Table of contents

Impurity seeding of noble gases is an effective way of decreasing the heat loads onto the divertor targets in fusion devices. To investigate the effect of noble gases on deuterium retention, tungsten targets have been implanted by different noble gas ions and subsequently exposed to deuterium plasma. Irradiation induced defects and deuterium retention in tungsten targets have been characterized by positron annihilation Doppler broadening and thermal desorption spectroscopy. Similar defect distributions are observed in tungsten irradiated by neon and argon, while it is comparatively low in the case of helium. The influence of helium pre-irradiation on deuterium trapping is found to be small based on the desorption spectrum compared with that of the pristine one. Neon and argon pre-irradiation leads to an enhancement of deuterium trapping during plasma exposure. The influence on deuterium retention is found to be argon > neon > helium when comparing at a similar crystal damage level.

DEMO is the name for the first stage prototype fusion reactor considered to be the next step after ITER towards realizing fusion. For the realization of fusion energy especially, materials questions pose a significant challenge already today. Heat, particle and neutron loads are a significant problem to material lifetime when extrapolating to DEMO. For many of the issues faced, advanced materials solutions are under discussion or already under development. In particular, components such as the first wall and the divertor of the reactor can benefit from introducing new approaches such as composites or new alloys into the discussion. Cracking, oxidation as well as fuel management are driving issues when deciding for new materials. Here ${{\rm{W}}}_{{\rm{f}}}/{\rm{W}}$ composites as well as strengthened CuCrZr components together with oxidation resilient tungsten alloys allow the step towards a fusion reactor. In addition, neutron induced effects such as transmutation, embrittlement and after-heat and activation are essential. Therefore, when designing a component an approach taking into account all aspects is required.

After commissioning in 2005, the EAST superconducting tokamak had been operated with its water cooled divertors for eight campaigns up to 2012, employing graphite as plasma facing material. With increase in heating power over 20 MW in recent years, the heat flux going to the divertors rises rapidly over 10 MW m⁻² for steady state operation. To accommodate the rapid increasing heat load in EAST, the bolting graphite tile divertor must be upgraded. An ITER-like tungsten (W) divertor has been designed and developed; and firstly used for the upper divertor of EAST. The EAST upper W divertor is modular structure with 80 modules in total. Eighty sets of W/Cu plasma-facing components (PFC) with each set consisting of an outer vertical target (OVT), an inner vertical target (IVT) and a DOME, are attached to 80 stainless steel cassette bodies (CB) by pins. The monoblock W/Cu-PFCs have been developed for the strike points of both OVT and IVT, and the flat type W/Cu-PFCs for the DOME and the baffle parts of both OVT and IVT, employing so-called hot isostatic pressing (HIP) technology for tungsten to CuCrZr heat sink bonding, and electron beam welding for CuCrZr to CuCrZr and CuCrZr to other material bonding.

Both monoblock and flat type PFC mockups passed high heat flux (HHF) testing by means of electron beam facilities. The 80 divertor modules were installed in EAST in 2014 and results of the first commissioning are presented in this paper.

The current state of knowledge of the mechanical and thermal properties of high-strength, high conductivity Cu alloys relevant for fusion energy high heat flux applications is reviewed, including effects of thermomechanical and joining processes and neutron irradiation on precipitation- or dispersion-strengthened CuCrZr, Cu–Al₂O₃, CuNiBe, CuNiSiCr and CuCrNb (GRCop-84). The prospects for designing improved versions of wrought copper alloys and for utilizing advanced fabrication processes such as additive manufacturing based on electron beam and laser consolidation methods are discussed. The importance of developing improved structural materials design criteria is also noted.

The ITER baseline scenario, with 500 MW of DT fusion power and Q = 10, will rely on a Type I ELMy H-mode and will be achieved with a tungsten (W) divertor. W atoms sputtered from divertor targets during mitigated ELMs are expected to be the dominant source in ITER. W impurity concentration in the plasma core can dramatically degrade its performance and lead to potentially damaging disruptions. Understanding the physics of the target W source due to sputtering during ELMs and inter-ELMs is important and can be helped by experimental measurements with improved precision. It has been established that the ELMy target ion impact energy has a simple linear dependence with the pedestal electron temperature measured by Electron Cyclotron Emission (ECE). It has also been shown that Langmuir Probes (LP) ion flux measurements are reliable during ELMs due to the surprisingly low electron temperature. Therefore, in this paper, LP and ECE measurements in JET-ITER-Like-Wall (ILW) unseeded Type I ELMy H-mode experiments have been used to estimate the W sputtering flux from divertor targets in ELM and inter-ELM conditions. Comparison with similar estimates using W I spectroscopy measurements shows a reasonable agreement for the ELM and inter-ELM W source. The main advantage of the method involving LP measurements is the very high time resolution of the diagnostic (∼10 μs) allowing very precise description of the W sputtering source during ELMs.

For the next step fusion reactor the use of tungsten is inevitable to suppress erosion and allow operation at elevated temperature and high heat loads. Tungsten fibre-reinforced composites overcome the intrinsic brittleness of tungsten and its susceptibility to operation embrittlement and thus allow its use as a structural as well as an armour material. That this concept works in principle has been shown in recent years. In this contribution we present a development approach towards its use in a future fusion reactor. A multilayer approach is needed addressing all composite constituents and manufacturing steps. A huge potential lies in the optimization of the tungsten wire used as fibre. We discuss this aspect and present studies on potassium doped tungsten wire in detail. This wire, utilized in the illumination industry, could be a replacement for the so far used pure tungsten wire due to its superior high temperature properties. In tensile tests the wire showed high strength and ductility up to an annealing temperature of 2200 K. The results show that the use of doped tungsten wire could increase the allowed fabrication temperature and the overall working temperature of the composite itself.

Stress-relieved pure tungsten received three damage levels (0.10, 0.25 and 0.50 dpa) by self-tungsten ion beam irradiation at room temperature. Positron annihilation spectroscopy showed the formation of mono-vacancies and vacancy clusters after ion beam exposure. In the first irradiation step (0–0.10 dpa) some splitting up of large vacancy clusters occurred which became more numerous. For increasing dose to 0.25 dpa, growth of the vacancy clusters was seen. At 0.50 dpa a change in the defect formation seems to occur leading to a saturation in the lifetime signal obtained from the positrons. Nano-indentation on the cross-sections showed a flat damage depth distribution profile. The nano-indentation hardness increased for increasing damage dose without any saturation up to 0.50 dpa. This means that other defects such as dislocation loops and large sized voids seem to contribute.

Polycrystalline tungsten samples were characterized and exposed to a pure H beam or mixed H/He beam containing 6% He in GLADIS at a surface temperature of 600 °C, 1000 °C, or 1500 °C. After 5400 s of exposure time with a heat flux of 10.5 MW m⁻², the total accumulated fluence of 2 × 10²⁵ m⁻² was reached. Thereafter, edge localized mode (ELM)-like thermal shocks with a duration of 1 ms and an absorbed power density of 190 MW m⁻² and 380 MW m⁻² were applied on the samples in JUDITH 1. During the thermal shocks, the base temperature was kept at 1000 °C. The ELM-experiments with the lowest transient power density did not result in any detected damage. The other tests showed the beginning of crack formation for every sample, except the sample pre-exposed with the pure H-beam at 1500 °C in GLADIS. This sample was roughened, but did not show any crack initiation. With exception to the roughened sample, the category of ELM-induced damage for the pre-exposed samples is identical to the reference tests without pre-exposure to a particle flux.

Tritium distributions on the W-coated divertor tiles used with Be wall in JET 2011–2012 ITER-like wall (JET-ILW) campaign were measured using an imaging plate (IP) technique. The high intensity of photo-stimulated luminescence (PSL) from IP was observed at the regions covered by deposited Be layers. However, the PSL intensity was not simply proportional to the thickness of the deposited Be layers; the shadowed region of Tile 4 showed the highest PSL intensity though the thickness of deposited Be layer on this region was smaller than that on Tile 0 and the apron of Tile 1 by an order of magnitude. These observations indicated the influence of impurities such as oxygen on tritium retention in the deposited Be layers. The C tiles used in the 2007–2009 JET carbon wall (JET-C) campaign were also examined. The high PSL intensity was observed for the regions covered with deposited C layers in this case. The area of tile surfaces covered by the deposited tritium-rich layers on the W-coated-tiles used in the JET-ILW campaign was significantly smaller than that on the C tiles used in the JET-C campaign.

Tritium imaging plate technique (TIPT) has been applied to examine tritium (T) retention in individual particles made of titanium (Ti) with 30 and 100 μm in diameter and tungsten (W) with 50 μm in diameter. Distribution of T radioactivity observed by TIPT corresponded well to spatial distribution of the particles. In a limited case of uniform and high T concentration in the bulk of the individual particle, the amount of T is directly quantified from T radioactivity by a master curve method. Density and size of the particle and T concentration profiles in the bulk of the particle are important factors to change emission behavior of T β-ray and thus accurate quantification of the amount of T in the individual particle.

Density functional theory (DFT) studies show that in tungsten a mono vacancy can contain up to six hydrogen isotopes (HIs) at 300 K with detrapping energies varying with the number of HIs in the vacancy. Using these predictions, a multi trapping rate equation model has been built and used to model thermal desorption spectrometry (TDS) experiments performed on single crystal tungsten after deuterium ions implantation. Detrapping energies obtained from the model to adjust temperature of TDS spectrum observed experimentally are in good agreement with DFT values within a deviation below 10%. The desorption spectrum as well as the diffusion of deuterium in the bulk are rationalized in light of the model results.

The primary goal of the WEST project is to be a test bed to characterize the fatigue and lifetime of ITER-like W divertor components subjected to relevant thermal loads. During the first phase of exploitation (S2 2016), these components (W monoblock plasma facing unit—W-PFU) will be installed in conjunction with graphite components (G-PFU). Since the G-PFU will not be actively cooled, it is necessary to ensure the expected pulse duration allows the W-PFU to reach its steady state without overheating the G-PFU assembly structure or the embedded stainless-steel diagnostics. High heat flux tests were performed at the GLADIS facility to assess the thermal behavior of the G-PFU. Some operational limits based on plasma parameters were determined. It was found that it is possible to operate at an injected power such that the maximal incident heat flux on the lower divertor is 10 MW m⁻² for the required pulse length.

A model system representing the RAFM steel EUROFER-97 is produced by magnetron sputter deposition of iron and 1.5 at% tungsten and investigated in order to study the consequences of plasma exposures. The alloy is deposited as coatings with a thickness of 400 nm on polycrystalline, high purity iron substrates. To understand the erosion mechanisms and morphology changes the coatings were exposed to a linear plasma device with an ion flux of 3×10²¹ D⁺ m⁻² s⁻¹ and an electron temperature of 13 eV. Samples were exposed at sample temperatures of about 420 and 770 K at incident ion energy of 30 eV (floating potential), 70 and 190 eV. Additionally, the effect of ion fluence was investigated. The coatings before and after plasma exposure were investigated by electron microscopy and glow discharge optical emission spectroscopy (GD-OES). Microstructure observation revealed a complex morphology with distinct sharp spikes formed under the plasma exposure at incident ion energies of 70 and 190 eV. The tungsten enrichment by a factor of 3 in the spikes was visualized by backscatter electron observation and confirmed by both energy-dispersive x-ray spectroscopy and GD-OES. No visible erosion and, by that, tungsten enrichment was observed after the plasma exposure at an incident ion energy of 30 eV, as expected since it is below the threshold energy for sputtering of iron.

Grazing incidence small angle x-ray scattering was performed on tungsten samples exposed to helium plasma in the MAGPIE and Pisces-A linear plasma devices to measure the size distributions of resulting helium nano-bubbles. Nano-bubbles were fitted assuming spheroidal particles and an exponential diameter distribution. These particles had mean diameters between 0.36 and 0.62 nm. Pisces-A exposed samples showed more complex patterns, which may suggest the formation of faceted nano-bubbles or nano-scale surface structures.

The electron beam device JUDITH 1 was used to establish a testing procedure for the qualification of tungsten as plasma facing material. Absorbed power densities of 0.19 and 0.38 GW m⁻² for an edge localized mode-like pulse duration of 1 ms were chosen. Furthermore, base temperatures of room temperature, 400 °C and 1000 °C allow investigating the thermal shock performance in the brittle, ductile and high temperature regime. Finally, applying 100 pulses under all mentioned conditions helps qualifying the general damage behaviour while with 1000 pulses for the higher power density the influence of thermal fatigue is addressed. The investigated reference material is a tungsten product produced according to the ITER material specifications. The obtained results provide a general overview of the damage behaviour with quantified damage characteristics and thresholds. In particular, it is shown that the damage strongly depends on the microstructure and related thermo-mechanical properties.

First results are presented in relation with experimental and theoretical studies performed at the CORIA laboratory in the general framework of the determination of the chemical analysis of Tokamak plasma facing materials by laser-induced breakdown spectroscopy (LIBS) in picosecond regime. Experiments are performed on W in a specific chamber. This chamber is equipped with a UV-visible-near IR spectroscopic device. Boltzmann plots are derived for typical laser characteristics. We show that the initial excitation temperature is close to 12 000 K followed by a quasi steady value close to 8500 K. The ECHREM (Euler code for CHemically REactive Multicomponent laser-induced plasmas) code is developed to reproduce the laser-induced plasmas. This code is based on the implementation of a Collisional-Radiative model in which the different excited states are considered as full species. This state-to-state approach is relevant to theoretically assess the departure from excitation and chemical equilibrium. Tested on aluminum, the model shows that the plasma remains close to excitation equilibrium.

It has long been recognized that volumetric dissipation of the plasma heat flux from a fusion power system is preferable to its localized impingement on a material surface. Volumetric dissipation mitigates both the anticipated very high heat flux and intense particle-induced damage due to sputtering. Recent projections to a tokamak demonstration power plant suggest an immense upstream parallel heat flux, of order 20 GW m⁻², implying that fully detached operation may be a requirement for the success of fusion power. Building on pioneering work on the use of lithium by Nagayama et al and by Ono et al as well as earlier work on the gas box divertor by Watkins and Rebut, we present here a concept for a lithium vapor box divertor, in which lithium vapor extracts momentum and energy from a fusion-power-plant divertor plasma, using fully volumetric processes. At the high powers and pressures that are projected this requires a high density of lithium vapor, which must be isolated from the main plasma in order to avoid lithium build-up on the chamber walls or in the plasma. Isolation is achieved through a powerful multi-box differential pumping scheme available only for condensable vapors. The preliminary box-wise calculations are encouraging, but much more work is required to demonstrate the practical viability of this scheme, taking into account at least 2D plasma and vapor flows within and between the vapor boxes and out of the vapor boxes to the main plasma.

The damage produced by primary knock-on atoms (PKA) in W has been investigated from the threshold displacement energy (TDE) where it produces one self interstitial atom–vacancy pair to larger energies, up to 100 keV, where a large molten volume is formed. The TDE has been determined in different crystal directions using the Born–Oppenheimer density functional molecular dynamics (DFT-MD). A significant difference has been observed without and with the semi-core electrons. Classical MD has been used with two different empirical potentials characterized as 'soft' and 'hard' to obtain statistics on TDEs. Cascades of larger energy have been calculated, with these potentials, using a model that accounts for electronic losses (Sand et al 2013 Europhys. Lett.103 46003). Two other sets of cascades have been produced using the binary collision approximation (BCA): a Monte Carlo BCA using SDTrimSP (Eckstein et al 2011 SDTrimSP: Version 5.00. Report IPP 12/8) (similar to SRIM www.srim.org) and MARLOWE (RSICC Home Page. (https://rsicc.ornl.gov/codes/psr/psr1/psr-137.html) (accessed May, 2014)). The comparison of these sets of cascades gave a recombination distance equal to 12 Å which is significantly larger from the one we reported in Hou et al (2010 J. Nucl. Mater.403 89) because, here, we used bulk cascades rather than surface cascades which produce more defects (Stoller 2002 J. Nucl. Mater.307 935, Nordlund et al 1999 Nature398 49). Investigations on the defect clustering aspect showed that the difference between BCA and MD cascades is considerably reduced after the annealing of the cascade debris at 473 K using our Object Kinetic Monte Carlo model, LAKIMOCA (Domain et al 2004 J. Nucl. Mater.335 121).

We report measurements of the energy dependence of flux thresholds and incubation fluences for He-ion induced nano-fuzz formation on hot tungsten surfaces at UHV conditions over a wide energy range using real-time sample imaging of tungsten target emissivity change to monitor the spatial extent of nano-fuzz growth, corroborated by ex situ SEM and FIB/SEM analysis, in conjunction with accurate ion-flux profile measurements. The measurements were carried out at the multicharged ion research facility (MIRF) at energies from 218 eV to 8.5 keV, using a high-flux deceleration module and beam flux monitor for optimizing the decel optics on the low energy MIRF beamline. The measurements suggest that nano-fuzz formation proceeds only if a critical rate of change of trapped He density in the W target is exceeded. To understand the energy dependence of the observed flux thresholds, the energy dependence of three contributing factors: ion reflection, ion range and target damage creation, were determined using the SRIM simulation code. The observed energy dependence can be well reproduced by the combined energy dependences of these three factors. The incubation fluences deduced from first visual appearance of surface emissivity change were (2–4) × 10²³ m⁻² at 218 eV, and roughly a factor of 10 less at the higher energies, which were all at or above the displacement energy threshold. The role of trapping at C impurity sites is discussed.

The High Heat Flux Test Facility (HHFTF) was designed and established recently at Institute for Plasma Research (IPR) in India for testing heat removal capability and operational life time of plasma facing materials and components of the ITER-like tokamak. The HHFTF is equipped with various diagnostics such as IR cameras and IR-pyrometers for surface temperature measurements, coolant water calorimetry for absorbed power measurements and thermocouples for bulk temperature measurements. The HHFTF is capable of simulating steady state heat load of several MW m⁻² as well as short transient heat loads of MJ m⁻². This paper presents the current status of the HHFTF at IPR and high heat flux tests performed on the curved tungsten monoblock type of test mock-ups as well as transient heat flux tests carried out on pure tungsten materials using the HHFTF. Curved tungsten monoblock type of test mock-ups were fabricated using hot radial pressing (HRP) technique. Two curved tungsten monoblock type test mock-ups successfully sustained absorbed heat flux up to 14 MW m⁻² with thermal cycles of 30 s ON and 30 s OFF duration. Transient high heat flux tests or thermal shock tests were carried out on pure tungsten hot-rolled plate material (Make:PLANSEE) with incident power density of 0.49 GW m⁻² for 20 milliseconds ON and 1000 milliseconds OFF time. A total of 6000 thermal shock cycles were completed on pure tungsten material. Experimental results were compared with mathematical simulations carried out using COMSOL Multiphysics for transient high heat flux tests.

The theoretical model describing spatiotemporal dynamics of He clusters in tungsten, which takes into account He trap generation associated with the growth of He clusters, is presented. Application of this model to the formation of the layer of nano-bubbles underneath of the surface of thick He irradiated sample, before surface morphology starts to change, gives very good agreement with currently available experimental data. The role of thermophoresis in a long-term evolution of nano-bubble containing structures is discussed.

A successful joint between W and EUROFER using high temperature brazing technique has been achieved for structural application in future fusion power plants. Cu-based powder alloy mixed with a polymeric binder has been used as filler. Microstructural analysis of the joints revealed that the joint consisted mainly of primary phases and acicular structures in a Cu matrix. Interaction between EUROFER and filler took place at the interface giving rise to several Cu–Ti–Fe rich layers. A loss of hardness at the EUROFER substrate close to the joint due to a diffusion phenomenon during brazing cycle was measured; however, the joints had an adequate shear strength value.

The simultaneous effects of sputtering, implantation and solid-state diffusion determine the surface profiles of mixed-material systems under ion-bombardment at elevated temperatures due to the enhanced atomic mobility. To simulate the joint processes the Monte Carlo code SDTrimSP for the simulation of the ion–solid interaction has been augmented by a diffusion model for solid-state diffusion. The combined model has been applied to a tungsten–iron system under deuterium bombardment as model system for EUROFER. The simulation results reveal a strong dependence of the surface profile on initial tungsten concentration, ion energy, temperature and fluence but also on the impinging flux, a parameter which is often not appropriately taken into account. For reactor relevant parameters of low-energy (200 eV) deuterium fluxes of ${10}^{21}\mathrm{at}\;{{\rm{m}}}^{-2}\;{{\rm{s}}}^{-1}$ at 873 K a tungsten–iron system exhibits an increase of the tungsten surface concentration from initially 1% by a factor of more than 20, which drops at lower fluxes.

The experimental fusion reactor ITER, currently under construction in Cadarache, France, is transferring the nuclear fusion research to the power plant scale. ITER's first wall (FW), armoured by beryllium, is subjected to high steady state and transient power loads. Transient events like edge localized modes not only deposit power densities of up to 1.0 GW m⁻² for 0.2–0.5 ms in the divertor of the machine, but also affect the FW to a considerable extent. Therefore, a detailed study was performed, in which transient power loads with absorbed power densities of up to 1.0 GW m⁻² were applied by the electron beam facility JUDITH 1 on beryllium specimens at base temperatures of up to 300 °C. The induced damage was evaluated by means of scanning electron microscopy and laser profilometry. As a result, the observed damage was highly dependent on the base temperatures and absorbed power densities. In addition, five different classes of damage, ranging from 'no damage' to 'crack network plus melting', were defined and used to locate the damage, cracking, and melting thresholds within the tested parameter space.

The recycling of D ions impinging onto a W divertor surface is a key input parameter into the power and momentum balance at the target boundary during SOL modeling. It is described by the ratio R of the flux of recombining D₂ molecules to the non-reflected incident ion flux. In steady-state plasmas where the surface is in equilibrium with the incident flux, R equals one due to particle conservation. However, during transient events such as edge localized modes (ELMs) the evolution of R with time is not straightforward to predict. Therefore, detailed diffusion-trapping calculations were performed taking into account the variations in power influx and particle energy during an ELM. They showed that in contrast to the naive expectation, that the ELM would deplete the surface and subsequently lead to 'pumping' (R$\ll$ 1) of the incident flux by the empty surface, R ≈ 1 or even R$\gt 1$ occurs. This paper will first describe how the ELM was approximated in the 1D diffusion-trapping code and then discuss the evolution of R during an ELM and in the inter ELM phase. Also, an analytical picture of R will be developed which allows qualitatively understanding the evolution of R as calculated by the diffusion-trapping code.

We have investigated net and gross erosion of W in the outer strike-point (OSP) region of ASDEX Upgrade with the help of marker probes during low-density/high-temperature L-mode discharges. Post mortem analyses indicate net-erosion rates of 0.04–0.13 nm s⁻¹, with the highest rates measured close to the OSP. Re-deposition was some 30–40% of gross erosion, which is lower than what has earlier been obtained spectroscopically (∼50–60%), possibly due to the special plasma conditions of our experiment and intense flux of W atoms originating from the main chamber. Gross erosion was also estimated by passive emission spectroscopy, and around the OSP the results matched with post mortem data. However, the spectroscopic erosion profile in the poloidal direction was much steeper than the post mortem one. Preliminary ERO simulations have predicted net erosion of the same order of magnitude as experimental results but reproducing the poloidal erosion/re-deposition profiles requires further work.

We report for the first time on the ability of Raman microscopy to give information on the structure and composition of Be related samples mimicking plasma facing materials that will be found in ITER. For that purpose, we investigate two types of material. First: Be, W, Be₁W₉, and Be₅W₅ deposits containing a few percents of D or N, and second: a Mo mirror exposed to plasma in the main JET chamber (in the framework of the first mirror test in JET with ITER-like wall). We performed atomic quantifications using ion beam analysis for the first samples. We also did atomic force microscopy. We found defect induced Raman bands in Be, Be₁W₉, and Be₅W₅ deposits. Molybdenum oxide has been identified showing an enhancement due to a resonance effect in the UV domain.

In the present paper marker structures consisting of W/Mo layers were deposited on bulk W samples by using a modified CMSII method. This technology, compared to standard CMSII, prevents the formation of nano-pore structures at interfaces. The thicknesses of the markers were in the range 20–35 μm to balance the requirements associated with the wall erosion in ITER and thermo-mechanical performances. The coatings structure and composition were evaluated by glow discharge optical emission spectrometry (GDOES), and energy dispersive x-ray spectroscopy measurements (EDX). The adhesion of the coatings to the substrate has been assessed by scratch test method. In order to evaluate their effectiveness as potential markers for fusion applications, the marker coatings have been tested in an electron beam facility at a temperature of 1000 °C and a power density of about 3 MW m⁻². A number of 300 pulses with duration of 420 s (35 testing hours) were applied on the marker coated samples.

The W—for tungsten—Environment in Steady-state Tokamak (WEST) project is based on an upgrade of the Tore Supra tokamak from a carbon limiter to an X-point divertor device. A new set of actively cooled tungsten coated plasma facing components will cover a part of the vessel to provide a fully metallic environment. This paper deals with the validation program performed for tungsten coatings (≥15 μm) on a CuCrZr substrate. The first step was dedicated to the qualification under high heat flux tests of the coating on small inertially cooled samples. To study the thermal behavior and the non-uniformity, the second step was dedicated to the validation of the coating on large inertially cooled samples with geometry and shape (540 × 120 mm) representative of the WEST coated components. The last step was dedicated to the optimization of the coating and to the high heat flux tests up to 10.5 MW m⁻² on relevant coated actively cooled prototypes. Non-uniformity and thickness of the coating (15 and 30 μm) correspond to specifications. As no delamination was observed, coatings of 15 and 30 μm were qualified with regard to their application on WEST coated components. In order to decrease the risk of coating delamination under thermal loading, it was decided to cover the upper divertor and baffle targets with the thinnest coating option of 15 μm.

We have performed microstructural characterization using transmission electron microscopy (TEM) techniques to reveal nanometric features in the sub-surface region of tungsten samples exposed to high flux, low energy deuterium plasma. TEM examination revealed formation of a dense dislocation network and dislocation tangles, overall resulting in a strong increase in the dislocation density by at least one order of magnitude as compared to the initial one. Plasma-induced dislocation microstructure vanishes beyond a depth of about 10 μm from the top of the exposed surface where the dislocation density and its morphology becomes comparable to the reference microstructure. Interstitial edge dislocation loops with Burgers vector a₀/2〈111〉 and a₀〈100〉 were regularly observed within 6 μm of the sub-surface region of the exposed samples, but absent in the reference material. The presence of these loops points to a co-existence of nanometric D bubbles, growing by loop punching mechanism, and sub-micron deuterium flakes, resulting in the formation of surface blisters, also observed here by scanning electron microscopy.

The influence of the annealing temperature on deuterium retention was studied for self-ion damaged tungsten in the range from 600–1200 K. Samples were damaged by 20 MeV W ions at room temperature to the peak damage level of 0.5 dpa. Samples were then annealed at the desired temperature for 1 h and exposed to deuterium atom beam with the flux of $2.6\times {10}^{19}\ {\rm{D}}/{{\rm{m}}}^{2}\;{\rm{s}}$ for 144 h to populate the remaining defects. An unannealed sample was also used as a reference. Nuclear reaction analysis technique was used for deuterium depth profile analysis and thermal desorption was performed on the same samples to measure the amount of total retained D. Scanning transmission electron microscopy was used for the calculation of dislocation densities in the samples. After annealing at 1200 K approximately 66% of those initial defects which retain deuterium were annealed.

Tungsten (W) surfaces are analyzed with laser-induced breakdown spectroscopy (LIBS). Interactions of W with nanosecond (ns) and femtosecond (fs) laser pulses are found to be quite different in terms of the ambient Ar gas pressure dependence of the average ablation rate and W I line intensity. Collinear double-pulse LIBS (115 + 115 mJ) using two ns lasers (with interpulse separation Δt₁₂ = 5.32 μs) improves the signal-to-noise ratio over the whole Ar pressure range P_Ar = 6.7 × 10⁻¹ − 6.7 × 10⁴ Pa in contrast with single-pulse LIBS (SP-LIBS) with 230 mJ, where a signal enhancement by a factor of ∼2–3 is obtained only at P_Ar > 10³ Pa. SP-LIBS with a ns laser has succeeded in obtaining a sharp transition between thin W layer with a thickness of ∼100 nm and the graphite substrate. A He I (587.5 nm) line has been successfully detected with SP-LIBS with a ns laser from W containing He bubbles (∼20–30 nm layers) in the near-surface region.

Transient heat pulses with triangular, square, and ELM-like temporal shapes are investigated in order to further understand how transient plasma instabilities will affect plasma facing components in tokamaks. A solution to the 1D heat equation for triangular pulses allows the peak surface temperature to be written analytically for arbitrary rise times. The solution as well as ANSYS simulations reveal that a positive ramp (maximum rise time) triangular pulse has a higher peak surface temperature by a factor of $\sqrt{2}$ compared to that from a negative ramp (rise time = 0) pulse shape with equal energy density, peak power, and pulse width. Translating the results to ITER, an ELM or disruption pulse with the shortest rise time is the most benign compared to other pulse shapes with the same peak heat flux and same energy density.

Laser based methods are investigated as in situ diagnostic for plasma facing materials (PFMs) in magnetic fusion research to study PFM composition and retention. In laser induced ablation spectroscopy (LIAS) the wall material is ablated by a laser beam. The released material enters the edge plasma region of a fusion experiment and the resulting optical emission is observed. To conclude from the observed photons to the number of ablated atoms, a detailed knowledge of the velocity distribution of the ablated material is required. In this work the LIAS emission in discharges at TEXTOR was studied using an Ametek Phantom v711 camera. In this paper a method is developed to conclude from the observed emission the velocity distribution of the ablated species. The obtained velocity distribution is used for our numerical LIAS model, demonstrating good agreement with our experimental observations. Implications are discussed.

Radio-frequency (RF)-enhanced surface erosion of beryllium (Be) plasma-facing components is explored, for the first time, using the ERO code. The code is applied to measured, RF-enhanced edge Be line emission at JET Be outboard limiters, in the presence of high-power, ion cyclotron-resonance heating (ICRH) in L-mode discharges. In this first modelling study, the RF sheath effect from an ICRH antenna on a magnetically connected, limiter region is simulated by adding a constant potential to the local sheath, in an attempt to match measured increases in local Be I and Be II emission of factors of 2–3. It was found that such increases are readily simulated with added potentials in the range of 100–200 V, which is compatible with expected values for potentials arising from rectification of sheath voltage oscillations from ICRH antennas in the scrape-off layer plasma. Absolute erosion values are also estimated within the uncertainties in local plasma conditions.

Thin tungsten oxide layers with thicknesses up to 250 nm have been formed on W surfaces by thermal oxidation following a parabolic growth rate. The reflectance of the layers in the IR range 2.5–16 μm has been measured showing a decrease with the layer thickness especially at low wavelengths. Raman microscopy and x-ray diffraction show a nanocrystalline WO₃ monoclinic structure. Low energy deuterium plasma exposure (11 eV/D⁺) has been performed inducing a phase transition, a change in the sample colour and the formation of tungsten bronze (D_xWO₃). Implantation modifies the whole layer suggesting a deep diffusion of deuterium inside the oxide. After exposure, a deuterium release due to the oxidation of D_xWO₃ under ambient conditions has been evidenced showing a reversible deuterium retention.

Temperature dependence on deuterium (D) retention for He⁺ implanted tungsten (W) was studied by thermal desorption spectroscopy (TDS) to evaluate the tritium retention behavior in W. The activation energies were evaluated using Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code and found to be 0.55 eV, 0.65 eV, 0.80 eV and 1.00 eV. The heating scenarios clearly control the D retention behavior and, dense and large He bubbles could work as a D diffusion barrier toward the bulk, leading to D retention enhancement at lower temperature of less than 430 K, even if the damage was introduced by He⁺ implantation. By comparing the D retention for W, W with carbon deposit and tungsten carbide (WC), the dense carbon layer on the surface enhances the dynamic re-emission of D as hydrocarbons, and induces the reduction of D retention. However, by He⁺ implantation, the D retention was increased for all the samples.

Finite element modeling analysis has been employed to simulate the melt layer motion of tungsten and tungsten-based materials under high magnetic field. High heat flux of 2 GW m⁻² was loaded for 3 ms at 1000 K and provided a molten bath. Meanwhile, high magnetic field from 0 to 8 T was loaded during the simulation. Both positive and negative surface tension temperature coefficient was tested. The result shows that the convention forced by the surface tension is suppressed by the magnetic field. The high magnetic field performs as a resistance of the heat transfer, leading to a reduced molten bath. The magnetic field mitigates the melting behaviur of the tungsten materials.

A recently developed numerical model, based on the dislocation-driven nucleation of gas bubbles, is used to analyse experimental results on deuterium retention in tungsten under ITER relevant plasma exposure conditions. Focus is placed on understanding the relation between exposure temperature and flux on primary features of thermal desorption spectra: peak positions and intensities of the desorption flux. The model allows one to relate the peak positions with the size of plasma induced deuterium bubbles and envisage exposure conditions (temperature and flux) for their formation. Based on the performed analysis, dedicated experimental conditions to validate the model are proposed.

Exposure of tungsten to low energy (<100 eV) helium plasmas at temperatures between 900–1900 K in both laboratory experiments and tokamaks has been shown to cause severe nanoscale modification of the near surface resulting in the growth of tungsten tendrils. Tendril formation can lead to non-sputtered erosion and dust formation. Here we report on characterization of a compact electron cyclotron resonance (ECR) He plasma source with an ion flux of ∼2.5 × 10¹⁹ ions m⁻² s⁻¹, average fluence of 3 × 10²⁴ ions m⁻², and the surface morphology changes seen on the exposed tungsten surfaces. Exposures of polished tungsten disks at temperatures up to 1270 K have been performed and characterized using scanning electron microscopy and atomic force microscopy (AFM) scans. Bubbles and craters have been seen on the exposed tungsten surface growing to up to 150 nm in diameter. The ECR source has been tested for eventual use on a scanning tunneling microscopy experiment intended to study the early stages of surface morphology change due to He ion exposure.

Self-passivating tungsten based alloys will provide a major safety advantage compared to pure tungsten when used as first wall armor of future fusion reactors, due to the formation of a protective oxide layer which prevents the formation of volatile and radioactive WO₃ in case of a loss of coolant accident with simultaneous air ingress. Bulk WCr10Ti2 alloys were manufactured by two different powder metallurgical routes: (1) mechanical alloying (MA) followed by hot isostatic pressing (HIP) of metallic capsules, and (2) MA, compaction, pressureless sintering in H₂ and subsequent HIPing without encapsulation. Both routes resulted in fully dense materials with homogeneous microstructure and grain sizes of 300 nm and 1 μm, respectively. The content of impurities remained unchanged after HIP, but it increased after sintering due to binder residue. It was not possible to produce large samples by route (2) due to difficulties in the uniaxial compaction stage. Flexural strength and fracture toughness measured on samples produced by route (1) revealed a ductile-to-brittle-transition temperature (DBTT) of about 950 °C. The strength increased from room temperature to 800 °C, decreasing significantly in the plastic region. An increase of fracture toughness is observed around the DBTT.

The effects of addition of ∼3% He⁺ in simultaneous D–He irradiation at various D and He ion energies were studied for polycrystalline W at 300 and 500 K. Combinations of 250–750 eV/D⁺ and 500–1000 eV/He⁺ were used to vary the D and He ion range relative to each other. Total D and He retention were measured by thermal desorption spectroscopy up to 1473 K, and select specimens implanted at 500 K were analyzed by nuclear reaction analysis and electron recoil detection analysis. At both 300 and 500 K, D retention was reduced and trapping changed due to the addition of He⁺; however, consistent with the literature, D and He diffused well beyond the ion ranges. Furthermore, varying the ion ranges had little effect on D retention, depth profile, and trapping. D diffusion into the bulk was reduced from far beyond 7 μm to less than 2 μm with the addition of He⁺.

Estimating the tritium amount retained in the plasma facing components and their surface layer composition is of crucial importance for ITER. Laser-induced breakdown spectroscopy (LIBS) is an analytical technique suitable for in situ measurements of both these quantities. For improving its sensitivity, the double pulse (DP) variant can be used, instead of the standard single pulse (SP). In this work Mo samples coated with 1.5–1.8 μm thick W–Al (as a proxy for Be) mixed layer, with co-deposited deuterium were analyzed under vacuum (∼5 × 10⁻⁵ mbar) by SP and DP LIBS, showing enhancement of the spectral intensity for the latter. Calibration free method was applied to the LIBS data for getting the elemental concentration of W and Al. Results are in satisfactory agreement with those obtained from preliminary, ion beam analysis measurements. Deuterium concentration was tentatively estimated by accounting for the intensity ratio between Dα and nearby WI lines.

Nanostructured W surfaces prepared by He bombardment exhibit characteristic angular distributions of hydrogen ion reflection upon injection of 1 keV H⁺ beam. A magnetic momentum analyzer that can move in the vacuum chamber has measured the angular dependence of the intensity and the energy of reflected ions. Broader angular distributions were observed for He-irradiated tungsten samples compared with that of the intrinsic polycrystalline W. Both intensity and energy of reflected ions decreased in the following order: the polycrystalline W, the He-bubble containing W, and the fuzz W. Classical trajectory Monte Carlo simulations based on Atomic Collision in Amorphous Target code suggests that lower atom density near the surface can make the reflection coefficients lower due to increasing number of collisions.

The vacuum UV (VUV)-near Infrared (NIR) laser induced breakdown spectroscopy (LIBS) technique was applied to investigate the composition of W-based samples with a protective carbon layer. The sample was analyzed under pressures from 5 to 10⁵ Pa and atmosphere (air, He). The spectra were recorded with three spectrometers at delays from 200 ns to 10 μs at atmospheric pressures and from 100 to 500 ns at low pressures. The electron density was determined from the measured spectra using Stark broadening and the electron temperature from the W I–W III Saha–plot in the VUV–NIR spectral range. The better precision was achieved due to usage W III spectral lines of tungsten. The achieved results are more reliable than results obtained without W III spectral lines. The calibration free LIBS method was then applied to determine the W and C contents of the analyzed sample.

The impact on the deuterium retention of simultaneous exposure of tungsten to a steady-state plasma and transient cyclic heat loads has been studied in the linear PSI-2 facility with the main objective of qualifying tungsten (W) as plasma-facing material. The transient heat loads were applied by a high-energy laser, a Nd:YAG laser (λ = 1064 nm) with an energy per pulse of up to 32 J and a duration of 1 ms. A pronounced increase in the D retention by a factor of 13 has been observed during the simultaneous transient heat loads and plasma exposure. These data indicate that the hydrogen clustering is enhanced by the thermal shock exposures, as seen on the increased blister size due to mobilization and thermal production of defects during transients. In addition, the significant increase of the D retention during the simultaneous loads could be explained by an increased diffusion of D atoms into the W material due to strong temperature gradients during the laser pulse exposure and to an increased mobility of D atoms along the shock-induced cracks. Only 24% of the retained deuterium is located inside the near-surface layer (d<4 μm). Enhanced blister formation has been observed under combined loading conditions at power densities close to the threshold for damaging. Blisters are not mainly responsible for the pronounced increase of the D retention.

Infrared (IR) images of the ITER wide angle viewing system are modeled for the baseline plasma equilibrium and partially detached tungsten divertor, taking into account the three-dimensional structure of the first wall and the divertor. The modeling includes a comprehensive chain of calculations from the heat load specifications up to the synthetic, reflection-free IR images of the surface temperature, T_surf. The effect of the optical blur due to finite IR detector size and diffraction/aberrations—approximated by a Gaussian filter—on the measured T_surf is investigated. The optical blur characterized by σ = 0.7 pixel (approximately twice the diffraction limit) leads to underestimation of T_surf,max on the inner vertical divertor target and near the upper X-point by <6% and <4%, respectively. This is within the required measurement accuracy of 10%. Larger underestimation of T_surf,max (<12%) is observed on the outer vertical divertor target. The study demonstrates the importance of keeping the performance of the optical system as close as possible to the diffraction limit.

The kinetics of tungsten carbide formation was investigated for tungsten coatings on carbon fibre composite with a molybdenum interlayer as they are used in the ITER-like Wall in JET. The coatings were produced by combined magnetron sputtering and ion implantation. The investigation was performed by preparing focused ion beam cross sections from samples after heat treatment in argon atmosphere. Baking of the samples was done at temperatures of 1100 °C, 1200 °C, and 1350 °C for hold times between 30 min and 20 h. It was found that the data can be well described by a diffusional random walk with a thermally activated diffusion process. The activation energy was determined to be (3.34 ± 0.11) eV. Predictions for the isothermal lifetime of this coating system were computed from this information.

As a complementary method to Rutherford back scattering (RBS), glow discharge optical emission spectrometry (GDOES) was used to investigate the depth profiles of W, Mo, Be, O and C concentrations into marker coatings (CFC/Mo/W/Mo/W) and the substrate of divertor tiles up to a depth of about 100 μm. A number of 10 samples cored from particular areas of the divertor tiles were analyzed. The results presented in this paper are valid only for those areas and they cannot be extrapolated to the entire tile. Significant deposition of Be was measured on Tile 3 (near to the top), Tile 6 (at about 40 mm from the innermost edge) and especially on Tile 0 (HFGC). Preliminary experiments seem to indicate a penetration of Be through the pores and imperfections of CFC material up to a depth of 100 μm in some cases. No erosion and a thin layer of Be (<1 μm) was detected on Tiles 4, 7 and 8. On Tile 1 no erosion was found at about 1/3 from bottom.

x-ray micro-laminography was qualified and implemented as a complementary solution for the 3D microstructural analysis of tungsten coated carbon-fibre reinforced carbon (W/CFC) samples retrieved from JET ITER-like wall. As expected, the W layers spatially correlate with the morphology of the CFC substrate. Three main cases were distinguished; (i) tungsten layers coated parallel to PAN fibre bundles tend to have a quasi-continuous, weakly waved surface (waves amplitude <100 μm); (ii) tungsten layers coated onto the relatively porous felt region appear to smoothly follow even the surface of the largest pores of around 250 μm and (iii) samples coated perpendicular to the PAN fibre bundles display frequently and strong crater-like discontinuities of the metal layer. The characteristics dimensions of these gaps range in the order of 300–400 μm both in the coating plane and perpendicular to it. On some craters the bottom W layer is broken and the generated debris can be found even deeper than one mm into the CFC substrate. These W particles, sized of 20–40 μm, are always found in the large gaps located between the fibre bundles perpendicular to the coated surface.

Erosion and deposition were studied in the JET divertor during the first JET ITER-like wall campaign 2011 to 2012 using marker tiles. An almost complete poloidal section consisting of tiles 0, 1, 3, 4, 6, 7, 8 was studied. The data from divertor tile surfaces were completed by the analysis of samples from remote divertor areas and from the inner wall cladding. The total mass of material deposited in the divertor decreased by a factor of 4–9 compared to the deposition of carbon during all-carbon JET operation before 2010. Deposits in 2011 to 2012 consist mainly of beryllium with 5–20 at.% of carbon and oxygen, respectively, and small amounts of Ni, Cr, Fe and W. This decrease of material deposition in the divertor is accompanied by a decrease of total deuterium retention inside the JET vessel by a factor of 10 to 20. The detailed erosion/deposition pattern in the divertor with the ITER-like wall configuration shows rigorous changes compared to the pattern with the all-carbon JET configuration.

Rotating collectors and quartz microbalances (QMBs) are used in JET to provide time-dependent measurements of erosion and deposition. Rotation of collector discs behind apertures allows recording of the long term evolution of deposition. QMBs measure mass change via the frequency deviations of vibrating quartz crystals. These diagnostics are used to investigate erosion/deposition during JET-C carbon operation and JET-ILW (ITER-like wall) beryllium/tungsten operation. A simple geometrical model utilising experimental data is used to model the time-dependent collector deposition profiles, demonstrating good qualitative agreement with experimental results. Overall, the JET-ILW collector deposition is reduced by an order of magnitude relative to JET-C, with beryllium replacing carbon as the dominant deposit. However, contrary to JET-C, in JET-ILW there is more deposition on the outer collector than the inner. This reversal of deposition asymmetry is investigated using an analysis of QMB data and is attributed to the different chemical properties of carbon and beryllium.

The broad array of expected loading conditions in a fusion reactor such as ITER necessitates high requirements on the plasma facing materials (PFMs). Tungsten, the PFM for the divertor region, the most affected part of the in-vessel components, must thus sustain severe, distinct exposure conditions. Accordingly, comprehensive experiments investigating sequential and simultaneous thermal and particle loads were performed on double forged pure tungsten, not only to investigate whether the thermal and particle loads cause damage but also if the sequence of exposure maintains an influence. The exposed specimens showed various kinds of damage such as roughening, blistering, and cracking at a base temperature where tungsten could be ductile enough to compensate the induced stresses exclusively by plastic deformation (Pintsuk et al 2011 J. Nucl. Mater.417 481–6). It was found out that hydrogen has an adverse effect on the material performance and the loading sequence on the surface modification.

Tungsten button samples were exposed to He ELMing H-mode plasma in DIII-D using 2.3 MW of electron cyclotron heating power. Prior to the exposures, the W buttons were exposed to either He, or D, plasma in PISCES-A for 2000 s at surface temperatures of 225–850 °C to create a variety of surfaces (surface blisters, subsurface nano-bubbles, fuzz). Erosion was spectroscopically measured from each DiMES sample, with the exception of the fuzzy W samples which showed almost undetectable WI emission. Post-exposure grazing incidence small angle x-ray scattering surface analysis showed the formation of 1.5 nm diameter He bubbles in the surface of W buttons after only a single DIII-D (3 s, ∼150 ELMs) discharge, similar to the bubble layer resulting from the 2000 s. exposure in PISCES-A. No surface roughening, or damage, was detected on the samples after approximately 600 ELMs with energy density between 0.04–0.1 MJ m⁻².

Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with little obvious damage except in the areas where unipolar arcing occurred. Arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces.

Powder injection molding (PIM) was used to produce pure and particle reinforced W materials to be qualified for the use as plasma facing material. As alloying elements La₂O₃, Y₂O₃, TiC, and TaC were chosen with a particle size between 50 nm and 2.5 μm, depending on the alloying element. The fabrication of alloyed materials was done for different compositions using powder mixtures. Final sintering was performed in H₂ atmosphere at 2400 °C resulting in plates of 55 × 22 × 4 mm³ with ∼98% theoretical density. The qualification of the materials was done via high heat flux testing in the electron beam facility JUDITH-1. Thereby, ELM-like 1000 thermal shock loads of 0.38 GW m⁻² for 1 ms and 100 disruption like loads of 1.13 GW m⁻² for 1 ms at a base temperature of 1000 °C were applied. The obtained damage characteristics, i.e. surface roughening and crack formation, were qualified versus an industrially manufactured pure reference tungsten material and linked to the material's microstructure and mechanical properties.

JET components are removed periodically for surface analysis to assess material migration and fuel retention. This paper describes issues related to handling JET components and procedures for preparing samples for analysis; in particular a newly developed procedure for cutting beryllium tiles is presented. Consideration is also given to the hazards likely due to increased tritium inventory and material activation from 14 MeV neutrons following the planned TT and DT operations (DTE2) in 2017. Conclusions are drawn as to the feasibility of handling components from JET post DTE2.

Volatile compounds of tungsten (WF₆) and molybdenum (MoF₆) were used as tracers of high-Z metal migration in the TEXTOR tokamak in several gas injection experiments when puffing was done through a test limiter. The experiments with W were performed prior to major shut-downs, while the MoF₆ injection was followed by the final shutdown in connection with TEXTOR decommissioning. In all cases a set of surface probes and limiter tiles was retrieved and analysed with electron and ion beam techniques. The focus was on the local deposition in the vicinity of the gas inlet and in the inlet system. Depth profiles in the deposits and metal distribution maps clearly show that only near the gas inlet significant amounts of Mo are deposited along the scrape-off layer flow and E × B drift directions, which could be reproduced by ERO-code modelling. Correlation between the injection scenario and the deposition patterns is presented.

The work presents results of a broad TEXTOR dust survey in terms of its composition, structure, distribution and fuel content. The dust particles were collected after final shutdown of TEXTOR in December 2013. Fuel retention, as determined by thermal desorption, varied significantly, even by two orders of magnitude, dependent on the dust location in the machine. Dust structure was examined by means of scanning electron microscopy combined with energy-dispersive x-ray spectroscopy, focused ion beam and scanning transmission electron microscopy. Several categories of dust have been identified. Carbon-based stratified and granular deposits were dominating, but the emphasis in studies was on metal dust. They were found in the form of small particles, small spheres, flakes and splashes which formed 'comet'-like structures, clearly indicating directional effects in the impact on surfaces of plasma-facing components. Nickel-rich alloys from the Inconel liner and iron-based ones from various diagnostic holders were the main components of metal-containing dust, but also molybdenum and tungsten debris were detected. Their origin is discussed.

In the next fusion devices, all the plasma facing components will consist of bulk tungsten or tungsten coating on carbon. This paper focuses on the behaviour of tungsten coated on carbon fibre composite designed for the WEST project (Bucalossi et al 2011 Fusion Eng. Des.86 684–688) under intensive thermal cycling delivered by an electron beam. The use of scanning electron microscope has allowed in particular, the observation of several pore lines inside the coating. These pore lines have different aspects depending on the observed zone according to the localisation of the electron beam, accentuated lines with more numerous enlarged pores in zone exposed to the electron beam. An analogous trend is also observed for JET tungsten-coated samples under similar thermal cycles despite their different properties due to an alternative manufacturing method of the substrate. A systematic and attentive comparison on the coating changes after the application of the electron beam heating is presented. The observed comportments as the formation of the pore lines or the pore shapes are assumed to be inherent to simultaneous diffusion processes. In association with the pore line formation, a migration of the carbon substrate towards the surface is presumed and discussed.

Surface samples from a full poloidal set of divertor tiles exposed in JET through operations 2010–2012 with ITER-like wall have been investigated using SEM, SIMS, ICP-AES analysis and micro beam nuclear reaction analysis (μ-NRA). Deposition of Be and retention of D is microscopically inhomogeneous. With careful overlaying of μ-NRA elemental maps with SEM images, it is possible to separate surface roughness effects from depth profiles at microscopically flat surface regions, without pits. With (³He, p) μ-NRA at 3–5 MeV beam energy the accessible depth for D analysis in W is about 9 μm, sufficient to access the W/Mo and Mo/W interfaces in the coatings and beyond, while for Be in W it is about 6 μm. In these conditions, at all plasma wetted surfaces, D was found throughout the whole accessible depth at concentrations in the range 0.2–0.7 at% in W. Deuterium was found to be preferentially trapped at the W/Mo and Mo/W interfaces. Comparison is made with SIMS profiling, which also shows significant D trapping at the W/Mo interface. Mixing of Be and W occurs mainly in deposited layers.

In order to understand the interaction mechanisms between hydrogenic species and beryllium co-deposits, a 1D Diffusion Trapping Model of Isotopic eXchange in Be (DITMIX) is developed. Hydrogen depth profiles from DITMIX are in good agreement with those measured by ¹⁵N-NRA on pre-characterised 600 nm thick Be:H layers (H/Be = 0.04), which were irradiated by D ions with a low flux of 10¹⁷ m⁻² s⁻¹ and an energy of 5 keV D⁻¹, for different fluences and surface temperatures. Hence DITMIX provides a qualitative understanding of the isotope exchange mechanisms, although modelled versus measured D profiles show less agreement in the bulk, casting some doubt on the processes involved. For such low fluxes, DITMIX shows that the main factors determining isotopic exchange are the irradiation fluence and the surface temperature.

Many ideas for liquid surface PFCs are for divertors. First walls are likely to be more challenging technologically because long flow paths are necessary for fast flowing systems and the first wall must be an integral structure with the blanket. Maximum tolerable heat loads are a critical concern. This paper describes several processes at work in walls with fast-flowing or slow-flowing liquid plasma-facing surfaces, and the considerations imposed by heat transfer and the power balance for the PFC as well as the structure needed for an integrated first wall and blanket, and uses thermal modeling of a generic PFC structure to illustrate the issues and support the conclusions.

Simulations of impurity trajectories in deuterium plasmas in the vicinity of the surface are performed by taking into account the magnetic sheath in conditions relevant for ITER and WEST. We show that the magnetic sheath has a strong effect on the average impact angle of impurities in divertor conditions and that it can lead to an increase of $\approx 60\%$ at the gross erosion maximum for neon (Ne⁺⁴) compared to the case when only the cyclotron motion is considered. The evaluation of the net erosion has been undertaken by retaining local redeposition of tungsten (W). We investigate how it is affected by the sheath magnetic potential profile. The largest effect is however observed when an energy distribution is considered. In this case the number of particles that manage to exit the sheath is larger as it is dominated by the more energetic particles. The comparison with other work is also discussed. The application to a scenario of the WEST project is finally performed, which exhibits a moderate, however non negligible, erosion of the plasma facing components.

Energy transfer processes from ELM-like pulsed helium (He) plasmas with a pulse duration of ∼0.1 ms to aluminum (Al) and tungsten (W) surfaces were experimentally investigated by the use of a magnetized coaxial plasma gun device. The surface absorbed energy density of the He pulsed plasma on the W surface measured with a calorimeter was ∼0.44 MJ m⁻², whereas it was ∼0.15 MJ m⁻² on the Al surface. A vapor layer in front of the Al surface exposed to the He pulsed plasma was clearly identified by Al neutral emission line (Al i) measured with a high time resolution spectrometer, and fast imaging with a high-speed visible camera filtered around the Al i emission line. On the other hand, no clear evaporation in front of the W surface exposed to the He pulsed plasma was observed in the present condition. Discussions on the reduction in the surface absorbed energy density on the Al surface are provided by considering the latent heat of vaporization and radiation cooling due to the Al vapor cloud.

After a brief introduction giving some insight of the WEST project, we present the three types of plasma facing units (PFUs) developed for the WEST project taking into account the envisaged main scenarios: (1) high power short pulse scenario (a few seconds) where the objective is to maximize the power handling of the PFUs, up to 20 MW m⁻², (2) high fluence scenario (a few 100 s) on actively cooled ITER-like tungsten (W) PFUs, up to 10 MW m⁻² during 1000 s. For the graphite PFUs, the high heat flux tests have been done at GLADIS (ion beam test facility), and for the CuCrZr PFUs on the JUDITH (electron beam test facility). The tests were successful, as no damage occurred for the different load cases. This confirms that the modelling done during the design phase is appropriate to describe these PFUs. Series productions are expected to be achieved by the end of 2015 for the graphite and CuCrZr PFUs, and few ITER-like W PFUs are expected at the beginning of 2016. The lower divertor will be complemented with ITER-like W PFUs as soon as available from our partners so that different fabrication procedures could be evaluated in a real industrial process and a real tokamak environment.

Deuterium and helium retention in Japanese reduced activation ferritic martensitic (RAFM) steel (F82H) under simultaneous D–He irradiation at 500, 625, 750, and 818 K was studied. This study aims to clarify tritium retention behavior in RAFM steels to assess their use as plasma facing materials. The irradiation fluence was kept constant at 1 × 10²⁴ D m⁻². Four He desorption peaks were observed with He retention greatest at 625 K. At T > 625 K a monotonic decrease in He retention was observed. At all temperatures a systematic reduction in D retention was observed for the simultaneous D–He case in comparison to D-only case. This suggests that He implanted at the near surface in RAFM steels may reduce the inward penetration of tritium in RAFM steels that would result in lower tritium inventory for a given fluence.

Deuterium (D) retention behaviours for 14 MeV neutron irradiated tungsten (W) and fission neutron irradiated W were evaluated by thermal desorption spectroscopy (TDS) to elucidate the correlation between D retention and defect formation by different energy distributions of neutrons in W at the initial stage of fusion reactor operation. These results were compared with that for Fe²⁺ irradiated W with various damage concentrations. Although dense vacancies and voids within the shallow region near the surface were introduced by Fe²⁺ irradiation, single vacancies with low concentration were distributed throughout the sample for 14 MeV neutron irradiated W. Only the dislocation loops were introduced by fission neutron irradiation at low neutron fluence. The desorption peak of D for fission neutron irradiated W was concentrated at low temperature region less than 550 K, but that for 14 MeV neutron irradiated W was extended toward the higher temperature side due to D trapping by vacancies. It can be said that the neutron energy distribution could have a large impact on irradiation defect formation and the D retention behaviour.

Cleaning systems of metallic first mirrors are needed in more than 20 optical diagnostic systems from ITER to avoid reflectivity losses. Currently, plasma sputtering is considered as one of the most promising techniques to remove deposits coming from the main wall (mainly beryllium and tungsten). This work presents the results of plasma cleaning of rhodium and molybdenum mirrors exposed in JET-ILW and contaminated with typical tokamak elements (including beryllium and tungsten). Using radio frequency (13.56 MHz) argon or helium plasma, the removal of mixed layers was demonstrated and mirror reflectivity improved towards initial values. The cleaning was evaluated by performing reflectivity measurements, scanning electron microscopy, x-ray photoelectron spectroscopy and ion beam analysis.

In October 2014, JET completed a scoping study involving high power scenario development in preparation for DT along with other experiments critical for ITER. These experiments have involved intentional and unintentional melt damage both to bulk beryllium main chamber tiles and to divertor tiles. This paper provides an overview of the findings of concern for machine protection in JET and ITER, illustrating each case with high resolution images taken by remote handling or after removal from the machine. The bulk beryllium upper dump plate tiles and some other protection tiles have been repeatedly flash melted by what we believe to be mainly fast unmitigated disruptions. The flash melting produced in this way is seen at all toroidal locations and the melt layer is driven by j × B forces radially outward and upwards against gravity. In contrast, the melt pools caused while attempting to use MGI to mitigate deliberately generated runaway electron beams are localized to several limiters and the ejected material appears less influenced by j × B forces and shows signs of boiling. In the divertor, transient melting of bulk tungsten by ELMs was studied in support of the ITER divertor material decision using a specially prepared divertor module containing an exposed edge. Removal of the module from the machine in 2015 has provided improved imaging of the melt and this confirms that the melt layers are driven by ELMs. No other melt damage to the other 9215 bulk tungsten lamellas has yet been observed.

In this paper, production and study of tokamak relevant W particles is presented. Existing tokamak-produced dust being very scarce, extensive study of such nano-particles with high specific surface area (SSA) and sub micron size requires a specific and efficient alternative production technique, in order to obtain relevant particles to study. We present our production and collection setup based on pulsed laser ablation on bulk ITER-grade tungsten, and the various parameters impacting on the collected dust morphology and properties. We observed that optimum gas pressure is required to control the laser-induced plasma properties and favour the production of tungsten nano-particles with high SSA. The laser pulse duration is also a key parameter to limit the generation of tungsten liquid droplets during the ablation process. The nano-particules structure and general aspect are characterized via scanning electron microscopy and transmission electron microscopy. Lastly, this dust produced by laser ablation is loaded with tritium by gas exposure, and its retention capability and long-term evolution addressed and compared to metallurgically produced W powders with homogeneous size distribution.

ITER first wall (FW) panels are irradiated by energetic neutrons during the nuclear phase. Thus, an irradiation and high heat flux testing programme is undertaken by the ITER organization in order to evaluate the effects of neutron irradiation on the performance of enhanced heat flux (EHF) FW components. The test campaign includes neutron irradiation (up to 0.6–0.8 dpa at 200 °C–250 °C) of mock-ups that are representative of the final EHF FW panel design, followed by thermal fatigue tests (up to 4.7 MW m⁻²). Mock-ups were manufactured by the same manufacturing process as proposed for the series production. After a pre-irradiation thermal screening, eight mock-ups will be selected for the irradiation campaigns. This paper reports the preparatory work of HHF tests and neutron irradiation, assessment results as well as a brief description of mock-up manufacturing and inspection routes.

A thermal wall model is developed for the SolEdge2D-EIRENE edge transport code for calculating the surface temperature of the actively-cooled vessel components in interaction with the plasma. This is a first step towards a self-consistent evaluation of the recycling of particles, which depends on the wall surface temperature. The proposed thermal model is built to match both steady-state temperature and time constant of actively-cooled plasma facing components. A benchmark between this model and the Finite Element Modelling code CAST3M is performed in the case of an ITER–like monoblock. An example of application is presented for a SolEdge2D-EIRENE simulation of a medium-power discharge in the WEST tokamak, showing the steady-state wall temperature distribution and the temperature cycling due to an imposed Edge Localised Mode-like event.

A selected set of samples from JET-ILW divertor tiles exposed in 2011–2012 has been analysed using thermal desorption spectrometry (TDS). The highest amount of deuterium was found on the regions with the thickest deposited layers, i.e. on the horizontal (apron) part and on the top part of Tile 1, which resides deep in the scrape-off layer. Outer divertor Tiles 6, 7 and 8 had nearly an order of magnitude less deuterium. The co-deposited layers on the JET tiles and the W coatings contain C, O and Ni impurities which may change the desorption properties. The D₂ signals in the TDS spectra were convoluted and the positions of the peaks were compared with the Be and C amounts but no correlations between them were found. The remaining fractions of D in the analysed samples at ITER baking temperature 350 °C are rather high implying that co-deposited films may be difficult to be de-tritiated.

Post-mortem studies with ion beam analysis, thermal desorption, and secondary ion mass spectrometry have been applied for investigating the long-term fuel retention in the JET ITER-like wall components. The retention takes place via implantation and co-deposition, and the highest retention values were found to correlate with the thickness of the deposited impurity layers. From the total amount of retained D fuel over half was detected in the divertor region. The majority of the retained D is on the top surface of the inner divertor, whereas the least retention was measured in the main chamber on the mid-plane of the inner wall limiter. The recessed areas of the inner wall showed significant contribution to the main chamber total retention. Thermal desorption spectroscopy analysis revealed the energetic T from DD reactions being implanted in the divertor. The total T inventory was assessed to be $\gt 0.3\;{\rm{mg}}$.

Experiments in the JET tokamak equipped with the ITER-like wall (ILW) revealed that the inner and outer target plate at the location of the strike points represent after one year of operation intact tungsten (W) surfaces without any beryllium (Be) surface coverage. The dynamics of near-surface retention, implantation, desorption and recycling of deuterium (D) in the divertor of plasma discharges are determined by W target plates. As the W plasma-facing components (PFCs) are not actively cooled, the surface temperature (T_surface) is increasing with plasma exposure, varying the balance between these processes in addition to the impinging deuteron fluxes and energies. The dynamic behaviour on a slow time scale of seconds was quantified in a series of identical L-mode discharges (JET Pulse Number (JPN)$\;\#\;81938-73$) by intra-shot gas analysis providing the reduction of deuterium retention in W PFCs by 1/3 at a base temperature (T_base) range at the outer target plate between 65 °C and 150 °C equivalent to a T_surface span of 150 °C and 420 °C. The associated recycling and molecular D desorption during the discharge varies only at lowest temperatures moderately, whereas desorption between discharges rises significantly with increasing T_base. The retention measurements represent the sum of inner and outer divertor interaction at comparable T_surface. The dynamic behaviour on a fast time scale of ms was studied in a series of identical H-mode discharges (JPN $\;\#83623-83974$) and coherent edge-localized mode (ELM) averaging. High energetic ELMs of about 3 keV are impacting on the W PFCs with fluxes of $3\times {10}^{23}\;{{\rm{D}}}^{+}\;{{\rm{s}}}^{-1}{{\rm{m}}}^{-2}$ which is about four times higher than inter-ELM ion fluxes with an impact energy of about E_im = 200 eV. This intra-ELM ion flux is associated with a high heat flux of about 60 MW m⁻² to the outer target plate which causes T_surface rise by Δ T = 100 K per ELM covering finally the range between 160 °C and 1400 °C during the flat-top phase. ELM-induced desorption from saturated near-surface implantation regions as well as deep ELM-induced deuterium implantation areas under varying baseline temperature takes place. Subsequent refuelling by intra-ELM deuteron fluxes occurs and a complex interplay between deuterium fuelling and desorption can be observed in the temporal ELM footprint of the surface temperature (IR thermography), the impinging deuteron flux (Langmuir probes), and the Balmer radiation (emission spectroscopy) as representative for the deuterium recycling flux. In contrast to JET-C, a pronounced second peak, ≃ 8 ms delayed with respect to the initial ELM crash, in the D_α radiation and the ion flux has been observed. The peak can be related to desorption of implanted energetic intra-ELM D⁺ diffusing to the W surface, and performing local recycling.

Global gas balance experiments at ASDEX Upgrade (AUG) and JET have shown that a considerable fraction of nitrogen injected for radiative cooling is not recovered as N₂ upon regeneration of the liquid helium cryo pump. The most probable loss channels are ion implantation into plasma-facing materials, co-deposition and ammonia formation. These three mechanisms are investigated in laboratory and tokamak experiments and by numerical simulations. Laboratory experiments have shown that implantation of nitrogen ions into beryllium and tungsten leads to the formation of surface nitrides, which may decompose under thermal loads. On beryllium the presence of nitrogen at the surface has been seen to reduce the sputtering yield. On tungsten surfaces it has been observed that the presence of nitrogen can increase hydrogen retention. The global nitrogen retention in AUG by implantation into the tungsten surfaces saturates. At JET the steady state nitrogen retention is increased by co-deposition with beryllium. The tokamak experiments are interpreted in detail by simulations of the global migration with WallDYN. Mass spectrometry of the exhaust gas of AUG and JET has revealed the conversion of nitrogen to ammonia at percent-levels. Conclusions are drawn on the potential implications of nitrogen seeding on the operation of a reactor in a deuterium–tritium mix.

An overview is given on the recent progress on edge modelling activities for the JET ITER-like wall using the computational tools like the SOLPS or EDGE2D-EIRENE code. The validation process of these codes on JET with its metallic plasma-facing components is an important step towards predictive studies for ITER and DEMO in relevant divertor operational conditions, i.e., for detached, radiating divertors. With increased quantitative credibility in such codes more reliable input to plasma-wall and plasma-material codes can be warranted, which in turn results in more realistic and physically sound estimates of the life-time expectations and performance of a Be first-wall and a W-divertor, the same materials configuration foreseen for ITER. A brief review is given on the recent achievements in the plasma–wall interaction and material migration studies. Finally, a short summary is given on the availability and development of integrated codes to assess the performance of an JET-ILW baseline scenario also in view of the preparation for a JET DT-campaign.

The migration of wall material or seeding impurities plays an important role in the formation of mixed materials, the impurity contamination of the plasma and tritium retention. First, this work presents an improved model for the sputtering from mixed material surfaces in WallDYN. Second, we present dynamic SDTrimSP and WallDYN simulations of the nitrogen implantation in Be and the migration of nitrogen in tokamaks with a Be main wall. The simulations with the binary collision code SDTrimSP predict that N accumulates directly at the surface and that the Be erosion decreases with increasing N surface content. A first application of WallDYN to the nitrogen migration with an ITER-like wall indicates that the Be main wall may cause wall pumping of N by co-deposition with Be.