Plasma-wall interaction in magnetic fusion devices is responsible for wall changes and plasma pollution with major safety issues. It is investigated both in situ and ex situ, especially by realizing large scale dedicated post-mortem campaigns. Selected parts of the walls are extracted and characterized by several techniques. It is important to extract hydrogen isotopes, oxygen or other element content. This is classically done by ion beam analysis and thermal desorption spectroscopy. Raman microscopy is an alternative and complementary technique. The aim of this work is to demonstrate that Raman microscopy is a very sensitive tool. Moreover, if coupled to other techniques and tested on wellcontrolled reference samples, Raman microscopy can be used efficiently for characterization of wall samples. Present work reviews long experience gained on carbon-based materials demonstrating how Raman microscopy can be related to structural disorder and hydrogen retention, as it is a direct probe of chemical bonds and atomic structure. In particular, we highlight the fact that Raman microscopy can be used to estimate the hydrogen content and bonds to other elements as well as how it evolves under heating. We also present state-of-the-art Raman analyses of beryllium- and tungsten-based materials, and finally, we draw some perspectives regarding boron-based deposits.

The adsorption of nitrogen molecules on a (100) tungsten surface has been studied using a new potential energy surface in which long-range interactions are suitably characterized and represented by the Improved Lennard-Jones function. The new potential energy surface is used to carry out molecular dynamics simulations by adopting a semiclassical collisional method that explicitly includes the interaction with the surface phonons. The results of the sticking probability, evaluated as a function of the collision energy, are in good agreement with those obtained in the experiments and improve the already good comparison recently obtained with calculations performed using interactions from the Density Functional Theory method and corrected for long-range van derWaals contributions. The dependence of trapping probability on the surface temperature for a well-defined collision energy has also been investigated.

Erosion of plasma-facing components affects their lifetime and other plasma-material interaction (PMI) issues important for ITER. Microscale morphology is shown to have a significant effect on surface sputtering properties [1], thus influencing the erosion-deposition pattern in tokamaks. Linear plasma devices (LPDs) are a perfect testbed for the investigation of this topic due to their cost-effectiveness and well-controlled exposure conditions. Modelling of the experiments with PMI codes, like ERO2.0 [1-2], is then highly recommended to gain insight into relevant processes. Present work reports on the investigation of the role of roughness in the sputtering process of tungsten (W) by helium (He) plasma of the linear device GyM (B?80 mT). Helium is of great interest when studying PMI since it will be present in a fusion plasma as an intrinsic impurity and it will also be the main plasma species during ITER pre-fusion power operation. W coatings deposited on: silicon (Si) substrates with pyramids on the surface and four different average roughnesses (Ra?3, 300, 600, 900 nm), and graphite substrates with irregular surface and three different Ra (?7, 90, 280 nm), as well as reference polished bulk W samples (Ra?10 nm), have been exposed in GyM changing the incident He+ energy (EHe+) between one experiment and the other in 30 - 350 eV range (i.e. by applying different bias voltage values to the samples), for a fluence of 4.0e24 He+m-2. Net erosion of the samples has been estimated from mass loss data. Morphology modifications have been investigated by scanning electron and atomic force microscopies. Experimental outcomes have been finally compared to ERO2.0 results. Considering W/Si samples, surface modifications were limited to the formation of ripples at the nanoscale for EHe+>=250 eV. This allowed to evaluate the quasi-static effective sputtering yield (YW|GyM) from mass loss data, on the one hand, and run single time step ERO2.0 simulations, on the other hand. For EHe+<=200 eV, YW|GyM is negligible. For higher energies, YW|GyM decreases by increasing the mean value of the surface inclination angle distribution (?m), in agreement with ERO2.0 results. ?m is thus the key-parameter determining the erosion of the samples rather than Ra, as also pointed out in [3]. Moreover, YW|GyM is about one order of magnitude lower than that from ion beam experiments and binary collision approximation (BCA) calculations, confirming what was observed in other LPDs [4]. Since the energy and angular distributions of sputtered particles in ERO2.0 were provided by the BCA SDTrimSP code, the effective YW from the code is also higher than YW|GyM. Calibration of ERO2.0 input, as the sputtering distributions and the incoming plasma flux, was necessary to improve the quantitative agreement with experimental data. The exposures of W coatings deposited on graphite substrates and polished bulk W samples are currently ongoing and the results will be presented during the Conference.

The demand for transparent conductive films (TCFs) is dramatically increasing. In this work tungsten oxide (WO3-x) is studied as a possible option additional to the existed TCFs. We introduce WO3-x thin films fabricated by a non-reactive magnetron RF-sputtering process at room temperature, followed by thermal annealing in dry air. Films are characterized morphologically, structurally, electrically, optically, and dielectrically. Amorphous WO3-x thin films are shown to be n-type conductive while the transparency extends to the near-IR. By evaluating a figure of merit for transparent-conductive performance and comparing to some most-widely used TCFs, WO3-x turns out to outperform in the near-IR optical range

Ultra-high temperature ceramics (UHTCs) are candidate materials for use in extreme environment owing to their melting point exceeding 3000°C and excellent combination of thermo-mechanical properties. Boride-based ceramics have strengths of 500-600 MPa up to 1500°C, when doped with suitable secondary phases and densified using the proper sintering technique. However, upon testing at higher temperatures, strength generally collapses to 200 MPa. Here, the strength behavior of Zr/Hf-B2 based ceramics at temperatures up to 2100°C is presented and related to the microstructure tailoring in terms of secondary phases, grain morphology and grain size. Strengths over 1 GPa 1800°C were measured and fracture analysis and transmission electron microscopy showed this behavior to be due to a particular morphology of the grains, known as core-shell, which included a solid solution around the native boride grain. Densification and annealing treatment at high temperatures enabled to develop a hierarchical hybrid structure where metallic nanoparticles were homogeneously dispersed in micrometric ceramic grains resulting in unprecedented refractoriness.

Gross and net erosion of tungsten (W) and other plasma-facing materials in the divertor region have been investigated in deuterium (D) and helium (He) plasmas during dedicated experiments in L- and H-mode on ASDEX Upgrade and after full-length experimental campaigns on the WEST tokamak. Net erosion was determined via post-exposure analyses of plasma-exposed samples and full-size wall components, and we conclude that the same approach is applicable to gross erosion if marker structures with sub-millimeter dimensions are used to eliminate the contribution of prompt re-deposition. In H-mode plasmas, gross erosion during ELMs may exceed the situation in inter-ELM conditions by 1-2 orders of magnitude while net erosion is typically higher by a factor of 2-3. The largest impact on net erosion is attributed to the electron temperature while the role of the impurity mixtures is weaker, even though both on ASDEX Upgrade and WEST significant amounts of impurities are present in the edge plasmas. Impurities, on the other hand, will lead to the formation of thick co-deposited layers. We have also noted that with increasing surface roughness, net erosion is strongly suppressed and the growth of co-deposited layers is enhanced. In He plasmas, gross erosion is increased compared to D due to the higher mass and charge states of the plasma particles, resulting from larger energies due to sheath acceleration, but strong impurity fluxes can result in apparent net deposition in the divertor. Our results from ASDEX Upgrade and WEST are comparable and indicate typical net-erosion rates of 0.1-0.4 nm s(-1), excluding the immediate vicinity of the strike-point regions.

The evolution of the JET high performance hybrid scenario, including central accumulation of the tungsten (W) impurity, is reproduced with predictive multi-channel integrated modelling over multiple confinement times using first-principle based core transport models. Eight transport channels (Ti, Te, j, nD, nBe, nNi, nW, ) are modelled predictively, with self-consistent sources, radiation and magnetic equilibrium, yielding a system with multiple non-linearities: This system can reproduce the observed radiative temperature collapse after several confinement times. W is transported inward by neoclassical convection driven by the main ion density gradients and enhanced by poloidal asymmetries due to centrifugal acceleration. The slow evolution of the bulk density profile sets the timescale for W accumulation. Modelling this phenomenon requires a turbulent transport model capable of accurately predicting particle and momentum transport (QuaLiKiz) and a neoclassical transport model including the effects of poloidal asymmetries (NEO) coupled to an integrated plasma simulator (JINTRAC). The modelling capability is applied to optimise the available actuators to prevent W accumulation, and to extrapolate in power and pulse length. Central NBI heating is preferred for high performance, but gives central deposition of particles and torque which increase the risk of W accumulation by increasing density peaking and poloidal asymmetry. The primary mechanism for ICRH to control W in JET is via its impact through turbulence in reducing main ion density peaking (which drives inward neoclassical convection), increased temperature screening and turbulent W diffusion. The anisotropy from ICRH also reduces poloidal asymmetry, but this effect is negligible in high rotation JET discharges. High power ICRH near the axis can sensitively mitigate against W accumulation, and dominant ion heating (e.g. He-3 minority) is predicted to provide more resilience to W accumulation than dominant electron heating (e.g. H minority) in the JET hybrid scenario. Extrapolation to DT plasmas finds 17.5 MW of fusion power and improved confinement compared to DD, due to reduced ion-electron energy exchange, and increased Ti/Te stabilisation of ITG instabilities. The turbulence reduction in DT increases density peaking and accelerates the arrival of W on axis; this may be mitigated by reducing the penetration of the beam particle source with an increased pedestal density.

In preparation for the upcoming JET D-T campaign, great effort has been devoted during the 2015-2016 JET campaigns with the ITER-like wall (ILW) to the extension of the high performance H-mode phase in baseline and hybrid scenarios. Hybrid discharges were the only ones that have been stopped by the real-time vessel protection system due hot-spot formation on the outboard poloidal limiter. Generation of hot-spots was linked to the application of high neutral beams injection and ion cyclotron resonance heating (ICRH) power. In tokamaks with high-Z plasma components, the use of ICRH heating is also accompanied by an increased metallic impurity content. Simultaneous control of hot-spot temperature and the core impurity content was crucial due to the fact that the same plasma-wall interaction mechanism is responsible for both phenomena. Impurity data collected by SXR, EUV and VUV diagnostics were able to provide for the first time comprehensive information concerning tungsten and mid-Z impurities such as nickel, iron, and cooper. To determine absolute mid-Z impurity concentrations a new relative calibration technique, compatible with JET-ILW, has been developed based on cross-calibration with a calibrated spectrometer via the quasicontinuum of W in the 200-400 Å wavelength range. In hybrid discharges, it was found that local D2 gas injection, plasma current, separatrix density, and fast ion losses appeared to impact hot-spot temperature and core impurity levels. Analysis showed a reduced maximum hot-spot temperature and impurity concentration at higher gas rate. Changes in the plasma current had a strong impact on the plasma-wall interaction, both via modifications in the edge density and in the fast ion losses. At constant gas injection rate, both the hot-spot temperature and the core impurity content decreased with the separatrix density. The main mechanism responsible for the formation of the hot-spots was found to be linked to the fast ion losses, but RF sheath effects may also be playing a role in the high limiter temperatures observed in these experiments. © Institute of Plasma Physics and Laser Microfusion.

The first comparison of dust radius and surface temperature estimates, obtained from spectroscopic measurements of thermal radiation, with simulations of dust heating and vaporization by the MIGRAINe dust dynamics code is reported. The measurements were performed during controlled tungsten dust injection experiments in the cold and dense plasmas of Pilot-PSI, reproducing ITER divertor conditions. The comparison has allowed us to single out the dominating role of the work function contribution to the dust heating budget. However, in the plasmas of interest, dust was found to enter the strong vaporization regime, in which its temperature is practically insensitive to plasma properties and the various uncertainties in modeling. This makes the dust temperature a poor figure of merit for model validation purposes. On the other hand, simple numerical scalings obtained from orbital-motion-limited estimates were found to be remarkably robust and sufficient to understand the main physics at play in such cold and dense plasmas.

For the first time, over five confinement times, the self-consistent flux driven time evolution of heat, momentum transport and particle fluxes of electrons and multiple ions including Tungsten (W) is modeled within the integrated modeling platform JETTO (Romanelli et al 2014 Plasma Fusion Res. 9 1-4), using first principle-based codes: namely, QuaLiKiz (Bourdelle et al 2016 Plasma Phys. Control. Fusion 58 014036) for turbulent transport and NEO (Belli and Candy 2008 Plasma Phys. Control. Fusion 50 95010) for neoclassical transport. For a JET-ILW pulse, the evolution of measured temperatures, rotation and density profiles are successfully predicted and the observed W central core accumulation is obtained. The poloidal asymmetries of the W density modifying its neoclassical and turbulent transport are accounted for. Actuators of the W core accumulation are studied: removing the central particle source annihilates the central W accumulation whereas the suppression of the torque reduces significantly the W central accumulation. Finally, the presence of W slightly reduces main ion heat turbulent transport through complex nonlinear interplays involving radiation, effective charge impact on ITG and collisionality.

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma-material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.

Ion cyclotron resonance heating (ICRH) in the hydrogen minority scheme provides central ion heating and acts favorably on the core tungsten transport. Full wave modeling shows that, at medium power level (4MW), after collisional redistribution, the ratio of power transferred to the ions and the electrons vary little with the minority (hydrogen) concentration n(H)/n(e) but the high-Z impurity screening provided by the fast ions temperature increases with the concentration. The power radiated by tungsten in the core of the JET discharges has been analyzed on a large database covering the 2013-2014 campaign. In the baseline scenario with moderate plasma current (I-p. =. 2.5 MA) ICRH modifies efficiently tungsten transport to avoid its accumulation in the plasma centre and, when the ICRH power is increased, the tungsten radiation peaking evolves as predicted by the neo-classical theory. At higher current (3-4MA), tungsten accumulation can be only avoided with 5MW of ICRH power with high gas injection rate. For discharges in the hybrid scenario, the strong initial peaking of the density leads to strong tungsten accumulation. When this initial density peaking is slightly reduced, with an ICRH power in excess of 4 MW, very low tungsten concentration in the core (similar to 10(-5)) is maintained for 3 s. MHD activity plays a key role in tungsten transport and modulation of the tungsten radiation during a sawtooth cycle is correlated to the fishbone activity triggered by the fast ion pressure gradient.

This paper presents a laboratory-scale experimental investigation about the laser cleaning of diagnostic first mirrors from tokamak-like contaminants, made of oxidized tungsten compounds with different properties and morphology. The re-deposition of contaminants sputtered from a tokamak first wall onto first mirrors' surfaces could dramatically decrease their reflectivity in an unacceptable way for the proper functioning of plasma diagnostic systems. The laser cleaning technique has been proposed as a solution to tackle this issue. In this work, pulsed laser deposition was exploited to produce rhodium films functional as first mirrors and to deposit onto them contaminants designed to be realistic in reproducing materials expected to be re-deposited on first mirrors in a tokamak environment. The same laser system was also used to perform laser cleaning experiments, exploiting a sample handling procedure that allows one to clean some cm(2) in a few minutes. Cleaning effectiveness was evaluated in terms of specular reflectance recovery and mirror surface integrity. The effect of different laser wavelengths (lambda = 1064, 266 nm) on the cleaning process was also addressed, as well as the impact of multiple contamination/ cleaning cycles on the process outcome. A satisfactory recovery of pristine mirror reflectance (>= 90%) was obtained in the vis-NIR spectral range, avoiding at the same time mirror damaging. The results here presented show the potential of the laser cleaning technique as an attractive solution for the cleaning of diagnostic first mirrors.

New experiments at JET with the ITER-like wall show for the first time that ITER-relevant low field side resonance first harmonic ion cyclotron resonance heating (ICRH) can be used to control sawteeth that have been initially lengthened by fast particles. In contrast to previous (Graves et al 2012 Nat. Commun. 3 624) high field side resonance sawtooth control experiments undertaken at JET, it is found that the sawteeth of L-mode plasmas can be controlled with less accurate alignment between the resonance layer and the sawtooth inversion radius. This advantage, as well as the discovery that sawteeth can be shortened with various antenna phasings, including dipole, indicates that ICRH is a particularly effective and versatile tool that can be used in future fusion machines for controlling sawteeth. Without sawtooth control, neoclassical tearing modes (NTMs) and locked modes were triggered at very low normalised beta. High power H-mode experiments show the extent to which ICRH can be tuned to control sawteeth and NTMs while simultaneously providing effective electron heating with improved flushing of high Z core impurities. Dedicated ICRH simulations using SELFO, SCENIC and EVE, including wide drift orbit effects, explain why sawtooth control is effective with various antenna phasings and show that the sawtooth control mechanism cannot be explained by enhancement of the magnetic shear. Hybrid kinetic-magnetohydrodynamic

Evidence for low field side accumulation of tungsten is often observed in bolometry and soft x-ray emissivities of highly rotating JET ITER-like wall (ILW) plasmas. Poloidal variation of the density of high-Z impurities, such as tungsten, in the core of NBI heated plasmas is expected from neoclassical theory due to charge displacement and parallel electric field generated by the centrifugal force. We calculate the poloidally asymmetric distribution of tungsten using fluid equations and a 1D transport simulation with the JETTO/SANCO code. Peaking of tungsten on the outboard side of the plasma is found and verified with soft x-ray and bolometry measurements. We then study the effect of a poloidally asymmetric tungsten distribution on the distribution of the NBI heat source by simulations with the Monte Carlo code ASCOT. The simulations show that the poloidally asymmetric tungsten profile redistributes the fast NBI ions radially through shifting their ionization profile and poloidally through enhanced pitch-angle scattering at high energy. The amplitude of the redistribution is in the order of 10% for the highest n(W)/n(e) ratios of similar to 10(-4) measured in recent JET H-mode plasmas. As a result of the scattering of the beam particles, the core heat deposition is changed less than 10%, which does not have a significant impact to the performance of JET plasmas. The modelling is in qualitative agreement with measurements by the vertical neutron camera that sees a broadening in the 2.5 MeV neutron profile when tungsten peaks on the outboard side of the plasma.

The need for materials with special and often unique functionalities is particularly felt in forefront domains, particularly in areas where materials are subjected to extreme erosion and/or corrosion. Coating technologies are becoming more widespread since they help to improve existing materials and products separating the structural performances from the surface-protection performances [1]. Refractory metals and their alloys constitute a class of particularly interesting materials because, thanks to their outstanding properties, they can be used in hazardous environments and extreme conditions. In this work, we present an investigation on W and Ta based coatings on stainless steel substrates. Among many different applications, tungsten is used in electronic industry for x-ray optics and in harsh environments such as thermonuclear fusion reactors (tokamaks) where it is generally considered to be the most promising plasma facing material (PFM). This is due to its high cohesive energy, which makes it resistant to erosion and ensures an high sputtering threshold under hydrogen bombardment, its high melting point (the highest one when considering refractory metals), low coefficient of thermal expansion, good electrical conductivity, low affinity with hydrogen and excellent thermal conductivity. Unfortunately tungsten finds just a few applications in the standard industrial fields because of the low stability of its oxides (WO3, WO2). Tantalum has properties that make it useful for many applications, from electronics to mechanical and chemical systems. Its high melting point, toughness, low ductile-to-brittle transition temperature and exceptional resistance to chemical attack make it an attractive coating material for components exposed to high temperature, wear, and harsh chemical environments. Its excellent resistance against corrosion is ensured by the formation of a stable passive oxide film (Ta2O5). Tantalum is also used to produce a variety of alloys that have high melting points, are strong and have good ductility. In particular W is alloyed with Ta to lower brittle-to-ductile transition temperature [2,3]. W rich alloys (W-Ta) are not very well known and, to the authors' knowledge, there are just a few papers in literature. On the contrary Ta rich alloys (Ta-W) are well known materials and they excel due to their good mechanical properties and excellent corrosion resistance. With the 10 % of W the resulting alloy is 1.4 times stronger than pure tantalum but it remains easy to work even at high temperature (up to 1 600 °C). Physical vapor deposition (PVD) is one of the most promising coating technology and it is widely employed to improve mechanical, wear and corrosion properties of materials. Moreover it is possible to tailor coating features (crystallinity, grain orientation, etc.) to obtain properties that can be far apart from the bulk material ones. In this work, different films, with different W and Ta contents, have been produced via PVD using Direct Current Magnetron Sputtering (DCMS) and Pulsed Laser Deposition (PLD) [4]. Moreover pure W and Ta samples have been deposited by High-Power Impulse Magnetron Sputtering (HiPIMS) [5,6]. For these samples the various deposition parameters have not been fully optimized. However, the preliminary results look very promising. DC Magnetron Sputtering is a traditional PVD technique extensively used in industry. PLD is a relatively new technique that permits to deposit films with very complex stoichiometry (e.g. YBCO). Moreover it is able to deposit nanostructured materials tailoring film features at the nanoscale. HiPIMS is an innovative technique which utilizes magnetron sputtering cathodes and high peak power density of up to 3 kW cm-2 on the target. The plasma produces a particle flux with high degree of ionization. HiPIMS has been successfully used to enhance coating adhesion. It produces high-density microstructure films with smooth surfaces. Moreover, it has a few industrial applications in hard, electronic, and optical coatings. The produced refractory metal based coatings have been extensively analyzed. Microstructural characterization activities consisted of scanning electron microscopy (SEM) and X-ray diffraction (XRD). The nano-mechanical properties of the films (hardness and elastic modulus) have been analyzed by nanoindentation testing. Adhesion has been finally evaluated by scratch tests [7], using a fully-computerized UMT apparatus. The electrochemical properties of thin films have been also evaluated in 3.5 wt.% NaCl aqueous solution [8]. The goal of this work was to evaluate the properties of high tech nanostructured coatings of refractory metals. Moreover it was possible to stress how to drive the film characteristics needed for the specific application using three different techniques.

One of the aims of the ITER Tokamak is to demonstrate the feasibility of a prolonged fusion power production based on a deuterium (D)- tritium (T) plasma. Plasma facing materials (PFMs) are a key issue for this objective. ITER will use tungsten (W) in the divertor baffle and other regions, which will be subject to high fluxes of energetic particles. Although, W is considered the most suitable material for the divertor region, owing to its high sputtering threshold and melting point, the possible retention of tritium in tungsten needs to be assessed for safety requirements. Due to the plasma interaction, W coatings in the divertor could have different morphologies and/or nanostructures. The present research activity aims to study different nanostructured W coatings exposed to high-flux D ion to evaluate the hydrogen retention properties.

The modelling of a controlled tungsten dust injection experiment in TEXTOR by the dust dynamics code MIGRAINe is reported. The code, in addition to the standard dust-plasma interaction processes, also encompasses major mechanical aspects of dust-surface collisions. The use of analytical expressions for the restitution coefficients as functions of the dust radius and impact velocity allows us to account for the sticking and rebound phenomena that define which parts of the dust size distribution can migrate efficiently. The experiment provided unambiguous evidence of long-distance dust migration; artificially introduced tungsten dust particles were collected 120o toroidally away from the injection point, but also a selectivity in the permissible size of transported grains was observed. The main experimental results are reproduced by modelling.

Reaction of the acidic tungsten(II) hydride 2 with the nickel(II) pincer complex 1 in either THF or toluene after an initial dihydrogen bonding (DHB) interaction led to the formation of the Ni-W bimetallic species 3 (see picture). The first example of DHB between two metal hydrides with opposite polarity was analyzed by NMR and IR spectroscopy, X-ray crystallography, and DFT calculations.