Wall Of Fusion Devices Research Articles

Filamentary velocity scaling validation and spin dynamics in the DIII-D tokamak

Measured filament velocities in the DIII-D tokamak are compared against theoretical scalings, finding that the latter often represents an upper limit on experimental velocity distributions with most filaments possessing lower velocity. Filament spin from internal E × B drift is experimentally demonstrated to alter filament radial velocity. A critical spin velocity, where filament radial velocity peaks, is observed and corresponds to approximately 5 km/s. This transition is corroborated using a less direct measure of filament spin in the form of a temperature ratio. These techniques are combined to find that the critical spin velocity closely aligns with transport times along and across filaments becoming comparable. The normalized filament size distribution is consistent with the most stable size as dictated by Kelvin–Helmholtz and curvature-driven instabilities. Overall, the findings suggest filament stability and spin alter filamentary transport that may threaten the integrity of first walls in fusion devices.

Monitoring of tritium and impurities in the first wall of fusion devices using a LIBS based diagnostic

Laser-induced breakdown spectroscopy (LIBS) is one of the most promising methods for quantitative in-situ determination of fuel retention in plasma-facing components (PFCs) of magnetically confined fusion devices like ITER and JET. In this article, the current state of understanding in LIBS development for fusion applications will be presented, based on a complete review of existing results and complemented with newly obtained data. The work has been performed as part of a research programme, set up in the EUROfusion Consortium, to address the main requirements for ITER: (a) quantification of fuel from relevant surfaces with high sensitivity, (b) the technical demonstration to perform LIBS with a remote handling system and (c) accurate detection of fuel at ambient pressures relevant for ITER. For the first goal, the elemental composition of ITER-like deposits and proxies to them, including deuterium (D) or helium (He) containing W–Be, W, W–Al and Be–O–C coatings, was successfully determined with a typical depth resolution ranging from 50 up to 250 nm per laser pulse. Deuterium was used as a substitute for tritium (T) and in the LIBS experiments deuterium surface densities below 1016 D/cm2 could be measured with an accuracy of ∼30%, confirming the required high sensitivity for fuel-retention investigations. The performance of different LIBS configurations was explored, comprising LIBS systems based on single pulse (pulse durations: ps–ns) and double pulse lasers with different pulse durations. For the second goal, a remote handling application was demonstrated inside the Frascati-Tokamak-Upgrade (FTU), where a compact, remotely controlled LIBS system was mounted on a multipurpose deployer providing an in-vessel retention monitor system. During a shutdown phase, LIBS was performed at atmospheric pressure, for measuring the composition and fuel content of different area of the stainless-steel FTU first wall, and the titanium zirconium molybdenum alloy tiles of the toroidal limiter. These achievements underline the capability of a LIBS-based retention monitor, which complies with the requirements for JET and ITER operating in DT with a beryllium wall and a tungsten divertor. Concerning the capabilities of LIBS at pressure conditions relevant for ITER, quantitative determination of the composition of PFC materials at ambient pressures up to 100 mbar of N2, the D content could be determined with an accuracy of 25%, while for atmospheric pressure conditions, an accuracy of about 50% was found when using single-pulse lasers. To improve the LIBS performance in atmospheric pressure conditions, a novel approach is proposed for quantitative determination of the retained T and the D/T ratio. This scenario is based on measuring the LIBS plume emission at two different time delays after each laser pulse. On virtue of application of a double pulse LIBS system, for LIBS application at N2 atmospheric pressure the distinguishability of the spectra from H isotopes could be significantly improved, but further systematic research is required.

Infrared thermography in metallic environments of WEST and ASDEX Upgrade

Abstract Infra-red (IR) thermography is a widely used tool in fusion devices to monitor and to protect the plasma-facing component (PFC) from excessive heat loads. However, with the use of all-metal walls in fusion devices, deriving surface temperature from IR measurements has become more challenging. In this paper, an overview of infra-red measurements in the metallic tokamaks WEST and ASDEX Upgrade (AUG) is reported and the techniques carried out in the modeling and experimental fields to deal with this radiative and fully reflective environment are presented. Experimental characterizations of metallic samples in laboratory and experiments in WEST and AUG reveal that the behavior of both the emission and the reflectance can vary significantly with surface roughness, machining process and as the plasma operation progress. In parallel, the development of a synthetic IR diagnostic has allowed for a better interpretation of the IR images by assessing the reflection patterns and their origin. This has also proven that small-scale change in the emission pattern of beveled PFC can be confused with abnormal thermal events. Numerical solutions to evaluate the contribution of the reflections associated with a variable emissivity in a fully reflective and radiative environment are finally presented.

Diameters and Velocities of Droplets Emitted From the Cu Cathode of a Vacuum Arc

The cathode spots of vacuum arcs emit material in the form of plasma as well as droplets. Generated by arcs burning at the first wall of fusion devices, the droplets may effectively contaminate the fusion plasma. Essential characteristics of the droplets (diameter, velocity, and emission direction) and their interrelations are poorly known so far. In this paper, a new approach of optical droplet investigation is presented. Emitted from a pulsed vacuum arc, the droplets fly vertically inside a drift tube against gravity, finally passing two consecutive light beams. The time-of-flight and the detected intensity of scattered light allow a simultaneous determination of droplet velocity and size. Different solid angle directions have been realized by turning the cathode with respect to the flight tube axis. Using this technique, the parameters of droplets emitted from a Cu cathode into different directions were obtained. Two distinct groups could be identified at smaller ( 20°) angles between the surface plane and emission direction, respectively. They exhibit different velocity distributions and different relations between the particle diameter and the emission velocity.

Modeling of Multispecies Dynamics in Fusion-Related Materials with FACE

The new reaction-diffusion code FACE (First wAll simulation CodE) is developed for modeling plasma-material–interaction processes taking place in the first wall of fusion devices. The code simulates implantation, desorption, transport, and interaction of an arbitrary number of particle and quasi-particle species, such as dissolved gases and intrinsic or induced defects, in a wall material. It allows descriptive and predictive modeling of retention and outgassing of plasma constituents in plasma-exposed materials for analysis of experimental data and assessment of performance of plasma-facing components under various static and transient plasma conditions. The physical model, capabilities, future development, and example applications of the code are presented.

Physics of the edge plasma and first wall in fusion devices: synergistic effects

Various synergistic effects resulting from plasma–wall interactions in magnetic fusion devices are considered. The crucial role of the first wall out-gassing processes in the recovery of pedestal density in the high-confinement mode of tokamak operation after giant type-I edge localized modes (ELMs) transient events as well as in the setting the ELM period is discussed. The shielding effects of vapor plasma formed during interactions of extremely large plasma heat fluxes with material surfaces are analyzed. The strongly non-linear impact of secondary electron emission from the divertor target on the incident plasma heat flux is discussed.

Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices

Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the understanding and validate the models of the in vessel build-up of the T inventory in ITER and future D–T devices. So far, research in these areas is largely supported by post-mortem analysis of wall tiles. However, access to samples will be very much restricted in the next-generation devices (such as ITER, JT-60SA, W7-X, etc) with actively cooled plasma-facing components (PFC) and increasing duty cycle.This has motivated the development of methods to measure the deposition of material and retention of plasma fuel on the walls of fusion devices in situ, without removal of PFC samples. For this purpose, laser-based methods are the most promising candidates. Their feasibility has been assessed in a cooperative undertaking in various European associations under EFDA coordination. Different laser techniques have been explored both under laboratory and tokamak conditions with the emphasis to develop a conceptual design for a laser-based wall diagnostic which is integrated into an ITER port plug, aiming to characterize in situ relevant parts of the inner wall, the upper region of the inner divertor, part of the dome and the upper X-point region.

Application of laser-induced breakdown spectroscopy for characterization of material deposits and tritium retention in fusion devices

Laser-induced breakdown spectroscopy (LIBS) is discussed as a possible method to characterize the composition, tritium retention and amount of material deposits on the first wall of fusion devices. The principle of the technique is the ablation of the co-deposited layer by a laser pulse with P (power density)≥0.5GW/cm2 and the spectroscopic analysis of the light emitted by the laser induced plasma. The typical spatial extension of the laser plasma plume is in the order of 1cm with typical plasma parameters of ne≈3×1022m−3 and Te≈1–2eV averaged over the plasma lifetime which is below 1μs. In this study “ITER-Like” mixed deposits with a thickness of about 2μm and consisting of a mixture of W/Al/C and D on bulk tungsten substrates have been analyzed by LIBS to measure the composition and hydrogen isotopes content at different laser energies, ranging from about 2J/cm2 (0.3GW/cm2) to about 17J/cm2 (2.4GW/cm2) for 7ns laser pulses. It is found that the laser energies above about 7J/cm2 (1GW/cm2) are needed to achieve the full removal of the deposit layer and identify a clear interface between the deposit and the bulk tungsten substrate by applying 15–20 laser pulses while hydrogen isotopes decrease strongly after the first laser pulse. Under these conditions, the evolution of the spectral line intensities of W/Al/C/hydrogen can be used to evaluate the layer composition.

Comparison of hydrogen retention in W and W/Ta alloys

The extreme brittleness of tungsten (W) is one of the challenges of using W as first wall material. One attempt to alleviate this problem is to use W alloys with better mechanical properties. However these alloying elements must not degrade the favorable properties of W with respect to its application at the first wall of fusion devices: low sputter yield and hydrogen inventory.In this work we investigate the hydrogen retention in the recently proposed W/Ta alloys under deuterium ion bombardment. By directly comparing pure W and W/Ta alloys with 1% and 5% Ta content we found that the W/Ta alloys retain significantly more hydrogen than pure W under identical implantation conditions.Our finding of increased hydrogen retention together with the fact that the Ta alloying did not improve the brittleness makes W/Ta alloys an unacceptable choice for the first wall of fusion devices.

Characterization of hydrocarbon and mixed layers in TEXTOR by laser induced ablation spectroscopy

Laser-based methods are investigated for the development of an in situ diagnostic for spatially and temporally resolved characterization of the first wall in fusion devices. Here we report on the first systematic laser-induced ablation spectroscopy (LIAS) measurements carried out on various surface layers in the TEXTOR tokamak. These materials include a-C:D, mixed W/C/Al/D2, Oerlikon Balzers ‘Balinit’ diamond-like carbon layers and EK98 fine-grain graphite. In LIAS, the bulk or deposited material is evaporated during the plasma discharge by intense laser radiation. The light emitted by particles entering the edge of the ionizing tokamak plasma is then observed by optical spectroscopy. In the measurements taken, it was found that the studied layers can be identified by their characteristic line emission. A good correlation between the observed line intensity and layer thickness is found. The observed plumes show target material dependence. To analyze layers formed during tokamak operation, further investigation of the ablation process and reference materials for cross calibration is required.

Hydrogen diffusive transport parameters in W coating for fusion applications

Combined Magnetron Sputtering and Ion Implantation (CMSII) technology was used for W coating of carbon based materials (Carbon Fibre Composite – CFC and fine grain graphite) for the first wall in fusion devices. The coating thickness was 10–15 μm or 20–25 μm depending on the position of the tile at the wall. While such coatings successfully passed the demanding thermo-mechanical tests, not much is known about its hydrogen interaction. The latter is particularly important for the assessment of tritium retention. Due to the low hydrogen diffusivity and very small volume of W in the coated layer, the gaseous hydrogen permeation measurement at 400 °C was selected for the experimental technique, where increasing & decreasing transient and steady state permeation flux was monitored. Problems that could arise with the CFC membrane sealing were overcome by deposition of the identical W layer on the 0.5 mm Eurofer substrate. Two such membranes were investigated. Obtained hydrogen permeability in tungsten layer (∼10 −13 mol H 2/m s Pa 0.5) is comparable to the upper range of published data. Measured diffusivity (∼10 −14 m 2/s) is several orders of magnitude lower compared to the average of published data for tungsten, while the measured solubility (∼1 mol H 2/m 3 Pa 0.5) is several orders of magnitude higher. The explanation is given in terms of hydrogen trapping that has significant impact on hydrogen migration.

Surface chemistry of first wall materials – From fundamental data to modeling

The application of different materials at the first wall of fusion devices, like beryllium, carbon, and tungsten in the case of ITER, unavoidably leads to the formation of compounds. These compounds are created dynamically during operation and depend on the local parameters like surface temperature, incoming particle energies and species. In dedicated, well-defined laboratory experiments, using mainly X-ray photoelectron spectroscopy and Rutherford backscattering analysis for qualitative and quantitative chemical surface analysis, the parameter space in relevant element combinations are investigated. These studies lead to a deep understanding of the reaction mechanisms under the applied conditions and to a quantitative description of reaction and diffusion processes. These data can be parameterized and integrated into a modeling approach which combines dynamic surface chemistry with the modeling of the transport in the plasma. Two different approaches for surface reaction modeling are compared and benchmarked with experimental data.

Improvement of the thermo-mechanical properties of isotropic graphite, produced from carbonaceous mesophase, by the addition of different carbides

This work demonstrates the possibilities for optimization of doped fine grain graphites with high thermal conductivity and high thermal shock resistance for application as plasma-facing materials at those areas of the vessel wall in fusion devices receiving the highest particle and power loads. A mixture of carbonaceous mesophase powder and different fine grain carbides (B 4 C, TiC, VC, ZrC and WC) was used as starting material. These carbides act as catalysts, decreasing the graphitization temperature. VC acts as catalyst of the graphitization at the lowest temperature, and ZrC is the most effective catalyst of all investigated carbides. With increasing graphitization temperature, the open porosity of all doped materials becomes closed, suggesting the existence of a diffusion mechanism responsible for both the catalytic effect and the closing of the open porosity. Carbides addition does not strongly influence the mechanical properties of pure graphite.

Plasma Wall Interaction and Its Control by Wall Conditioning

The surrounding material walls in fusion devices must fulfil three important tasks:provide high vacuum conditions necessary to provide clean fusion plasmasabsorb the power produced by the α-particles in the fusion processes and injected by auxiliary heatingenable the exhaust of the helium ash by thermalisation of the helium plasma ions on material surfaces in the vicinity of helium pumps.The interaction of the plasma with the surrounding wall surfaces (PSI: plasma surface interaction) is therefore a necessary condition for fusion devices and not to avoid. In the plasma wall interaction a variety of bulk material and surface processes are involved on one side together with various special processes in the near surface plasma region on the other side. They can modify the properties of the boundary and main plasma in a feed back like behaviour.1,2 A prominent example is the release of impurities from the walls by plasma particle impact which increases the energy loss of the plasma by radiation and reduces thereby the particle fluxes to and impurity release from the walls.

Statistical Properties of Turbulence: A New Approach to Characterize Transport in Fusion Plasmas

A view of recent experimental results and progress in the characterization of the statistical properties of edge turbulence in fusion plasmas is given. The study of the dynamical interplay between fluctuations in gradients, turbulent transport and radial electric fields has shown that these parameters are strongly coupled both in tokamak and stellarator plasmas. The size of turbulent events increases when the plasma deviates from the average density gradient. The radial velocity of fluctuations is of the order of 20 m/s for transport events associated with a small deviation from the most probable gradient. On the contrary, the effective radial velocity increases up to 500 m/s for transport events in which the local gradient increases significantly above the most probable gradient. These results suggest a link between the size of transport events and the nature of transport (diffusive versus non-diffusive) in the plasma boundary region. The dynamical relationship between fluctuations in gradients and transport is strongly affected by the presence of sheared poloidal flows, heating power and the proximity to instability thresholds. There is also experimental evidence that the turbulent transport is strongly coupled with fluctuations in parallel flows, reflecting that parallel flows might, at least partially, be driven by turbulent mechanisms. Using the same methods developed for the study of turbulence, the statistical properties of the radial propagation of edge localized modes (ELMs) in the SOL of JET have also been investigated. ELM events propagate radially with effective velocities in the range of 1000 m/s, again showing a link between the radial velocity and the size of transport events. At such high velocities, the radial propagation competes strongly with parallel loss to the divertor plates and may therefore be an important mechanism transporting particles to the main chamber walls of fusion devices. This work emphasizes the importance of the statistical description of transport processes in fusion plasmas as an alternative approach to the traditional way of characterizing transport based on the computation of effective transport coefficients (i.e. diffusion coefficients) and on average quantities (i.e., average correlation lengths).

Removal of codeposited layers by ECR discharge cleaning

Redeposition of eroded carbon on the walls of fusion devices together with the D–T plasma forms amorphous, hydrogenated carbon films with D–T concentrations ranging for deposition at a surface temperature of 300 K from 0.3 up to 0.6, depending on the deposition conditions. The trapping of large amounts of hydrogen isotopes in these so-called codeposited layers may represent a potential safety hazard during D–T operation because large amounts of tritium will be retained in these layers. To erode carbon films and deplete the T concentration an experimental investigation was performed using O 2, D 2, H 2, H 2O, and O 2/H 2, electron cyclotron resonance low-pressure discharges. Real-time in-situ ellipsometry was applied to measure the optical response of the layers during deposition and erosion. This allows a precise determination of the optical constants and the film thickness. From these data the deposition and erosion rates are calculated. The dependence of the erosion rates on substrate temperature, applied ion energy, gas composition and discharge pressure for two different types of a-C:H films was measured. In the investigated parameter range oxygen discharges showed the highest erosion rates of all investigated source gases. The highest rates achieved at 300 K with the substrate at floating potential are 3.6 μm/h for polymer-like, soft C:H films (H/C about 1) and 1.7 μm/h for hard a-C:H films (H/C about 0.5). The erosion rates increase with increasing substrate temperature (roughly by a factor of 3–4, 300–650 K) and with increasing ion energy (by about a factor of 6–8, 0–200 eV). The paper presents the experimental results and a discussion of the parameter dependencies and the consequences for the removal of codeposited layers by low-pressure discharge cleaning in ITER.

Dynamic behavior of hydrogen atoms with a boronized wall

Dynamic behavior of hydrogen atoms in boron films, which is one of candidate materials of the first wall of fusion devices, has been studied. Hydrogen absorption behavior was investigated with a glow discharge in hydrogen. After a strong absorption in the beginning of the discharge, hydrogen atoms were slowly and continuously absorbed for 10 h without saturation. The slow absorption is due to deeper migration of hydrogen atoms into the film, enhanced by the ion bombardment. A helium glow discharge causes both ion induced desorption and the deeper migration, and thus, hydrogen atoms are accumulated in the film when the H 2 and He discharges are repeated alternately. Most of the retained hydrogen atoms were released by heating up to 400°C. A calculation result based on a simple model was compared with the experimental results. A recombination coefficient was obtained from the analysis of the experimental transient release of the H atoms just after the H 2 discharge. The time behavior at the initial phase of the He discharge was reproduced fairly well with use of a smaller cross-section for the He + ion induced detrapping than that for H 2 +.

Heat flux experiments on first wall mock-ups coated by plasma sprayed B 4C

The plasma compatibility of the first wall of fusion devices can be achieved either by using protective tiles of massive material like fibre reinforced carbon or by direct coating of the metallic wall. For ITER, the plasma spray technique provides the opportunity to coat large areas of the first wall with coating thicknesses of the order of 1 mm and may allow in-situ repair of the coating. Plasma sprayed boron carbide (B 4C) is regarded as a promising candidate material because of its low atomic number and high melting point. In the present work, actively cooled stainless steel AISI316L bodies were coated with a copper interlayer and a 1.2 mm top layer of B 4C. The homogeneously coated surface was 210 mm × 65 mm. The stainless steel bodies had two parallel channels for water cooling of the mock-ups. Thermomechanical testing was performed in the JEBIS-facility (JAERI Electron Beam Irradiation System). Surface heat loads of 0.5–1 MW/m 2 caused local delamination and cracking of the coating.

Erosion and redeposition at the vessel walls in fusion devices

The plasma induced erosion and redeposition at the vessel walls in today's fusion devices have been investigated both with the computer simulation code ERO, and in experiments. Well prepared carbon probes with implanted and evaporated markers in the surface layers have been exposed in the scrape-off layer (SOL) of several tokamaks such as JET, TEXTOR and ASDEX-Upgrade. The main plasma parameters (electron density and temperature, impurity concentration in the SOL) are simultaneously determined. After exposure to single plasma discharges, erosion and redeposition of the marker material were measured by surface layer analysis with MeV ion beam techniques. The experimental results were compared with the results from the ERO code. The measured erosion/redeposition could be described with ERO, which takes into account the impurity concentration in the SOL, the dynamical change of the surface composition (causing a modification of the sputtering yield during the exposure) and E × B drift effects.

Deposition of hydrogenated carbon layers from CH 3+ ion beams

Carbon deposition and hydrogen codeposition is investigated as function of ion energy, fluence and target temperature at normal incidence by bombardment of hydrocarbon molecules onto Si and graphite. An amorphous hydrogen/carbon (a-C:H) layer is formed in a thickness range of 40 to 130 nm by CH 3 + bombardment up to a fluence of 3 × 10 18 / cm 2. The deposition process and the re-erosion phenomenon of the carbon film is also studied. The experimental results are compared with TRIDYN computer simulation and previous experimental results of carbon sputtering by atomic H + and C + beams in order to have a good understanding of the interaction between hydrocarbon molecular ions and the carbon-based wall in fusion devices.