Plasma Facing Armor Research Articles

Insights into microstructural evolution of tungsten-tungsten carbide plasma facing composite materials prepared by field assisted sintering technique

Combining a stable, resilient W-W2C or WC as plasma-facing material, with thermally more conductive and ductile W as a structural material, may provide a solution for the demanding environment in the fusion reactor chamber. In order to ensure the structural stability of such layered composite, their phase composition and thermal stability at elevated temperatures and cyclic heat loads needs to be evaluated.In situ densification and joining of W and W-xWC materials have been performed by field-assisted sintering technique. Microstructural analysis revealed that the interface layer formed between W and pure WC consists primarily of large W2C grains. Initially, equiaxed W2C grains are formed following the decarburisation of WC due to the large C concentration gradient. Further C diffusion results in the formation of larger anisotropic grains growing preferentially in the 〈0001〉 direction. A similar microstructure was observed when joining WC-reinforced W (W-xWC, x = 7, 17, 28, 33 at. % C) with WC. On the other hand, the joining of W with WC-reinforced W resulted in a sound joint with no observable grain growth at the interface, even at large C concentrations (e.g. 28 at. %). As formed, microstructures were stable after thermal ageing at 1600 °C for 24 h with no observable change in major microstructural features.The combination of WC-reinforced W as high strength plasma facing armour material and W-structural material offers a promising composite solution for improved thermal stability under high cyclic heat loads expected in future fusion power plants.

Advances in material phase change modelling approach for EU-DEMO limiter’s plasma-facing components

Within the EU-DEMO first wall protection framework, work on limiter’s plasma-facing component design has started under plasma disruptive events (Richiusa et al., 2022). Starting from the rationale behind the TARTIFL&TTE software (Richiusa et al., 2022), this companion paper describes the progress on the engineering modelling of the plasma-facing material phase change under high heat flux, with the aid of COMSOL Multiphysics® software. The aim is to develop a reliable technique which can be used by designers for predicting how much of the solid armour undergoes phase change. This helps satisfy requirements for actively cooled components, such as at which armour depth safely locating the cooling system, and if its design can safely handle the heat transfer in the resulting component configuration after the disruption is extinguished. A few changes to the driving idea in Richiusa et al. (2022) will be also highlighted.The multiphysics software allows us to implement a 1D model which can be extended to 2D/3D geometries subjected to both uniform and non-uniform heat flux. It also offers the capability of conjugated heat transfer in solids and liquids by coupling the two different domains. Although this multiphysics approach is also investigated, no effort in melting layer motion modelling is done. Therefore, the equivalent way of validating this approach while reducing its computational time is working with one single solid domain, within which any liquid phase changes are tracked by an apparent heat capacity formulation. The vapour domain is not modelled. The material removal due to the evaporative mass flux is modelled by means of moving mesh frames which push the recessing liquid interface backwards according to gas kinetics-driven boundary conditions. The melt pool is not removed during the transient. Mass balance considerations drive the liquid-to-vapour interface velocity.The 3D Multiphisics implementation (3D-TARTIFL&TTE), is here supported by a benchmark activity and an application to the limiter’s plasma-facing armour, whose preferred chosen thickness is 20 mm.

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Development of refractory metals for application as plasma-facing armour material remains among priorities of fusion research programmes in Europe, China and Japan. Improving the resistance to high temperature recrystallization, enhancing material strength to sustain thermal fatigue cracking and tolerance to neutron irradiation are the key indicators used for the down selection of materials and manufacturing processes to be applied to deliver engineering materials. In this work we investigate the effect of neutron irradiation on mechanical properties and microstructure of several tungsten grades recently developed. Neutron irradiation campaign is arranged for screening purposes and therefore is limited to the fluence relevant for the ITER plasma facing components. At the same time, the neutron exposure covers a large span of irradiation temperatures from 600 up to 1000 °C. Four different grades are included in the study, namely: fine-grain tungsten strengthened by W-carbide (W–4wt.% W2C), fine-grain tungsten strengthened by Zr carbides (W–0.5% ZrC), W alloyed with 10 at.% chromium and 0.5 at.% yttrium (W–10Cr–0.5Y) and technologically pure W plate manufactured according to the ITER specification by Plansee (Austria). The strengthening by W2C and ZrC particles leads to an enhanced strength, moreover, the W–0.5ZrC material exhibits reduced DBTT (compared to ITER specification grade) and is available in the form of thick plate (i.e. high up-scaling potential). The W–10Cr–0.5Y grade is included as the material offering the self-passivation protection against the high temperature oxidation.

Probabilistic analysis of cleavage fracture in commercial polycrystalline tungsten

Polycrystalline tungsten is a promising candidate as plasma-facing armour material in future fusion reactors. Below the DBTT tungsten shows brittle or semi-brittle behaviour with no or little plastic deformation before cleavage fracture. Moreover, tungsten has an inherent low fracture toughness as well as a large scatter within the strength properties. This requires a statistical treatment of the strength values. Therefore, four-point bending tests in the temperature range of room temperature (RT) to 400 °C were performed to allow the Weibull analysis in brittle failure regime and/or the Beremin model analysis in case of onset of plastic behaviour while approaching the DBTT temperature. The analyses are supported by FE simulations. Weibull and Beremin parameter values are determined from the simulation results using post-processing codes.

Thermal diffusivity recovery and defect annealing kinetics of self-ion implanted tungsten probed by insitu transient grating spectroscopy

Tungsten is a promising candidate material for plasma-facing armour components in future fusion reactors. A key concern is irradiation-induced degradation of its normally excellent thermal transport properties. In this comprehensive study, thermal diffusivity degradation in ion-implanted tungsten and its evolution from room temperature (RT) to 1073 K is considered. Five samples were exposed to 20 MeV self-ions at RT to achieve damage levels ranging from 3.2 × 10−4 to 3.2 displacements per atom (dpa). Transient grating spectroscopy with insitu heating was then used to study thermal diffusivity evolution as a function of temperature. Using a kinetic theory model, an equivalent point defect density is estimated from the measured thermal diffusivity. The results showed a prominent recovery of thermal diffusivity between 450 K and 650 K, which coincides with the onset of mono-vacancy mobility. After 1073 K annealing samples with initial damage of 3.2 × 10−3 dpa or less recover close to the pristine value of thermal diffusivity. For doses of 3.2 × 10−2 dpa or higher, on the other hand, a residual reduction in thermal diffusivity remains even after 1073 K annealing. Transmission electron microscopy reveals that this is associated with extended, irradiation-induced dislocation structures that are retained after annealing. A sensitivity analysis shows that thermal diffusivity provides an efficient tool for assessing total defect content in tungsten up to 1000 K.

W + Cu and W + Ni Composites and FGMs Prepared by Plasma Transferred Arc Cladding.

Tungsten-based materials are the most prospective candidates for plasma-facing components of future fusion devices, such as DEMO. W-based composites and graded layers can serve as stress-relieving interlayers for the joints between plasma-facing armor and the cooling or structural parts. Coating/cladding techniques offer the advantages of eliminating the joining step and the ability to coat large areas, even on nonplanar shapes. In this work, W + Cu and W + Ni composites were prepared by pulsed plasma transferred arc (PTA) cladding on several different substrates. Optimization of the process was carried out with respect to powder mixture composition and process parameters like arc current, plasma gas composition, and traverse velocity. Dense claddings of several millimeters thickness and various W content were achieved. Moreover, multilayers with W content gradually varying from 47 to 92% were formed. The structure, compositional profiles, and thermal properties of the claddings were characterized.

Stress analysis of divertor plasma-facing component designs using tungsten particle-reinforced copper composite heat sink

The W-Cu composites are considered as advanced heat sink material for plasma-facing components, depending notably on the reduced macroscopic coefficient of thermal expansion compared with monotonic Cu materials. One class of such W-Cu composite materials is W particle-reinforced Cu composites which can be joined to pure tungsten parts as plasma-facing armor. After heat flux tests with 20 MW/m2 and with a target of 500 cycles, it is found that fracture occurred after around 100 cycles either on the side of W-armor or on the interface between W-armor and W-Cu composite heat sink. In this work, the stress as well as temperature processes during the cyclic heat flux loadings have been investigated with finite-element method for two designs with W-Cu composite materials, and compared with the former design with W-armor and CuCrZr-heat sink. Possible causes for the fracture have been proposed.

Hardening and Strain Localisation in Helium-Ion-Implanted Tungsten

Tungsten is the main candidate material for plasma-facing armour components in future fusion reactors. In-service, fusion neutron irradiation creates lattice defects through collision cascades. Helium, injected from plasma, aggravates damage by increasing defect retention. Both can be mimicked using helium-ion-implantation. In a recent study on 3000 appm helium-implanted tungsten (W-3000He), we hypothesized helium-induced irradiation hardening, followed by softening during deformation. The hypothesis was founded on observations of large increase in hardness, substantial pile-up and slip-step formation around nano-indents and Laue diffraction measurements of localised deformation underlying indents. Here we test this hypothesis by implementing it in a crystal plasticity finite element (CPFE) formulation, simulating nano-indentation in W-3000He at 300 K. The model considers thermally-activated dislocation glide through helium-defect obstacles, whose barrier strength is derived as a function of defect concentration and morphology. Only one fitting parameter is used for the simulated helium-implanted tungsten; defect removal rate. The simulation captures the localised large pile-up remarkably well and predicts confined fields of lattice distortions and geometrically necessary dislocation underlying indents which agree quantitatively with previous Laue measurements. Strain localisation is further confirmed through high resolution electron backscatter diffraction and transmission electron microscopy measurements on cross-section lift-outs from centre of nano-indents in W-3000He.

The Role of Laser Texturing in Improving the Adhesion of Plasma Sprayed Tungsten Coatings

Plasma sprayed W-based coatings have the potential application as joining interlayers for the plasma-facing components of future fusion reactors. Particularly in the form of tungsten-steel functionally graded materials (FGMs), they would reduce the stress concentration between the W-based plasma-facing armor and steel-based construction. For mechanical integrity, good adhesion is essential. For this, conventional grit blasting treatment of the substrate may not be sufficient. Therefore, alternative treatments capable of reaching significantly higher roughness without damaging the substrate are sought. In this study, the effects of laser texturing on the adhesion of plasma sprayed W and W-steel FGMs on W and steel substrates are investigated. A variety of surface profiles were achieved by controlled laser ablation in several types of patterns. Their geometric features were systematically varied and measured, and the extent of their filling by the coating material was evaluated. On the most promising patterns, coating adhesion tests were performed in shear loading mode and the best performing patterns were identified. Significant improvement over grit-blasted surface was obtained on patterned surfaces with a high effective contact area, promoting mechanical interlocking. On steel substrates, generally higher adhesion was observed, surpassing already the coating cohesive strength; this is attributed to a combination of mechanical interlocking and metallurgical bonding.

Hardening and Strain Localisation in Helium-Ion-Implanted Tungsten

Tungsten is the main candidate material for plasma-facing armour components in future fusion reactors. In service, bombardment with energetic neutrons will create collision cascades that leave behind lattice defects. Helium, injected from the plasma and produced by transmutation, strongly interacts with these defects, modifying their behaviour and retention. Helium-ion-implantation provides an effective tool for examining helium-defect interactions and their effect on the properties of tungsten armour. We use nano-indentation to probe the mechanical properties of the shallow helium-ion-implanted layer. Comparison of spherical indents in unimplanted and helium-implanted regions of the same single crystal shows a large increase in hardness and substantial pile-up in the implanted material. The complex lattice distortions beneath indents are probed non-destructively using 3D-resolved synchrotron X-ray micro-diffraction. Reduced lattice rotations and indentation-induced residual strains in the ion-implanted material indicate a more confined plastic zone. This is confirmed by HR-EBSD and TEM observations. Together our results suggest that dislocation motion is initially obstructed by helium-induced defects. The obstacle strength of these defects is reduced by the passage of dislocations, leading to a strain-softening and slip channel formation. A constitutive law for 3D crystal plasticity finite element simulations is developed based on this hypothesis. Simulations of the indentation process successfully capture the helium-implantation induced changes in deformation behaviour observed in experiments. Importantly, the effects we observe are markedly different from previous observations for self-ion-implanted tungsten, highlighting that the exact nature of the implantation damage plays a critical role in determining mechanical property change.

Helium-implantation-induced lattice strains and defects in tungsten probed by X-ray micro-diffraction

Tungsten is the main candidate material for plasma-facing armour components in future fusion reactors. Bombardment with energetic fusion neutrons causes collision cascade damage and defect formation. Interaction of defects with helium, produced by transmutation and injected from the plasma, modifies defect retention and behaviour. Here we investigate the residual lattice strains caused by different doses of helium-ion-implantation into tungsten and tungsten‑rhenium alloys. Energy and depth-resolved synchrotron X-ray micro-diffraction uniquely permits the measurement of lattice strain with sub-micron 3D spatial resolution and ~10−4 strain sensitivity. Increase of helium dose from 300 appm to 3000 appm increases volumetric strain by only ~2.4 times, indicating that defect retention per injected helium is ~3 times higher at low helium doses. This suggests defect retention is not a simple function of implanted helium dose, but strongly depends on material composition and presence of impurities. Conversely, analysis of W-1 wt% Re alloy samples and of different crystal orientations shows that both the presence of rhenium, and crystal orientation, have a comparatively small effect on defect retention. These insights are key for the design of armour components in future reactors where it will be essential to account for irradiation-induced dimensional change when predicting component lifetime and performance.

On Oxidation Resistance Mechanisms at 1273 K of Tungsten-Based Alloys Containing Chromium and Yttria

Tungsten (W) is currently deemed the main candidate for the plasma-facing armor material of the first wall of future fusion reactors, such as DEMO. Advantages of W include a high melting point, high thermal conductivity, low tritium retention, and low erosion yield. However, was an accident to occur, air ingress into the vacuum vessel could occur and the temperature of the first wall could reach 1200K to 1450K due to nuclear decay heat. In the absence of cooling, the temperature remains in that range for several weeks. At these temperatures, the radioactive tungsten oxidizes and then volatilizes. Smart W alloys are therefore being developed. Smart alloys are supposed to preserve properties of W during plasma operation while suppressing tungsten oxide formation in case of an accident. This study focuses on investigations of thin film smart alloys produced by magnetron sputtering. These alloys provide an idealistic system with a homogeneous distribution of the elements W, chromium (Cr), and yttrium (Y) on an atomic scale. The recommended composition is W with 12 weight % of Cr and 0.5 weight % of Y. Passivation and a suppression of WO3 sublimation is shown. For the first time, the mechanisms yielding the improved oxidation resistance are analyzed in detail. A protective Cr2O3 layer forms at the surface. The different stages of the oxidation processes up to the failure of the protective function are analyzed for the first time. Using 18O as a tracer, it is shown for the first time that the oxide growth occurs at the surface of the protective oxide. The Cr is continuously replenished from the bulk of the sample, including the Cr-rich phase which forms during exposure at 1273K. A homogenous distribution of yttria within the W-matrix, which is preserved during oxidation, is a peculiarity of the analyzed alloy. Further, an Y-enriched nucleation site is found at the interface between metal and oxide. This nucleation sites are deemed to be crucial for the improved oxidation resistance.

Heat loading behavior and thermomechanical analyses on plasma spray tungsten coated reduced-activation ferritic/martensitic steel

Tungsten coating with a thickness of 1 mm on reduced-activation ferritic/martensitic steel (RAF/M) F82H (Fe-8Cr-2W), which is a leading structural material candidate for DEMO, have been produced by Atmospheric Plasma Spraying (APS) and Vacuum Plasma Spraying (VPS). Heat loading experiments on thermal response and fatigue using an electron beam have been carried out on W-coated F82H brazed on oxygen free high purity copper (OFHC) block with a cooling tube to evaluate their possibility as a plasma-facing armor in the fusion device. In addition, quantitative analyses about temperature profiles and thermal stress have been carried out using FEA. Thermal response experiments show that the temperatures increased monotonically with increasing heat flux. Surface temperature of the VPS-W/F82H/OFHC is always lower than that of the APS-W/F82H/OFHC. Surface modification, exfoliation and crack are not formed by thermal fatigue experiments up to 200 cycles at a heat flux of 3.2 MW/m2. Thermal response experiments under steady state condition have been successfully modeled by FEA. Interfacial strength of VPS-W and F82H is also discussed using the FEA results.

Oxidation resistance of bulk plasma-facing tungsten alloys

Tungsten (W) currently is the main candidate as plasma-facing armour material for the first wall of future fusion reactors, like DEMO. Advantages of W include a high melting point, high thermal conductivity, low tritium retention, and low erosion yield. However, in case of an accident, air ingress into the vacuum vessel can occur and the temperature of the first wall can reach 1200 K to 1450 K due to nuclear decay heat. In the absence of cooling, the temperature will remain in that range for several weeks. At these temperatures the radioactive tungsten oxide volatilizes. Therefore, ‘smart’ W alloys are developed that aim to preserve the properties of W during plasma operation coupled with suppressed tungsten oxide formation in case of an accident.This study focusses on oxidation studies at 1273 K of samples produced by mechanical alloying followed by field assisted sintering. In a first step the sintering is optimized for tungsten (W) – chromium (Cr) -yttrium (Y) alloys. It is shown that the best oxidation resistance is achieved with submicron grain sizes. This yields a closed, protective oxide layer. In a second step the influence of the grinding process during sample preparation is analysed. It is shown that scratches initiate failure of the protective oxide. In a third step the oxidation and sublimation is measured for weeks – for the first time the sublimation is directly measured in order to determine the potential hazard in comparison to pure W. It is shown that the oxidation is suppressed in comparison to pure W. However, sublimation at a rate of 1×10−6mgcm−2s−1 starts after a few days. Nevertheless, the progess in smart alloys is evident: sublimation is delayed by about two days and complete mechanical destruction of the first wall is avoided.

Deuterium Removal from W-Doped Carbon Films by Thermo-Oxidation

If both carbon and tungsten were to be part of the plasma-facing armor in a future fusion reactor, it is inevitable that carbon co-deposits containing tungsten impurities will form. This work examines the effectiveness of thermo-oxidation in removing hydrogen from W-containing carbon co-deposits. Amorphous deuterated hydrocarbon (a-C:D) films were created with a CD4/Ar direct-current glow discharge and doped with W sputtered from a W mesh in front of the specimen. The W concentration in the specimens ranged from 0 to 35 at. % W/(W + C). The films were oxidized at 350°C, in 2 Torr pure O2 for time increments totaling 8 h. The D content of the films was measured before and at various stages of the oxidation exposure using laser thermal desorption spectroscopy. Essentially all deuterium was removed from films containing very little or no W doping [<0.1% W/(W + C)]. For films with more W [few percent W/(W + C)], oxidation was less effective at removing D. For two specimens with 2.4% and 35% W/(W + C), oxidation was completely ineffective at removing D.

Effect of processing factors on the microstructure and gradual diffusion of tungstenized layers

The refractory tungsten (W) coating due to its low sputtering yield is not easy to be obtained, and the bonds between the W coating and its substrate are weak and easily broken. A novel gradual tungstenized layer as the plasma facing armor material is prepared by using the double glow plasma surface alloying technology (DGPA). The effects of process factors on the microstructure and micro-hardness of the alloyed layers are investigated. It is found that TixW1-x gradual layers are formed in the alloyed layer. Due to the fact that the tungsten content in the gradual tungstenized layer gradually decreases from surface to the substrate, the layer is composed of both pure tungsten coating and transitional coatings (e.g. TixW1-x (x>0.5) and TixW1-x (x<0.5) coatings). Hence, the gradual layer can enhance the bond strength between the layer and substrate, and the improved mechanical properties such as high hardness can also be obtained. The fact is that vacancies, dislocation and crystal defects could form on the surface of the substrate due to the effect of the double glow plasma bombardment. Therefore, a particular attention is paid to the diffusion mechanisms of W atoms in the W gradual layers and a diffusion model is proposed to explain the forming process of the gradual layers according to the Fick's first law. The findings reported here demonstrate the potential of an effective surface modification method for improving the diffusivity of W atoms. The present work provides a better understanding of the gradual diffusion process of tungsten element.

Microstructural stability of spark-plasma-sintered Wf/W composite with zirconia interface coating under high-heat-flux hydrogen beam irradiation

Tungsten is considered as the most suitable material for the plasma-facing armour of future fusion reactors. However, in spite of many advantageous properties, pure tungsten has a major drawback, namely, brittleness at lower temperatures and embrittlement by neutron irradiation. Tungsten fibre-reinforced tungsten (Wf/W) composites are thought to be a promising candidate material for armour owing to the pseudo-toughness effect which is based on controlled cracking of coated interfaces. In this material concept, the reliability of the material during service relies on the fabrication quality as well as the stability of microstructure, particularly, of the interfacial coating under high-heat-flux loads.In this paper, the durability and chemical stability of Wf/W composite specimens under cyclic heat-flux loads up to 20 MW/m² (surface temperature: 1260 °C) was investigated using hydrogen neutral beam. The bulk material was fabricated by means of spark-plasma-sintering (SPS) method using fine tungsten powder and a stack of tungsten wire meshes as reinforcement where the surface of the wire was coated with zirconia thin film to produce an engineered interface. The impact of plasma beam irradiation on microstructure was examined for two kinds of specimens produced at different sintering temperatures, 1400 °C and 1700 °C. Results of microscopic (SEM) and chemical (EDX) analysis are presented comparing the microstructure and element distribution maps obtained before and after heat flux loading. Effects of different sintering temperatures on damage behaviour are discussed. The present composite materials are shown to be applicable as plasma-facing material for high-heat-flux components.

Transmutation of Tungsten in Fusion and Fission Nuclear Environments

The amount and type of gaseous and metallic transmutants produced in tungsten (W) when used as a plasma-facing armor in magnetic (MFE) and inertial (IFE) confinement fusion systems were determined and compared to those obtained following irradiation in fission reactors. Up to ∽8% metallic transmutants are generated at the expected lifetime of the fusion blanket. Irradiation in fission reactors to the same fast neutron fluence yields a much larger amount of metallic transmutation products than in fusion systems. While the dominant component in fusion systems is rhenium (Re), osmium (Os) is the main transmutation product in fission reactors. The impact on the W properties needs to be assessed. The results of this work will help guide irradiation experiments in fission reactors to properly simulate the conditions in fusion systems by possible direct implantation of transmutation products in irradiated samples. In addition, the results represent a necessary input for modeling activities aimed at understanding the expected effects on properties.

Electrodeposition of metallic tungsten coating from binary oxide molten salt on low activation steel substrate

Tungsten is considered a promising plasma facing armor material for future fusion devices. An electrodeposited metallic tungsten coating from Na2WO4–WO3 binary oxide molten salt on low activation steel (LAS) substrate was investigated in this paper. Tungsten coatings were deposited under various pulsed currents conditions at 1173K in atmosphere. Cathodic current density and pulsed duty cycle were investigated for pulsed current electrolysis. The crystal structure and microstructure of tungsten coatings were characterized by X-ray diffractometry, scanning electron microscopy, and energy X-ray dispersive analysis techniques. The results indicated that pulsed current density and duty cycle significantly influence tungsten nucleation and electro-crystallization phenomena. The average grain size of the coating becomes much larger with increasing cathodic current density, which demonstrates that appropriate high cathodic current density can accelerate the growth of grains on the surface of the substrate. The micro-hardness of tungsten coatings increases with the increasing thickness of coatings; the maximum micro-hardness is 482HV. The prepared tungsten coatings have a smooth surface, a porosity of less than 1%, and an oxygen content of 0.024wt%.

The JET ITER-like wall experiment: First results and lessons for ITER

The JET programme is strongly focused on preparations for ITER construction and exploitation. To this end, a major programme of machine enhancements has recently been completed, including a new ITER-like wall, in which the plasma-facing armour in the main vacuum chamber is beryllium while that in the divertor is tungsten—the same combination of plasma-facing materials foreseen for ITER. The goal of the initial experimental campaigns is to fully characterise operation with the new wall, concentrating in particular on plasma-material interactions, and to make direct comparisons of plasma performance with the previous, carbon wall. This is being done in a progressive manner, with the input power and plasma performance being increased in combination with the commissioning of a comprehensive new real-time protection system. Progress achieved during the first set of experimental campaigns with the new wall, which took place from September 2011 to July 2012, is reported.