Porous layers emergent from the bulk tungsten (W) interface subsequent to helium (He) plasma exposure, under conditions relevant to thermonuclear experiments, were found to significantly modulate the structural attributes of the W─as deduced through nanoscale indentation measurements. Plastic deformation of nanofibers constituting the fuzzy surface layer, along with an increased yield strength with layer depth, was quantified. A power-law relationship between the indentation modulus (Eind) and the porosity of the fuzzy W was indicated. The surface roughness of the bulk W increases with temperature prior to fuzz formation, determined by AFM. A decrease in the Eind of the bulk W underneath the fuzzy layer with increasing sample temperature when exposed to plasma, in agreement with molecular-dynamics computations, was observed.

DEUTERIUM RETENTION IN N-SEEDED AND UNSEEDED TUNGSTEN LAYERS OBTAINED WITH DCMS AND M-HIPIMS TECHNIQUE

The retention of hydrogen isotopes in tungsten co-deposited layers is a subject severely understudied in the literature compared to hydrogen isotope implantation in bulk tungsten specimens. However, impurity seeding in future next-generation fusion devices can reduce the sputtering threshold of tungsten and potentially increase the contribution of the tungsten layers to the global tritium retention problem. This paper reports on the development of fusion-relevant tungsten-deuterium and nitrogen-seeded tungsten deuterium layers using direct current (DCMS) and multipulse high-power impulse magnetron sputtering techniques (HiPIMS). The four layers resulting from co-deposition in argon-deuterium and argon-deuterium-nitrogen atmosphere were investigated from the deuterium retention and release point of view. Supplementary the crystalline structure of all samples and particularly the composition of the nitrogen-seeded layers were also studied. The tungsten-deuterium layers produced using m-HiPIMS technology showed a higher crystalline coherence compared to layers deposited with DCMS in part owned to the enhanced ion bombardment of the forming layers during deposition. On the other hand, the nitrogen-seeded layers showed without exception the formation of the tungsten-nitride compound. Here on the contrary the crystalline coherence as revealed by X-ray diffraction measurements measurement showed that nitride formation is favored for DCMS compared with m-HiPIMS. The elemental areal densities of the layer constituents were analyzed using Rutherford Backscattering Spectrometry. In this case, only the two nitrogen-seeded layers were analyzed revealing that a higher nitrogen content by a factor of ~2 is retained in DCMS deposited layer compared to HiPIMS.The release of deuterium was higher in the case of m-HiPIMS layers and it was shown that despite the deposition method the release behavior occurs predominantly at temperatures above 800 K suggesting a high energy trapping of deuterium in tungsten layers. Nitrogen seeding influences both the retention mechanisms and increases the deuterium retention by two orders of magnitude in seeded layers compared to unseeded ones.

Feasibility of raised inner strike point equilibria scenario in ITER for detritiation from beryllium co-deposits

In ITER, tritium retention primarily occurs through co-deposition with beryllium. To avoid exceeding the strict tritium inventory limit, efficient tritium recovery techniques are essential. Baking is the ITER baseline for tritium recovery, but its effectiveness in removing tritium from thick beryllium layers is limited. A raised strike point scenario is considered an alternative method for removing tritium from the ITER inner vertical divertor target by heating components via plasma flux. This paper presents SOLPS-ITER code simulations conducted under various conditions, assessing the divertor performance and tritium outgassing of the raised strike point scenario. As the strike point is raised, recycled neutrals are not efficiently baffled by the dome and scrape-off layer, significantly changing the neutral trajectory and ionization source distribution. This improves detachment accessibility but worsens core-edge compatibility compared to the baseline scenario. However, in the partially detached condition, the impact of raising the strike point, perpendicular transport, and q 95 on target heat flux is not significant, as it primarily scales with the input power. Target heat flux is translated to target surface temperature using a simplified heat transfer model that considers the 3D target monoblock geometry and active cooling condition, excluding Be layer thermal properties. For partially detached divertor conditions, the bulk tungsten monoblock surface temperature remains below the baking temperature, which is insufficient for efficient tritium outgassing under the actively cooled ITER divertor condition. However, considering the potential thermal contact resistance between the beryllium and tungsten layers, which may significantly impact temperature distribution, the temperature of the beryllium layer can be raised to a level sufficient for efficient tritium outgassing. Therefore, the raised strike point scenario can be considered as an alternative in-vessel tritium removal technique.

Spin-torque generation using a compositional gradient at the interface between titanium and tungsten thin films

We experimentally demonstrate spin-torque generation using a compositional gradient at the interface between titanium and tungsten thin films. The width of the compositional gradient interface (CGI) between films is varied from 1.4 to 2.0 nm via sputtering. Spin-torque ferromagnetic resonance is observed in the ferromagnetic ${\mathrm{Ni}}_{95}{\mathrm{Cu}}_{5}$ alloy fabricated on a Ti/W bilayer with the CGI. The positive spin torque increases with decreasing CGI width, but a negative spin torque is superimposed owing to the negative spin Hall effect in bulk tungsten. A structural undulation in the CGI eliminates this variation in positive spin torque. The CGI width dependence of the spin torque is associated with the generation of spin and/or orbital angular momentum flow at the CGI. Spin-torque generation using a CGI expands the range of material choice for magnetic nonvolatile memory applications.

Front face shaping of the inner wall tiles in the COMPASS Upgrade tokamak

The central column of the COMPASS Upgrade tokamak, where the plasma discharges will be started and shut down, is covered by a continuous first wall composed of 8 bulk tungsten guard limiters with recessed Inconel tiles in between. In diverted configurations, the plasma will be limited for a short duration (∼0.2 s–0.4 s) during the plasma current ramp-up and ramp-down phases but high heat fluxes due to short power decay lengths are foreseen. An adapted shaping is therefore needed to spread the heat loads over a maximum area. A novel method taking into account the different magnetic equilibria and the incident angle of the field lines was used to design a roof shape able to ensure a constant heat flux on the toroidal direction, accounting for the λqnear−SOL and λqfar−SOL. Due to the diversity of plasma scenarios considered on COMPASS-U, the optimized shape of the inner wall tiles strongly differs from one scenario to another, thus a compromise has to be found among all foreseen scenarios. The tile shaping also needs to take into account possible short wave misalignments. Calculations with the 3D magnetic field line tracing code PFCFlux validates the design, confirming the constant toroidal profile of deposited heat flux for the COMPASS-U workhorse scenarios, or slightly decreasing or increasing heat flux profiles (from the center to the sides of the tile) for scenarios with smaller or higher λqnear−SOL, respectively, within the expected large range (1.7 mm<λqnear−SOL<5 mm), but still providing acceptable loads. Estimated heat fluxes are in the range of 5 MW/m2 to 35 MW/m2 on the 8 guard limiters, receiving 70%–80% of the power deposited to the inner wall. In case of shortwave misalignments up to 0.5 mm in the radial direction, calculations show that leading edges are still protected, confirming the robustness of the design.

Damage effects of high-intensity laser pulse on W for fusion applications: modelling and experiments

In future nuclear fusion reactors plasma-facing materials (PFMs) will be exposed to transient high energy events such as disruptions, edge localized modes (ELMs) and vertical displacement events (VDEs) which may cause melting and vaporization leading to component damage and plasma contamination. The interaction between plasma and refractory metals, namely bulk tungsten (W) and W deposited by plasma spraying (PS-W), has been simulated through a single laser pulse of high energy (≈ 4 J) and short duration (≈ 15 ns). The morphology of the zone affected by laser pulse has been then examined by scanning electron microscopy (SEM) and 3D optical surface profilometry. Both an analytical and a numerical thermal model of the phenomenon based on the specific conditions of heat transfer from the laser spot into each irradiated material were developed. The attention has been focused on the erosion occurring on the material surface because this is the primary cause of damage of PFMs and plasma contamination. Numerical, analytical, and experimental results were found in good agreement.