Tungsten Plasma Facing Components Research Articles

Development of ITER first wall heat load feedback control

Real-time monitoring and control of thermal loads on tungsten plasma-facing components is mandatory for ITER to avoid their overheating during plasma discharges. For the Start of Research Operations (SRO) phase, dedicated First Wall Heat Load Control (FWHLC) functions are included in the ITER Plasma Control System (PCS) to prevent the development of excessive plasma heat loads on the inertially cooled first wall (FW). In this work, we propose a FWHLC design with a model-based approach, which features a simplified thermal model for the FW armour. This control-oriented model simulates the thermal response of ITER FW to slow changes in plasma equilibrium and energy, which during ITER operation will be monitored by the real-time wide angle infrared camera system. This allows computationally efficient runs for rapid controller performance assessment but can also assist operation scenario design by pre-empting potential heat load issues. FWHLC is designed to react to measured and predicted FW temperatures overcoming predesigned limits with real-time variations of the plasma shape or auxiliary power references, aiming to reduce thermal loads before a premature plasma termination is required to protect the FW. The new scheme is tested in closed loop simulations, where perturbations to nominal ITER low current scenarios are introduced in the wall thermal model to trigger the response of FWHLC, supporting the effectiveness of the proposed policy.

Data streaming infra-red video bolometer (IRVB) of Korea Superconducting Tokamak Advanced Research (KSTAR).

Due to the increasing demands for active plasma control operations, in situ diagnostics are highly sought after. Tungsten plasma-facing components have been utilized in the Korea Superconducting Tokamak Advanced Research (KSTAR) lower divertor since the 2023 campaign. Plasma radiation is a key parameter for plasma control, especially in radiation front control experiments. Therefore, the KSTAR infra-red video bolometer (IRVB) needs to be reconfigured into an in situ data streaming diagnostic. This requires comprehensive changes in both infra-red camera control and data analysis compared to the previous system. To ensure the stability of the reconfigured system, functional parts are grouped and separated into individual processes to protect camera acquisitions from errors in other processes. In addition, to enhance the speed of data streaming analysis, the analysis code has been optimized and converted into graphics processing unit (GPU) code. Besides the data streaming analysis, the system is also designed to support post-shot analysis with the entire frame data for the same shot to address frame drop issues encountered during data streaming. Radiation front control experiments with N2 gas seeding are successful results of the data streaming IRVB for its commissioning. This paper focuses on the development of the data streaming IRVB system.

Helium plasma operations on ASDEX Upgrade and JET in support of the non-nuclear phases of ITER

For its initial operational phase, ITER has until recently considered using non-nuclear hydrogen (H) or helium (He) plasmas to keep nuclear activation at low levels. To this end, the Tokamak Exploitation Task Force of the EUROfusion Consortium carried out dedicated experimental campaigns in He on the ASDEX Upgrade (AUG) and JET tokamaks in 2022, with particular emphasis put on the ELMy H-mode operation and plasma-wall interaction processes as well as comparison to H or deuterium (D) plasmas. Both in pure He and mixed He + H plasmas, H-mode operation could be reached but more effort was needed to obtain a stable plasma scenario than in H or D. Even if the power threshold for the LH transition was lower in He, entering the type-I ELMy regime appeared to require equally much or even more heating power than in H. Suppression of ELMs by resonant magnetic perturbations was studied on AUG but was only possible in plasmas with a He content below 19%; the reason for this unexpected behaviour remains still unclear and various theoretical approaches are being pursued to properly understand the physics behind ELM suppression. The erosion rates of tungsten (W) plasma-facing components were an order of magnitude larger than what has been reported in hydrogenic plasmas, which can be attributed to the prominent role of He2+ ions in the plasma. For the first time, the formation of nanoscale structures (W fuzz) was unambiguously demonstrated in H-mode He plasmas on AUG. However, no direct evidence of fuzz creation on JET was obtained despite the main conditions for its occurrence being met. The reason could be a delicate balance between W erosion by ELMs, competition between the growth and annealing of the fuzz, and coverage of the surface with co-deposits.

Design of a multi-energy soft X-ray diagnostic for profile measurements during long-pulse operation in the WEST tokamak

The W Environment in Steady-state Tokamakwas designed and built to test ITER-like tungsten plasma facing components in a long pulse (∼1000 s) scenario. Recently, a multi-energy soft x-ray diagnostic (MESXR) was installed in the WEST (W Environment in Steady-state Tokamak), to understand the sources, transport and confinement of high-Z impurities. The purpose of this work is to describe the engineering challenges posed by the long pulse scenario for the integration of the MESXR diagnostic and how they were addressed and solved. The calculations of vacuum and thermal stress as well as heat transfer are presented and discussed.

High-velocity dust impacts in plasma facing materials: Insights from molecular dynamics simulations

This research investigates the interaction between high-speed tungsten (W) dust and plasma-facing components (PFCs) in fusion reactors, particularly focusing on W walls. Through molecular dynamics (MD) simulations, the study covers a broad spectrum of W dust velocities to evaluate their effect on wall materials with various crystal orientations. We found that high-speed impacts cause considerable damage, including sputtering, degradation, and deformation. The study introduces a damage model derived from experimental and simulation data that reveals the patterns and mechanisms of damage caused by dust impacts. The proposed model significantly improves our understanding of dust-wall interactions and underscores the importance of MD simulations as a reliable technique for exploring such phenomena in the challenging conditions of fusion devices. These insights are crucial to predict and mitigate damage to PFCs, helping to develop more resilient and efficient components. Overall, the research offers valuable knowledge on the atomic-level dynamics of dust impacts and represents a notable advancement in the durability and efficiency of materials used in fusion energy technologies.

Recent DIII-D progress toward validating models of tungsten erosion, re-deposition, and migration for application to next-step fusion devices

Fundamental mechanisms governing the erosion and prompt re-deposition of tungsten impurities in tokamak divertors are identified and analyzed to inform the lifetime of tungsten plasma-facing components in ITER and other future devices. Various experiments conducted at DIII-D to benchmark predictive models are presented, leveraging the DiMES removable sample exposure probe capability and the Metal Rings Campaign, in which toroidally symmetric rows of tungsten-coated tiles were installed in the DIII-D divertor. In tokamak divertors, the width of the electric sheath is of the order of the main ion Larmor radius, and a vast majority of sputtered tungsten impurities are typically ionized within the sheath. Therefore, W prompt redeposition is mainly governed by the ratio of the characteristic ionization mean-free path of neutral tungsten to the width of the sheath. In-situ monitoring of the prompt redeposition of tungsten impurities in divertors is demonstrated via the use of WII/WI line ratios and the ionizations/photon (S/XB) method in L-mode discharges. Even with this relatively limited set of emission measurements, net erosion measurements were found to be a consistent upper bound to an analytic scaling based on the ratio of the W ionization length, and the width of the magnetic sheath rather than the ratio of and the W+ gyro-radius. In the far-scrape-off layer (SOL) of the ITER divertor, however, it is calculated that the measurement of photon emissions associated with the ionization of tungsten impurities up to may be required. Finally, W deposition patterns on DiMES collector probes, interpreted via DIVIMP-WallDYN modelling, reveal the key roles of progressive W erosion/re-deposition staps and E × B drifts in regulating long-range high-Z material migration.

Performance of a liquid Sn divertor target during ASDEX upgrade L-mode and H-mode operation

One of the ways to extend the lifetime of the divertor for DEMO could be to replace the solid tungsten plasma-facing components with liquid tin (Sn) confined in a tungsten capillary porous structure (CPS). Testing a CPS in a divertor plasma configuration is crucial for the development of a liquid metal divertor (LMD) to understand how the main plasma is affected. Only a limited Sn concentration is allowed in the plasma core, due to the high radiative losses associated with the high atomic number of Sn (50). Therefore, it is necessary to test a small-scale LMD filled with Sn in a tokamak environment, which has not previously been done. In ASDEX Upgrade, a liquid tin module (LTM) has been exposed by means of the divertor manipulator. During plasma flat-top, the outer strike point (OSP) was placed onto the pre-heated LTM and held there for a time interval between 2 and 3.4s over multiple discharges. Photographs of the LTM taken after each discharge, revealed macroscopic Sn leakage onto the adjacent tile. Simulations with the HeatLMD code predicted an acceptable tin erosion near the LTM with thermal sputtering dominating over evaporation. However, spectroscopic measurements revealed an order of magnitude higher erosion. Since this remained constant when the OSP was held on the LTM so that the surface temperature increased, evaporation could be excluded as the main source of Sn erosion. Comparison between discharges with different durations of OSP location on the LTM revealed an increase in core radiation up to 1.5MW due to Sn. The 1.5D-impurity transport code STRAHL was used to interpret this increase in total plasma radiation and revealed a Sn concentration in the main plasma of up to 1.4×10-4. Given that the LTM only covered about 1/650 of the outer divertor circumference, extrapolating to a full toroidal divertor implies erosion is above acceptable limits. The unexpectedly high Sn fraction in the main plasma is attributed to the ejection of Sn droplets reaching the main plasma, which may have originated from either the CPS or leaked tin. This conclusion is also supported by splashes of tin droplets, which were observed on the adjacent divertor tile and one ∼0.5m downstream. Therefore, to make a Sn-filled LMD a viable alternative to solid tungsten, the formation of droplets must be reduced by two orders of magnitude.

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.

Energetics of silicon in the bulk and near surfaces of tungsten: a first-principles study

Siliconization of the tokamak walls is a candidate method to improve plasma confinement in fusion tokamaks containing tungsten plasma facing components (W PFCs). To understand the interactions of silicon (Si) with W, the Si behavior in bulk W, and near three low-index W surfaces ((100), (110) and (111)) has been investigated using first-principles density functional theory. In bulk W, Si interstitial atoms have a low solution ability and high mobility, and Si atoms can be strongly trapped by W vacancies. The interaction between two Si adatoms is responsible for the stability of adatom superstructures on W surfaces, consistent with previous experimental observation (Tsong and Casanova 1981 Phys. Rev. Lett. 47 113). Although the coverage dependence of Si adsorption and diffusion energetics on surfaces is related to surface orientation, the W(110) surface has lower Si adsorption affinity and higher Si diffusivity than either the W(111) or W(100) surfaces. The most stable Si adatom superstructure on W surfaces is: square c(2 × 2) pattern on W(100) covered with 0.5 ML Si; rectangular c(4 × 2) pattern on W(110) with 0.25 ML Si; and rhombus p(1 × 1) pattern on W(111) with 1 ML Si. The coverage dependence of Si mobility on/toward W surfaces is generally related to the stability of the Si superstructures as a function of coverage on each surface. Interestingly, Si adatoms prefer to transport below the surface and into W subsurface by an exchange mechanism with W atoms, indicating the likelihood of epitaxial growth of W silicide layers on W surfaces during the operation of W PFCs.

Analysis of high heat flux tested tungsten mono-blocks by hardness and microstructural profiling

The integrity of tungsten plasma facing component mockups was investigated by microstructure and hardness measurement before and after high heat flux (HHF) test. The mockups consisting of five ITER-like tungsten mono-blocks were fabricated employing gas pressure casting and hot radial pressing (HRP) method. The tungsten blocks were machined from rolled W plates of two commercial suppliers. The tested mockups showed discernible results in response to different heat flux conditions of which increased heat flux was of importance. Post-mortem analysis of the tungsten armor of the HHF tested mockups were conducted through microstructure and hardness profiling from surface along the depth. The Vickers hardness profile measured from the surface into the depth showed notable decrease around 300 μm ∼ 1600 μm depth after HHF test compared to that before the test. The depth and magnitude of hardness change was dependent on the heat loading condition and the position in the tungsten block. The behavior of hardness indicating the materials property of tungsten showed correlation with the microstructural change measured by electron back scatter diffraction (EBSD).

Characterisation of electron cyclotron wall conditioning plasma in ASDEX Upgrade

Electron cyclotron wall condition (ECWC) discharges are characterised in ASDEX Upgrade with full tungsten plasma facing components and X2 polarised waves launched from the equatorial ports, relevant to ECWC conditions in ITER Pre-Fusion Power Operation phase 1. The characterisation of the deuterium plasmas is based on experimental inputs such as electron density measurements, in-vessel pressure measurements, poloidal field maps obtained from the measured coil currents, as well as advanced tomographic methods on camera images filtered at the hydrogen Balmer lines. TOMATOR-1D simulations and collective Thomson scattering radiometer spectra complement the findings. The cold, high density and partially ionized toroidal plasmas show significant levels of stray radiation. The measured radiation includes waves at half of the gyrotron frequency suggesting the occurrence of parametric decay instabilities at 2nd harmonic upper hybrid resonance that locates at the low field side of the resonance. A displacement of the plasma emission along the resonance layer is observed at higher discharge power in discharges with a vertical poloidal magnetic field only. By optimizing the poloidal field pattern, along with the location of the electron cyclotron heating (ECH) resonance, the strongest surface interaction regions for the charged particles can be controlled. Directing plasma flux to inner wall surfaces, and same for the inner divertor apron, is found less effective in ASDEX Upgrade due to magnetic mirror effects and outward convective flows. Modeling however predicts the presence of an intense and uniform flux of low energy atoms produced at the ECH absorption layer that may be effective for conditioning the high field side surfaces after use of the disruption mitigation system.

Numerical feasibility study on signal discrimination in reflective environment of tungsten PFC

In the reflective environments of tungsten plasma facing components (PFCs), diagnostics of tokamaks will suffer from signal distortions since the reflectivity of tungsten PFCs are not negligible especially in the infrared (IR) range [1,2]. However, considering that most optical reflective systems are linear, it might be possible to eliminate the reflections if all ray information is provided. So, the possibility depends on the ray information of the system, and it can be derived using ray tracing simulations. For the feasibility study, the own ray tracing program of KFE has to be built since the conventional ray tracing programs have a limitation in providing full ray information. Among diagnostics, temperature measurement of PFCs is an issue for tungsten PFCs since it is a fundamental measurement for machine safety. So, as a feasibility study of signal discriminations, the emitted radiance was examined, and it has shown to be quite successful. This paper shows the possibility of reflection discrimination using ray path information.

A model of ballistic helium transport during helium-induced fuzz growth in tungsten

Ballistic helium transport in tungsten during fuzz growth was investigated to provide insight into the helium-induced nanostructuring process which adversely affects the near-surface properties of tungsten plasma-facing components. An analytical model of helium ion transport in the vacuum region within the fuzz layer was developed, whereby He ions are assumed to move in straight-line trajectories with mean free paths determined from fuzz/nanotendril dimensions. Reflection of ions from tendril sides was considered, with He ions allowed to implant into the bulk at the base of the fuzz layer if they maintain ≥5eV of energy necessary to overcome the He-W surface barrier potential, resulting in an increase in the effective He mean free path. Using the model results for helium implantation in the bulk, a helium fluence-fuzz thickness relation was achieved, which matches well with experimental data in the literature, and implies that the fuzz growth rate is consistent with ballistic implantation into the bulk through the growing porous fuzz layer.

Comparative Analysis of Spectroscopic Studies of Tungsten and Carbon Deposits on Plasma-Facing Components in Thermonuclear Fusion Reactors

Studies on the erosion products of tungsten plasma-facing components (films, surfaces, and dust) for thermonuclear fusion reactors by spectroscopic methods are considered and compared with those of carbon deposits. The latter includes: carbon–deuterium CDx (x ~ 0.5) smooth films deposited at the vacuum chamber during the erosion of the graphite limiters in the T-10 tokamak and mixed CHx-Me films (Me = W, Fe, etc.) formed by irradiating a tungsten target with an intense H-plasma flux in a QSPA-T plasma accelerator. It is shown that the formerly developed technique for studying CDx films with 15 methods, including spectroscopic methods, such as XPS, TDS, EPR, Raman, and FT-IR, is universal and can be supplemented by a number of new methods for tungsten materials, including in situ analysis of the MAPP type using XPS, SEM, TEM, and probe methods, and nuclear reaction method. In addition, the analysis of the fractality of the CDx films using SAXS + WAXS is compared with the analysis of the fractal structures formed on tungsten and carbon surfaces under the action of high-intensity plasma fluxes. A comparative analysis of spectroscopic studies on carbon and tungsten deposits makes it possible to identify the problems of the safe operation of thermonuclear fusion reactors.

Electron-impact ionization of the tantalum atom

Electron-impact ionization cross sections for the ground configuration of the tantalum (Ta) atom are calculated using a combination of non-perturbative time-dependent close-coupling and perturbative distorted-wave methods. Direct ionization of the 6s and 5d subshells leading to single ionization are presented. The inclusion of a polarization potential in the 6s direct ionization leads to a small reduction in the cross section. Angular factors are included that allow the configuration-average direct ionization cross sections to be split into LS term resolution. The magnitude of the ionization cross section suggests that the ionization mean free path for neutral Ta will be similar to that for tungsten (W) plasma facing components in fusion experiments, leading to similarly large fractions of prompt redeposition.

Development of an Integrated Multidiagnostic to Assess the High-Z Impurity Fluxes in the Metallic Environment of WEST Using IMAS

WEST is an actively cooled, long-pulse tokamak with nearly all plasma-facing components (PFC) made of tungsten. One of the aims of WEST is to study plasma operations with tungsten PFCs in preparation for long-pulse operations on high-Z divertor devices, such as ITER. For long-pulse operation, the high-Z impurity content and transport to the core plasma are critical concerns that require further measurement and interpretation to improve plasma performance and PFC durability. This work details the impurity influxes in WEST during a series of discharges in which the lower hybrid (LH) injected power was incrementally increased. An analysis has been performed of measurements collected from an array of edge diagnostics. Visible spectroscopy was utilized to measure the spectral radiances generated by fuel particles ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$D$ </tex-math></inline-formula> ) and impurities ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$W$ </tex-math></inline-formula> , <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$O$ </tex-math></inline-formula> , and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$C$ </tex-math></inline-formula> ) at the divertor and at the antennas with a newly developed spectral peak-fitting tool used to analyze the data in WEST. The scrape-off layer (SOL) plasma conditions ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$n_{e}$ </tex-math></inline-formula> and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$T_{e}$ </tex-math></inline-formula> ) measured at the divertor target with flush Langmuir probes and near the outer mid-plane (OMP) using reciprocating Langmuir probes (RCPs) are used to evaluate the number of ionizations per photons (S/XB) coefficients required to estimate the impurity fluxes obtained with the collisional-radiative model ColRadPy. The array of edge diagnostics discussed in this work, coupled with SOL plasma modeling tools, represents a multidiagnostic interpretative modeling workflow that will continue to be applied to upcoming experimental campaigns on the WEST experiment to assess high-Z impurity transport.

Electrodeposition of Functionally Graded Brazing Interlayers for Enhanced Joint Strength between Fusion Plasma-Facing Materials and Heat Sinks

In response to the significant plasma science and engineering breakthroughs witnessed over the past decade, the fusion research community has developed an ambitious roadmap to achieve large-scale, energy-positive fusion reactors that will enable economically competitive fusion electricity. Critical to the success of commercial fusion energy is the development of materials able to serve as plasma facing components (PFCs) that can withstand the unprecedented high heat, particle, and neutron fluxes within the fusion reactor. Tungsten is a leading candidate for the plasma-facing portion of these PFCs because of its high melting point, high sputtering resistance, and low tritium retention. Copper or copper-based alloys (e.g., copper-chromium-zirconium alloy C18150) and reduced activation ferritic/martensitic (RAFM) steels have been proposed for the heat sink materials behind the tungsten armor due to their high thermal conductivity. Brazing is considered a promising method for joining the tungsten armor and heat sink layer. However, the extreme mismatch in the coefficient of thermal expansion (CTE) between tungsten and these potential heat sink materials makes directly joining these two layers challenging. This thermal stress can be reduced by incorporating a joining layer of functionally graded materials between the tungsten and heat sink that imparts gradual changes in CTE. Copper/tungsten or copper/tungsten carbide composites are leading candidates for these functionally graded materials. For these composites, gradual changes in the CTE are achieved by varying the volume fraction of tungsten/tungsten carbide in the composite, from tungsten-rich at the tungsten armor to tungsten-poor at the heat sink.This talk will discuss the development of a scalable, electrochemical approach for fabricating interlayers with functionally graded CTE that can decrease the thermal mismatch between tungsten PFCs and copper or RAFM-based heat sinks. In this approach, the copper/tungsten or copper/tungsten carbide composite are electrodeposited onto the tungsten PFC. This component is then bonded to the heat sink via brazing. A key objective of the work is to achieve compositional control during electrodeposition through the use of highly tunable pulse and pulse-reverse electric fields. Engineering of the electric field enables control of the mass transport and crystallization phenomena in the deposition process and thus enables control of composite properties such as composition, porosity, and surface roughness. Results from electrodeposition trials and mechanical testing of brazed joints under ambient conditions as an initial screen for interlayer performance will be discussed, in anticipation of high-heat flux testing of joint performance under fusion-relevant conditions.

Investigation of plasma wall interactions between tungsten plasma facing components and helium plasmas in the WEST tokamak

ITER will operate with a tungsten divertor, a material featuring surface morphology changes when exposed to helium plasmas, in particular the formation of the so called tungsten fuzz under specific conditions. Investigating interactions between tungsten plasma facing components and helium plasmas in a tokamak environment is therefore a key point to consolidate predictions for the ITER divertor performance and lifetime. To this end, a dedicated helium campaign was performed in the full tungsten WEST tokamak, cumulating ∼2000 s of repetitive L mode discharges. It is shown that conditions for tungsten fuzz formation, as derived from linear devices experiments (incident helium energy E inc > 20 eV, helium fluence >1024 He/m2, surface temperature T surf > 700 °C), were met in the outer strike point (OSP) area of the inertially cooled tungsten divertor elements of WEST. Preliminary inspection of the components after the campaign did not show visible signs of surface modification, but points to significant net erosion in the OSP area. An extensive post mortem analysis is now ongoing to confirm these first indications. These results underline that the complex balance between erosion/redeposition (in particular linked to impurities) and tungsten fuzz formation needs to be taken into account in tokamak conditions.

Onset of fuzz formation in plasma-facing tungsten as a surface morphological instability

Based on simulation results and theoretical analysis according to an atomistically informed continuum-scale model for the surface morphological response of tungsten plasma-facing component (PFC) in nuclear fusion devices, we demonstrate that the onset of the well-known surface nanostructure (known as ``fuzz'') formation in PFC tungsten is the outcome of a stress-induced surface morphological instability. Specifically, we show that formation and growth of nanotendrils emanating from the exposed surface of PFC tungsten, a precursor to fuzz formation, is caused by a long-wavelength surface morphological instability triggered by compressive stress in the PFC near-surface layer due to over-pressurized helium (He) bubbles forming in this near-surface region through implantation of low-energy He ions from the plasma. Using linear stability theory (LST), we predict the onset of surface growth in response to low-amplitude perturbations from a planar PFC surface morphology and calculate the average spacing between nanotendrils growing from the PFC surface, in excellent agreement with self-consistent dynamical simulations of surface morphological response according to the fully nonlinear surface evolution model starting with random fluctuations from the planar surface morphology that result in nanoscale surface roughness. In addition, we examine the morphological response of the PFC surface to low-amplitude perturbations of very long wavelengths from its planar morphology and interpret fully the simulation results on the basis of a weakly nonlinear tip-splitting instability theory, which predicts a post-instability nanotendril pattern formation with nanotendril separation consistent with the LST predictions regardless of the initial surface perturbation. Finally, we compare our simulation predictions for surface nanotendril growth with experimental measurements of fuzz layer growth on the exposed surface of PFC tungsten. Our simulation results are in very good agreement with the surface growth measurements at the early stages of fuzz growth, further establishing the onset of fuzz formation in PFC tungsten as the outcome of a stress-driven PFC surface morphological instability.

Double pulse laser-induced breakdown spectroscopy for the analysis of plasma-facing components

Laser-induced breakdown spectroscopy is applied successfully for plasma-wall interaction studies in several fusion devices and post-mortem analyses of plasma-facing materials. However, the quantitative as well as qualitative analysis of low hydrogen isotope contents in tungsten plasma-facing components is still challenging. A promising approach to increase the optical signal in laser-induced breakdown spectroscopy is to apply a second laser pulse to the laser-produced plasma. We present two setups for post mortem plasma-facing component analyses using different laser pulse properties and different excitation geometries. The enhancement factors and changes in spectral line shapes for double pulse compared to single pulse laser-induced breakdown spectroscopy are presented.