Divertor Materials Research Articles

Tuning mechanical properties of ultrafine-grained tungsten by manipulating grain boundary chemistry

Tungsten is, due to a combination of high strength and good physical properties, frequently considered for high-performance applications in the harshest environments. Oftentimes its inherent brittleness and low ductility stand in the way of a successful deployment in these fields. Since tungsten has been proposed as divertor material for nuclear fusion reactors, an improvement of ductility and fracture toughness is essential. An obvious first step to increase these properties is to reduce the grain size to the ultrafine-grained regime. As this still leaves the material with a relatively low-energy intercrystalline fracture mode, this work takes a step further. With the help of doping elements, which are identified from ab-initio simulations, an attempt to increase grain boundary cohesion of ultra-fine grained tungsten to improve ductility is made. After fabrication of the doped samples from powders using severe plastic deformation, thorough microstructural investigations and extensive mechanical characterization, utilizing various small-scale testing techniques, are combined to assess the properties of the materials. We report that the addition of boron and hafnium can significantly increase the bending strength and bending ductility of ultra-fine grained tungsten. An additional heat treatment of the boron doped sample amplifies this effect even further, drastically increasing the strength and overall mechanical properties due to a combination of hardening-by-annealing and increased grain boundary segregation. Thus, an effective way to adaptively improve the mechanical properties of tungsten by manipulating grain boundary chemistry is reported, validating grain boundary segregation engineering as a powerful tool for enhancing damage tolerance in brittle materials.

Microstructure deformation and near-pore environment of resolidified tungsten in high heat flux conditions

Tungsten is a candidate divertor material for many tokamak reactors and has potential to be durable to high heat flux conditions. However, plasma-materials interaction and edge localized modes (1–10 GW/m2) can cause significant hardening and melting damage. Tungsten in a high heat flux (46.3 GW/m2) and helium plasma environment was investigated. We studied possible residual helium and microstructure deformation in resolidified tungsten. Following this, a 1–30 keV Ga focused ion beam was used for TEM sample milling. In as-received tungsten, dark spots of possible lattice strain and defects were in the grains. After high heat exposure under plasma pressure, intergranular features, dark spots in TEM, diffraction patterns, and elongated FIB induced pore-artifact structures (second phase) formed. Helium appears to be absent from the resolidified tungsten matrix due to erosive melting damage and high temperature conditions. A simple four-step microstructure deformation process from the elastic bulk side to the near-pore environment is proposed. The presence of defects in TEM images, grain size reduction, and microstructure deformation affected hardening. The near-pore environment likely experienced grain refinement, and further disordered nanophases are possible under severe deformation conditions.

Non-linear MHD modelling of edge localized modes suppression by resonant magnetic perturbations in ITER

Edge localized modes (ELMs) suppression by resonant magnetic perturbations (RMPs) was studied with the non-linear magneto-hydro-dynamic (MHD) code JOREK for the ITER H-mode scenarios at 15 MA, 12.5 MA, 10 MA/5.3 T. The main aim of this work was to demonstrate that ELMs can be suppressed by RMPs while the divertor 3D footprints of heat and particle fluxes remain within divertor material limits. The unstable peeling–ballooning modes responsible for ELMs without RMPs were modelled first for each scenario using numerically accessible parameters for ITER. Then the stabilization of ELMs by RMPs was modelled with the same parameters. RMP spectra, optimized by the linear MHD MARS-F code, with main toroidal harmonics N = 2, N = 3, N = 4 have been used as boundary conditions of the computational domain of JOREK, including realistic RMP coils, main plasma, scrape off layer (SOL) divertor and realistic first wall. The model includes all relevant plasma flows: toroidal rotation, two fluid diamagnetic effects and neoclassical poloidal friction. With RMPs, the main toroidal harmonic and the non-linearly coupled harmonics remain dominant at the plasma edge, producing saturated modes and a continuous MHD turbulent transport thereby avoiding ELM crashes in all scenarios considered here. The threshold for ELM suppression was found at a maximum RMP coils current of 45 kAt–60 kAt compared to the coils maximum capability of 90 kAt. In the high beta poloidal steady-state 10 MA/5.3 T scenario, a rotating QH-mode without ELMs was observed even without RMPs. In this scenario with RMPs N = 3, N = 4 at 20 kAt maximum current in RMP coils, similar QH-mode behaviour was observed however with dominant edge harmonic corresponding to the main toroidal number of RMPs. The present MHD modelling was limited in time by few tens of ms after RMPs were switched on until the magnetic energy of the modes saturates. As a consequence the thermal energy was still evolving on this time scale, far from the ITER confinement time scale and hence only the form of 3D footprints on the divertor targets can be indicated within this set-up. Also note, that the divertor physics was missing in this model, so realistic values of fluxes are out of reach in this modelling. However the stationary 3D divertor and particle fluxes could be simply extrapolated from these results to the stationary situation considering that a large power fraction should be radiated in the core and SOL and only about 50 MW power is going to the divertor, which is an arbitrary, but reasonable number used here. The 3D footprints with RMPs show the characteristic splitting with the main RMP toroidal symmetry. The maximum radial extension of the footprints typically was ∼20 cm in inner divertor and ∼40 cm in outer divertor with stationary heat fluxes decreasing further out from the initial strike point from ∼5 MW m−2 to ∼1 MW m−2 assuming a total power in the divertor and walls is 50 MW. The heat fluxes remain within the divertor target and baffle areas, however with rather small margin in the outer divertor which could be an issue for the first wall especially in transient regimes when part of the plasma thermal energy is released due to switching on the RMP coils. This fact should be considered when RMPs are applied with a more favorable application before or soon after the L–H transition, although optimization is required to avoid increasing the L–H power threshold with RMPs.

Determination of the characteristic magnetic pre-sheath length at divertor surfaces using micro-engineered targets on DiMES at DIII-D

The magnetic pre-sheath (MPS) length, L MPS, is a critical parameter to define the sheath potential, which controls the ion trajectory of low-Z species (D, T, He, and C), as well as the prompt re-deposition of high-Z species. To determine L MPS, we fabricated micro-trenches (30 × 30 × 4 μm) via focused ion beam milling on a silicon surface and exposed them to L-mode deuterium plasmas in DIII-D via the divertor material evaluation system (DiMES) removable sample exposure probe. The areal distribution of impurity depositions, mainly consisting of carbon, was measured by energy-dispersive x-ray spectroscopy (EDS) to reveal the deuterium ion shadowing effect on the trench floors. The carbon deposition profiles showed that the erosion was maximized for the azimuthal direction of φ = −40° (referenced to the toroidal magnetic field direction) as well as the polar angle of θ = 80°. A Monte Carlo equation-of-motion (EOM) model, based on a collisionless MPS, was used to calculate the azimuthal and polar deuterium ion angle distributions (IADs) at the surface for a range of L MPS = k × ρ i, where ρ i is the ion gyro radius and k = 0.5–4. Then, gross erosion profiles were calculated by a Monte Carlo micro-patterning and roughness (MPR) code for ion sputtering using as input the calculated azimuthal and polar IADs for each value of k. Good agreement with the experimental C deposition profiles was obtained for the case k = 2.5–3.5. This result is consistent with a previous kinetic modeling prediction of k ∼ 3, as well as previous analytical investigations that predicted the L MPS to be several ion gyro radii. A validation of theoretical sheath models supports its applicability to ITER and pilot plant divertors to successfully predict plasma–materials interactions.

New proposed ITER divertor design using carbon insert on tungsten to mitigate ELMs and secondary radiation effects on nearby components

Building a successful device for the magnetic fusion energy production is a great challenge. ITER is an international project of the tokamak based magnetic fusion design being developed for the demonstration of the feasibility of thermonuclear technologies for future realization of successful commercial fusion energy. A key obstacle to a successful magnetic fusion energy production is however, the performance during abnormal events including plasma disruptions and edge-localized modes (ELMs). A credible reactor design must tolerate at least a few of these transient events without serious consequences such as melting of the structure. This paper investigates and compares the performance of the current ITER tokamak design during two types of transient events, i.e., ELMs occurring at normal operation and disruptions during abnormal operation. We simulated the divertor components response using our integrated 3D HEIGHTS package. The simulations include self-consistent modeling of the interaction of the released core plasma particles with the initial solid divertor material, energy deposition processes, vaporization of divertor material, secondary plasma formation and MHD evolution, incident core particles collisions and scattering from this dense secondary plasma, photon radiation of secondary plasma, and the resulting heat loads on nearby components. Our simulations showed that using a small carbon insert around the strike point can significantly reduce the overall expected damage on the tungsten dome structure, reflector plates, and prevent tungsten vaporization and its potential core plasma contamination.

Helium-induced swelling and mechanical property degradation in ultrafine-grained W and W-Cu nanocomposites for fusion applications

Besides high dose radiation and extreme thermal loads, a major concern for materials deployed in novel nuclear fusion reactors is the formation and growth of helium bubbles. This work investigates the swelling and mechanical property degradation after helium implantation of ultrafine-grained W and nanocrystalline W-Cu, possible candidates for divertor and heat-sink materials in fusion reactors, respectively. It is found that ultrafine-grained W and single crystalline W experience similar volumetric swelling after helium implantation but show different blistering behavior. The W-Cu nanocomposite, however, shows a reduced swelling compared to a coarse-grained composite due to the effective annihilation of radiation-induced vacancies through interfaces. Furthermore, the helium-filled cavity structures lead to considerable softening of the composite.

Dynamic interaction between the dislocations and cavities in tungsten during tensile test

The irradiation hardening factor of the cavities (helium bubbles) in tungsten, as the divertor candidate material, in nuclear fusion reactor, has been investigated through an in-situ TEM observation during tensile test. The obstacle barrier strength α has been calculated by the bow-out behavior of a dislocation based on the line tension model. The obstacle barrier strength α of the cavity with a diameter of 2 nm in W has been measured to be 0.53 ± 0.02. The fraction of cross-slip behavior except the obstacle-cutting mechanism owing to the dynamic interaction between a dislocation of screw type and a cavity has been about 11 %. The fraction of cross-slip behavior of the screw dislocation in W has been smaller than those of other BCC structure materials.

Nitrogen retention and ammonia production on tungsten

We report a systematic study that quantifies nitrogen retention and ammonia production on tungsten and that sheds light on the mechanism for ammonia formation on ITER’s divertor material. Saturation of the nitrogen-implanted layer in polycrystalline tungsten is observed at room temperature for a nitrogen ion fluence in the low 1021 N+ m−2 range. Nitrogen desorption from this N-implanted layer occurs in the 800–1100 K temperature range and exhibits a zero-order kinetics with an activation energy of 1.45 eV and a prefactor of 5 × 1024 m−2 s−1. Following nitrogen and deuterium co-implantation, deuterated ammonia production is observed during temperature programmed desorption between 350 K and 650 K in conjunction with deuterium desorption. In contrast, nitrogen desorption still occurs above 800 K. Significant production of ammonia is obtained only when the nitrogen layer created by ion implantation is approaching saturation and the amount of nitrogen lost to ammonia production is only in the percent range. This result is understood by repeating cycles of deuterium implantation and thermo-desorption below the desorption temperature of the nitrogen layer. The exponential decay of the amount of produced ammonia with cycle number demonstrates that nitrogen diffusion to the surface is negligible in the ammonia production temperature range and that ammonia formation occurs at the outermost surface layer. The maximum quantity of ammonia produced from the present N implanted layer is below 2 × 1018 ND3 m−2, which is limited by the nitrogen atom surface density. Surface vibrational spectroscopy demonstrates the presence of ammonia precursors on the nitrogen-implanted tungsten surface upon deuterium implantation. These ammonia precursors can be created also at room temperature through the dissociative chemisorption of thermal D2 catalysed by nitrogen present at the tungsten surface and, more efficiently, by adsorption of deuterium atoms.

Behavior of high-entropy W-rich alloys Wx(TaVCrTi)y under He+ irradiation

A group of high-entropy W-rich (HE-W) alloys were investigated as potential divertor materials in this study. Pure W and four HE-W alloys: W90(TaVCrTi)10, W80(TaVCrTi)20, W70(TaVCrTi)30, and W60(TaVCrTi)40 were manufactured by mechanical alloying and spark plasma sintering. The HE-W alloys showed the body-centered cubic structure with W-based and Ti-rich grains. The HE-W alloys had the ultrafine grain size, which were much finer than that of pure W. The HE-W alloys exhibited excellent surface stability under He+ irradiation. Even under an irradiation fluence of 0.9 × 1025 ions/m2, the surface of the HE-W alloys remained nearly flat. The excellent He+ surface stability of the HE-W alloys was due to their self-healing mechanism and the effect of the irradiation temperature on the irradiation damage was not significant.

Combined high fluence and high cycle number transient loading of ITER-like monoblocks in Magnum-PSI

It is highly desirable to understand the long term evolution of the divertor material under the extreme steady-state and transient heat and particle loads expected during ITER operation. Here the impact of ELM-like transient loading under combined high-flux plasma and transient ELM-like heat loading in Magnum-PSI was explored to determine how plasma affects the fatigue cracking threshold of tungsten due to ELMs. Mock-ups consisting of five ITER-like monoblocks in a chain were simultaneously exposed to high flux plasma and a high power pulsed laser which closely simulated the ELM impact in terms of heat flux and duration. Loading conditions were chosen to enable comparison to existing data from electron-beam loading, while the influence of surface base temperature (750 °C, 1150 °C or 1500 °C) and impurity seeding (addition of 6.5 ion% He+ and/or 8 ion% Ne+) were also investigated. The plasma loading leads to differences in surface morphology and indicates synergistic effects on the extent of the surface damage. Base temperatures at or above 1150 °C are found to lead to a significant reduction in the fatigue cracking threshold by a factor of two or more compared to at 750 °C. Cracked surfaces are found to be more than ten times rougher than the original microstructure, and additionally when seeding impurities are added surface roughness can be significantly increased by up closely factor of two compared to roughening using pure H plasma. Overall the results indicate that avoiding fatigue cracking in ITER will be very challenging, and that understanding the level to which this can therefore be tolerated is vital for anticipating divertor lifetime and reliability.

Thermal-hydraulic modelling and analysis of ITER tungsten divertor monoblock

• A mathematical model has been developed to investigate the thermal-hydraulic performance of ITER tungsten divertor monoblock. • The model predicts the thermal response of the divertor structural materials and coolant tube. • The model accounts for the melting, vaporization, and solidification of the upper layer of the divertor facing plasma. • The model is verified against a previous calculation in the literature for DEMO divertor at an incident surface heat flux of 10 MW/m 2 . • The model is used to simulate the divertor materials response subjected to a VDE. The divertor is a fundamental component of fusion power plants, being primarily responsible for power exhaust and impurity removal via guided plasma exhaust. Due to its position and functions, the divertor has to sustain very high heat flux arising from the plasma (up to 20 MW/m 2 ), while experiencing an intense nuclear deposited power, which could jeopardize its structure and limit its lifetime. Therefore, attention has to be paid to the thermal–hydraulic design of its cooling system. In this work a mathematical model has been developed to investigate the steady state and transient thermal–hydraulic performance of ITER tungsten divertor monoblock. The model predicts the thermal response of the divertor structural materials and coolant tube. The coolant tube is divided into specified axial regions and the divertor plate is divided into specified radial zones, and then a two-dimensional heat conduction calculation is performed to predict the temperature distribution for both steady and transient states. A two-dimensional numerical finite difference technique is adapted in Cartesian coordinate system where the implicit scheme is used for transient calculation. The model also accounts for the melting, vaporization, and re-solidification of the upper layer of the divertor facing plasma. The selected heat transfer correlations cover all possible operating conditions of ITER under both normal and off-normal situations. The model is verified against a previous calculation in the literature for DEMO divertor at an incident surface heat flux of 10 MW/m 2 . The model is then used to predict the steady state thermal behaviour of the divertor under incident surface heat fluxes ranges from 2 to 20 MW/m 2 for a bare cooling tube and cooling tube with swirl-tape insertion. It calculates the maximum tube surface heat flux and the minimum critical heat flux ratio for all cases as well. The model is also used to simulate the divertor materials response subjected to high heat flux during a vertical displacement event (VDE) where 60 MJ/m 2 plasma energy is deposited over 500 ms.

The updated ITPA global H-mode confinement database: description and analysis

The multi-machine International Tokamak Physics Activity (ITPA) Global H-mode Confinement Database has been upgraded with new data from JET with the ITER-like wall and ASDEX Upgrade with the full tungsten wall. This paper describes the new database and presents results of regression analysis to estimate the global energy confinement scaling in H-mode plasmas using a standard power law. Various subsets of the database are considered, focusing on type of wall and divertor materials, confinement regime (all H-modes, ELMy H or ELM-free) and ITER-like constraints. Apart from ordinary least squares (OLS), two other, robust regression techniques are applied, which take into account uncertainty on all variables. Regression on data from individual devices shows that, generally, the confinement dependence on density and the power degradation are weakest in the fully metallic devices. Using the multi-machine scalings, predictions are made of the confinement time in a standard ELMy H-mode scenario in ITER. The uncertainty on the scaling parameters is discussed with a view to practically useful error bars on the parameters and predictions. One of the derived scalings for ELMy H-modes on an ITER-like subset is studied in particular and compared to the IPB98(y, 2) confinement scaling in engineering and dimensionless form. Transformation of this new scaling from engineering variables to dimensionless quantities is shown to result in large error bars on the dimensionless scaling. Regression analysis in the space of dimensionless variables is therefore proposed as an alternative, yielding acceptable estimates for the dimensionless scaling. The new scaling, which is dimensionally correct within the uncertainties, suggests that some dependencies of confinement in the multi-machine database can be reconciled with parameter scans in individual devices. This includes vanishingly small dependence of confinement on line-averaged density and normalized plasma pressure (β), as well as a noticeable, positive dependence on effective atomic mass and plasma triangularity. Extrapolation of this scaling to ITER yields a somewhat lower confinement time compared to the IPB98(y, 2) prediction, possibly related to the considerably weaker dependence on major radius in the new scaling (slightly above linear). Further studies are needed to compare more flexible regression models with the power law used here. In addition, data from more devices concerning possible ‘hidden variables’ could help to determine their influence on confinement, while adding data in sparsely populated areas of the parameter space may contribute to further disentangling some of the global confinement dependencies in tokamak plasmas.

Bubble formation in liquid Sn under different plasma loading conditions leading to droplet ejection

Liquid metals have been proposed as potential divertor materials for future fusion reactors, and surface stability is a vital requirement for such liquid metal divertors (LMDs). Capillary porous structures (CPSs) have been applied to the design of liquid metal targets as they can avoid MHD instability by surface tension and provide a stable liquid surface. However, our previous work has found that liquid Sn surfaces can be very unstable in hydrogen plasma even in cases without magnetic fields. To increase our understanding of the interaction of liquid Sn surfaces with plasmas, in this work we systematically investigated the surface behaviors of liquid Sn in different plasma exposures in linear plasma devices, either in Nano-PSI at low flux and without magnetic field, or in Magnum-PSI with strong magnetic field strength. Surface instability leading to droplet ejection has been observed and recorded in the experiments. The ejection of droplets is not dependent on magnetic fields and plasma currents, and is found to be dependent on the plasma species and plasma flux and surface temperature. The CPS meshes applied in the experiments cannot completely avoid droplet ejection but can decrease droplet size and lower droplet production rate. In H plasma, droplets were observed once Sn melted even at low fluxes. For the case of N plasma, the appearance of droplets started at a temperature marginally higher than tin–nitride decomposition temperature. Only at high fluxes (∼1023–24 m−2 s−1) and high temperatures (900–1000 °C) were a few droplets observed in Ar or He plasma. For all cases, the ejection velocities of most droplets were around 1–5 m s−1. Bubble formation, growth and bursting in the plasma-species-supersaturated liquid Sn is proposed as the primary mechanism for the ejection of droplets. Plasma-enhanced solubility is responsible for the achievement of H/N-supersaturated liquid Sn, while high plasma flux implantation is responsible for Ar/He-supersaturated liquid Sn. Once the concentration of plasma species in liquid Sn reaches a certain supersaturation level, nucleation and growth of bubbles occur due to the desorption of dissolved plasma species from the liquid Sn. The formation and bursting of bubbles have been directly observed in the experiment. The sizes of most bubbles were estimated in the range of 40–400 μm or even smaller. A bubble growth model based on Sievert’s and Henry’s laws is invoked to describe bubble growth in liquid Sn.

Evaluation of silicon carbide as a divertor armor material in DIII-D H-mode discharges

Silicon carbide (SiC) represents a promising but largely untested plasma-facing material (PFM) for next-step fusion devices. In this work, an analytic mixed-material erosion model is developed by calculating the physical (via SDTrimSP) and chemical (via empirical scalings) sputtering yield from SiC, Si, and C. The Si content in the near-surface SiC layer is predicted to increase during D plasma bombardment due to more efficient physical and chemical sputtering of C relative to Si. Silicon erosion from SiC thereby occurs primarily from sputtering of the enriched Si layer, rather than directly from the SiC itself. SiC coatings on ATJ graphite, manufactured via chemical vapor deposition, were exposed to repeated H-mode plasma discharges in the DIII-D tokamak to test this model. The qualitative trends from analytic modeling are reproduced by the experimental measurements, obtained via spectroscopic inference using the S/XB method. Quantitatively the model slightly under-predicts measured erosion rates, which is attributed to uncertainties in the ion impact angle distribution, as well as the effect of edge-localized modes. After exposure, minimal changes to the macroscopic or microscopic surface morphology of the SiC coatings were observed. Compositional analysis reveals Si enrichment of about 10%, in line with expectations from the erosion model. Extrapolating to a DEMO-type device, an order-of-magnitude decrease in impurity sourcing, and up to a factor of 2 decrease in impurity radiation, is expected with SiC walls, relative to graphite, if low C plasma impurity content can be achieved. These favorable erosion properties motivate further investigations of SiC as a low-Z, non-metallic PFM.

Assessment of ITER divertor performance during early operation phases

During the ITER design phase, the focus of ITER boundary plasma modeling activities has been on divertor performance under baseline H-mode, fusion power operation (FPO) conditions. However, early ITER operation will be primarily with hydrogen fuel in L-mode, in the pre-fusion power operation 1 (PFPO-1) phase. Here, the SOLPS-ITER code is used to evaluate divertor performance during this non-active phase. To verify the assumptions used in the existing high power simulation database, gas throughput scans were performed for two types of divertor surface material (beryllium and tungsten) and two gas puff locations (divertor and main chamber). The adoption of beryllium target surfaces simulates the effect of main chamber material erosion and migration and, along with main chamber gas injection, is the current default for the high power database. Depending on the divertor surface material, the atom to molecule ratio of the recycled neutral particles varies. This modifies the momentum and power loss mechanisms arising from plasma–neutral interactions. However, since the effect of atomic and molecular reactions are compensatory, the ‘total’ power and momentum losses are relatively insensitive to the target surface material. Similarly, the impact of gas puff location on divertor plasma parameters is not significant, though main chamber injection provides an additional ionization source in the upstream scrape-off layer (SOL) and leads to moderate changes in the upstream density and far SOL parameters. However, these effects can be neglected within the available range of the gas puff and pump rates in ITER. Since beryllium and tungsten are materials at both extremes in terms of surface reflection properties, the conclusions may be applicable to other divertor surface materials. An important additional finding of the study is that the insensitivity of upstream density to divertor neutral pressure found in the FPO database is also recovered in these PFPO-1 simulations.

Modeling, analysis, and code/data validation of DIII-D tokamak divertor experiments on ELM and non-ELM plasma tungsten sputtering erosion

We analyzed recent DIII-D tokamak tungsten divertor probe experiments using advanced, coupled, sputter erosion/redeposition, plasma, and surface response code packages. Modeling is done for ELMing H-mode, and L-mode plasmas, impinging on various size tungsten deposits on Divertor Material Evaluation System (DiMES) carbon probes. The simulations compute 3D, full kinetic, sub-gyromotion, impurity sputtering and transport, including changes in tungsten surface composition and response due to mixed deuterium and carbon ions irradiation. Per our analysis, ELM (edge localized mode) plasma sputtering in DIII-D mostly involves free-streaming high energy (∼500–1000 eV) D+ and C+6 ions, with high near-surface plasma density. L-Mode sputtering is due to impurity sputtering (C, W) only, with lower density. All cases show complete redeposition of tungsten on the divertor, with significant redeposition on the tungsten spots themselves, and low self-sputtering. Comparison of ELM plasma gross tungsten erosion simulation results with in-situ spectroscopic data is good, as are code/data comparisons of net erosion using post-exposure Rutherford backscattering (RBS) data for the L-mode probes. The analysis, extrapolated to a full tungsten divertor, implies low net erosion and negligible plasma contamination from sputtering. These results support the use of high-Z plasma facing surfaces in ITER and beyond.

On thermal axisymmetric liquid-metal divertors

Abstract Motivated by liquid-metal divertor concepts, we consider a detailed model of a thin liquid-metal film flowing down a toroidally-symmetric solid substrate, driven by gravity as well as thermal and mechanical interaction with the SOL. In the relevant physical limit where the film is thin relative to the divertor extent, and the induced magnetic field is weaker than the applied field, the governing equations may be approximated asymptotically to yield tractable reduced-order dynamics. The analysis incorporates gravitational, thermocapillary and thermoelectric forces, and reveals several nontrivial findings. Firstly, we find that the influence on the liquid metal flow of the toroidal component of the applied field is negligible compared to that of the poloidal component. This has important ramifications for the behaviour of the liquid metal. Secondly, we identify a “spiral-type flow” due to a flow component in the toroidal direction. This flow has been previously reported in the literature in the isothermal case when using extreme parameters. Our analysis explains why these extreme parameter choices were required to observe the isothermal flow, and reveals that, for typical parameters in the thermal case, the flow becomes significant due to thermoelectric forces. Thirdly, we deduce full dimensional velocity scalings. Finally, we present a case study that employs the reduced-order model to explore the effects of sweeping the plasma strike point up and down the divertor. We simulate fully transient dynamics within the divertor and demonstrate how, in some regions of parameter space, the sweeping might promote dry-out or droplet ejection of the liquid metal. We discuss how the case study can inform both divertor material choice as well as tokamak operating parameters.

Hardening effect of multi-energyW2+-ion irradiation on tungsten–potassium alloy**Project supported by the National Natural Science Foundation of China (Grant Nos. 11975160 and 11775149). One of the authors, Kun Zhang, was supported by the Fundamental Research Funds for the Central Universities, China.

Tungsten is one of the most promising plasma-facing materials (PFMs) to be used in the nuclear fusion reactor as divertor material in the future. In this work, W2+-ions bombardment is used to simulate the neutron irradiation damage to commercial pure tungsten (W) and rolled tungsten–potassium (W–K). The 7 MeV of 3 × 1015 W2+-ions/cm2, 3 MeV of 4.5 × 1014 W2+, and 2 MeV of 3 × 1014 W2+-ions/cm2 are applied at 923 K in sequence to produce a uniform region of 100 nm–400 nm beneath the sample surface with the maximum damage value of 11.5 dpa. Nanoindentation is used to inspect the changes in hardness and elastic modulus after self-ion irradiation. Irradiation hardening occurred in both materials. The irradiation hardening of rolled W–K is affected by two factors: one is the absorption of vacancies and interstitial atoms by potassium bubbles, and the other is the interaction between potassium bubbles and dislocations. Under the condition of 11.5 dpa, the capability of defect absorption can reach a threshold. As a result, dislocations finally dominate the hardening of rolled W–K. Specific features of dislocation loops in W–K are further observed by transmission electron microscopy (TEM) to explain the hardening effect. This work might provide valuable enlightenment for W–K alloy as a promising plasma facing material candidate.

Effect of Ta addition on the fuzz formation of additively manufactured W-based materials

As a divertor plasma-facing material, W will experience high-flux plasma irradiation. In particular, severe surface morphology change like fuzz formation can be induced by He plasma irradiation. In this study, fuzz formation on additively manufactured W and W–Ta was investigated. Rolled W, laser-powder-bed-fused (LPBFed) W and W–Ta were exposed to high flux (∼1023 m−2 s−1) He plasma in the linear plasma generator Magnum-PSI with ion energy 12–13 eV at 1273 K. The mean thickness of the fuzz at grain interiors of rolled W, LPBFed W and W–Ta was measured as 0.37 μm, 0.71 μm and 0.23 μm, respectively. The fuzz suppression in LPBFed W–Ta can be attributed to the synergetic effect of solid-solution, dislocation, and secondary-phase nanoparticles. Abnormally grown fuzz was observed near the pre-existing cracks of LPBFed W, while no such structure was found in LPBFed W–Ta. It is found that dislocations play a crucial role in inhibiting fuzz growth. This is confirmed by the difference in fuzz structure in rolled W and LPBFed W, where rolled W has a much greater dislocation density compared to LPBFed W. This work suggests that the fuzz growth kinetics may be tuned by tailoring the microstructures using the LPBF technique.

Fusion Materials Development at Forschungszentrum Jülich

Material issues pose a significant challenge for the design of future fusion reactors. From a historic point of view, the material mix used for the first wall of a fusion reactor has continually evolved, from original steel vessels to carbon and other low‐Z materials such as beryllium to tungsten as the primary candidate for a reactor's first wall armor and divertor material. For materials considered for fusion applications, a highly integrated approach is necessary. Resilience against neutron damage, good power exhaust, and oxidation resistance during accidental air ingress are design relevant issues while deciding on new materials or improving upon baseline materials. Neutron‐induced effects, e.g., transmutation adding to embrittlement, retention, and changes to thermomechanical properties, are crucial to material performance. In this contribution, the recent progress (2013–2019) in fusion materials development for current and future fusion devices, at Forschungszentrum Jülich GmbH, with activities focussing on advanced materials and their characterization is given. It is a continuation and extension of the work given by Coenen et al.