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In 2021 JET exploited its unique capabilities to operate with T and D–T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D–T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements—new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors—as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D–T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D–T plasmas to date and expanding our understanding of isotopes and D–T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D–T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D–T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D–T operations is presented in (King et al Nucl. Fusion submitted).

The fracture behavior of plasma-facing components (PFCs) under extreme plasma-material interaction conditions is of great concern to ITER and future fusion reactors. This was explored in the current study by exposing pure tungsten (W), W-1%TiC and W-2%Y2O3 composites to a combined steady-state/transient hydrogen plasma up to a base surface temperature of ~2220 K, and up to 5000 transient pulses for 1000 s using the linear plasma generator Magnum-PSI. The applied heat loads were characterized by combining sheath physics, thermographic information and finite element analyses, with which the thermal stress was evaluated. Combining microstructural investigation and thermo-mechanical numerical analyses, a physical picture of fracture is developed. The transient heat loads drive surface crack initiation, whose depth can be estimated by a simple analytical model for pure tungsten, while the cooling period following the steady-state heat load induces tensile stresses, opening existing surface cracks deeper. The fracture process is mediated by the microstructure whereby the ceramic particles stabilize the microstructure but promote surface crack initiation due to suppressed plasticity at the grain boundaries and the particle-matrix interfaces. The surface cracks relieve the subsequent cycles of transient thermal stress but intensify the steady-state thermal stress, therefore, promoting deep crack propagation. These results help to understand failure mechanisms in PFCs under extreme operation conditions which are valuable for developing advanced PFCs. [ABSTRACT FROM AUTHOR]

In the ITER tokamak, injection of nitrogen is foreseen to decrease the heat loads on the divertor surfaces. However, once dissociated, nitrogen atoms react with hydrogen isotopes to form ammonia isotopologues. The formation of tritiated ammonia may pose some issues with regards to tritium inventory and operation duty cycle. In this paper, we report a study of the effect of three parameters of relevance for the fusion environment on the ammonia production, including the presence of a catalytic surface, sample temperature and noble gas addition. Results of ammonia formation from N2/H2 RF plasma (both with and without tungsten or stainless steel surface) show the importance of the presence of a catalyst in the ammonia formation process. By increasing the temperature of the W samples up to 1270 K, ammonia formation demonstrated a continuous decrease due to two major factors. For high temperatures above 650 K and 830 K, for stainless steel and W, respectively, the reduction results from the thermal decomposition of ammonia. For the lower temperature range, the temperature rise leads to the formation of more stable nitrides that do not tend to react further with hydrogen to form NH2 and NH3. Interestingly, the addition of helium or argon to the N2/H2 plasma show opposite effects on the ammonia production. He effectively decreases the percentage of NH3 by acting as a barrier for the surface processes. On the other hand, argon impacts the plasma processes more, probably by increasing the active nitrogen species in the plasma and as a consequence the percentage of formed ammonia. [ABSTRACT FROM AUTHOR]

The CIMPLE-PSI laboratory in India’s northeast corner has been engaged over the last decade in the development of advanced experimental systems for controlled plasma fusion-relevant plasma surface interaction (PSI) studies and material testing. The CIMPLE-PSI experimental device was commissioned recently, where a magnetized collimated helium plasma beam is produced under steady-state conditions inside a linear vacuum chamber, which can be made to interact with remotely placed material targets, under controlled experimental conditions. This paper reports on the design, development of the major sub-systems of this device and the performance of the integrated system. The peak helium ion flux and heat flux were measured as 1024 m−2 s−1 and 5.1 MW m−2, respectively; this confirms that extreme ITER divertor-like parameters are successfully reproduced by this simulation device. Steady-state operation of the electromagnet allowed prolonged operation of the plasma leading to a very high fluence of 0.3  ×  1028 m−2, which may be enhanced further in the future. Tungsten samples were exposed in this device under helium plasma that had led to the formation of nanometer-sized tungsten fibre foam structures on the target. The technique of grazing incidence small angle x-ray scattering was successfully utilized to measure the average size of the helium bubbles remaining buried in the exposed sample and its variation with depth from the top of the surface. [ABSTRACT FROM AUTHOR]

Recrystallisation has been identified as a critical issue for the ITER divertor but currently the synergistic effects of heat and particle bombardment on tungsten recrystallisation are still unclear. In this study, tungsten was exposed to helium plasma under a range of sample temperatures (573 K–1073 K) and ion fluences (1024–1025 m−2) followed by annealing at a range of temperatures (1373 K–1673 K). It is quantitatively found that larger helium nano-bubbles form at higher sample temperatures during plasma exposure due to increased nano-bubble mobility and coalescence. The retardation of recrystallisation showed the opposite trend, with less recrystallisation occurring in samples exposed at lower temperatures and with smaller nano-bubbles. This result is attributed to the Zener drag force from the helium nano-bubbles opposing grain boundary movements during annealing with the effect of the force being inversely proportional to bubble size. Ion fluence is not found to affect the extent of recrystallisation at any temperature. The presence of the sub-surface helium nano-bubbles increased the microstrain in the tungsten which was eliminated during subsequent annealing as a clear sign of the recovery stage proceeding. [ABSTRACT FROM AUTHOR]

Ammonia formation was studied in a series of dedicated nitrogen seeded H-mode discharges at ASDEX Upgrade. The evolution of ammonia formation was investigated with a reference phase before the seeding, and a long, stable flat-top nitrogen-seeded phase. It was monitored with divertor spectroscopy and analysis of the exhaust gas. The amount of the detected ammonia increased continuously over the course of five discharges with the same nitrogen seeding rate. The same trend was observed in the nitrogen density in the core plasma, as measured by charge exchange recombination spectroscopy and other signals, linked to the effects of nitrogen seeding. The results show that the rate of ammonia formation exhibited the same trend as the nitrogen density in the plasma. This density, in turn, was strongly influenced by the nitrogen wall inventory. The spatial distribution of the detected ammonia suggests that a significant contribution to the net ammonia formation is made in plasma-shaded areas, through surface reactions of neutral species. [ABSTRACT FROM AUTHOR]

Gas seeding is often used in tokamaks to reduce the power load onto the divertor target plates. Nitrogen is the preferred seeding species because of its favourable radiative properties as well as its apparent beneficial effect on plasma confinement. However, nitrogen molecules are chemically reactive with hydrogen and its isotopes to form stable ammonia compounds. Since ammonia is a polar molecule, sticking on metal surfaces can be expected, increasing as a consequence the tritium retention which could pose a serious risk for ITER operation and maintenance. It is, therefore, important to understand the adsorption mechanism of ammonia on surfaces, investigate when the surface saturation occurs and whether ammonia adsorbs as a molecule or undergoes a dissociation on the surface. In this contribution, ammonia sticking on different fusion-relevant materials is presented. The results show a pressure-dependent ammonia sticking on tungsten, boron and stainless steel followed by a partial desorption from these surfaces while on gold and beryllium, ammonia molecules weakly adsorb and completely desorb. A detailed explanation of the two interaction mechanisms is addressed. Furthermore, the time dependence of ammonia desorption as well as the chemical state of non-desorbed residuals were investigated with x-ray photoelectron spectroscopy. Tungsten, boron and stainless steel surfaces showed a continuous dissociation process from NH3 to NH2, NH, N and surface nitrides. [ABSTRACT FROM AUTHOR]

The motion and temperature evolution of beryllium droplets produced by first wall surface melting after ITER major disruptions and vertical displacement events mitigated during the current quench are simulated by the MIGRAINe dust dynamics code. These simulations employ an updated physical model which addresses droplet-plasma interaction in ITER-relevant regimes characterized by magnetized electron collection and thin-sheath ion collection, as well as electron emission processes induced by electron and high-Z ion impacts. The disruption scenarios have been implemented from DINA simulations of the time-evolving plasma parameters, while the droplet injection points are set to the first-wall locations expected to receive the highest thermal quench heat flux according to field line tracing studies. The droplet size, speed and ejection angle are varied within the range of currently available experimental and theoretical constraints, and the final quantities of interest are obtained by weighting single-trajectory output with different size and speed distributions. Detailed estimates of droplet solidification into dust grains and their subsequent deposition in the vessel are obtained. For representative distributions of the droplet injection parameters, the results indicate that at most a few percents of the beryllium mass initially injected is converted into solid dust, while the remaining mass either vaporizes or forms liquid splashes on the wall. Simulated in-vessel spatial distributions are also provided for the surviving dust, with the aim of providing guidance for planned dust diagnostic, retrieval and clean-up systems on ITER. [ABSTRACT FROM AUTHOR]

The tungsten (W) material in the high heat flux regions of the ITER divertor will be exposed to high fluxes of low-energy particles (e.g. H, D, T, He, Ne and/or N). Combined with long-pulse operations, this implies fluences well in excess of the highest values reached in today’s tokamak experiments. Shaping of the individual monoblock top surface and tilting of the vertical targets for leading-edge protection lead to an increased surface heat flux, and thus increased surface temperature and a reduced margin to remain below the temperature at which recrystallization and grain growth begin. Significant morphology changes are known to occur on W after exposure to high fluences of low-energy particles, be it H or He. An analysis of the formation conditions of these morphology changes is made in relation to the conditions expected at the vertical targets during different phases of operations. It is concluded that both H and He-related effects can occur in ITER. In particular, the case of He-induced nanostructure (also known as ‘fuzz’) is reviewed. Fuzz formation appears possible over a limited region of the outer vertical target, the inner target being generally a net Be deposition area. A simple analysis of the fuzz growth rate including the effect of edge-localized modes (ELMs) and the reduced thermal conductivity of fuzz shows that the fuzz thickness is likely to be limited by the occurrence of annealing during ELM-induced thermal excursions. Not only the morphology, but the material mechanical and thermal properties can be modified by plasma exposure. A review of the existing literature is made, but the existing data are insufficient to conclude quantitatively on the importance and extent of these effects for ITER. As a consequence of the high surface temperatures in ITER, W recrystallization is an important effect to consider, since it leads to a decrease in material strength. An approach is proposed here to develop an operational budget for the W material, i.e. the time the divertor material can be operated at a given temperature before a significant fraction of the material is recrystallized. In general, while it is clear that significant surface damage can occur during ITER operations, the tolerable level of damage in terms of plasma operations currently remains unknown. [ABSTRACT FROM AUTHOR]

The effect of helium (He) plasma exposure, and associated surface modifications, on the thermal shock resistance of tungsten (W) under ITER relevant steady state and transient heat and particle loads was studied. W samples were exposed to steady state and pulsed He plasmas at surface base temperatures from 670 to 1170 K. The same exposures were repeated in hydrogen (H) to allow a direct comparison of the role of the ion species on the thermal shock resistance. Exposure to He plasma pulses caused the formation of fine cracking network on W samples which occurred at a higher density and smaller depths compared to H pulsed plasma irradiation. The peak temperature reached during an ELM-like plasma pulse increased by a factor ~1.45 over the 100 s of He plasma exposure, indicating a deterioration of the thermal properties. Transient loading experiments were also performed using a high power pulsed laser during He plasma exposure, showing a significant modification of the target thermal response caused by the surface damage. The effect of He-induced morphology changes on the thermal response modification was found to be very small compared to that of transient-induced damage. [ABSTRACT FROM AUTHOR]

For the final design of the ITER divertor it is important to determine whether shaping of each tungsten monoblock to eliminate leading edges is required or not. In order to aid this decision, two experiments were performed in DIFFER’s linear plasma devices to study heat loads on misaligned water cooled blocks at glancing incidence. First, a series of tungsten blocks were exposed to a high parallel heat flux (26 MW ) hydrogen plasma beam in the Magnum-PSI linear device. The blocks were exposed at an oblique angle between and with respect to the plasma beam, approaching the low angle of incidence expected in the ITER divertor strike-point regions. One block was vertically misaligned with respect to the others by a value between 0 and mm, and the surface temperature evolution monitored. This was compared to a finite element thermo-mechanical model of the system, to investigate any discrepancies between the two. The importance of Larmor orbit smoothing effects could also be considered, as the Larmor radius was larger than half the misalignment height. Next, a row of identical tungsten blocks were exposed at an angle of to  ∼500–700 μs pulsed hydrogen plasmas of peak power between 275–625 MW in the Pilot-PSI linear device, to simulate the effect of edge localized mode (ELM) transients. One block was misaligned with respect to the others by 1 mm. In all cases, the model and experiment were found to be in excellent agreement (to  ∼1%), demonstrating clearly that an optical approximation for plasma power loading on misaligned edges is appropriate. A simple model including Larmor smoothing was applied, the results of which give a similar prediction to the optical approximation for the temperature of the misaligned edges, within the error of the temperature measurements, though a small decrease in edge temperature is calculated. Therefore it can be concluded that the application of the optical approximation is generally appropriate as an input physics assumption for the study of shaping alternatives. [ABSTRACT FROM AUTHOR]

The fuel outgassing efficiency from plasma-facing components exposed in JET-ILW has been studied at ITER-relevant baking temperatures. Samples retrieved from the W divertor and Be main chamber were annealed at 350 and 240 °C, respectively. Annealing was performed with thermal desoprtion spectrometry (TDS) for 0, 5 and 15 h to study the deuterium removal effectiveness at the nominal baking temperatures. The remained fraction was determined by emptying the samples fully of deuterium by heating W and Be samples up to 1000 and 775 °C,respectively. Results showed the deposits in the divertor having an increasing effect to the remaining retention at temperatures above baking. Highest remaining fractions 54 and 87 were observed with deposit thicknesses of 10 and 40 μm, respectively. Substantially high fractions were obtained in the main chamber samples from the deposit-free erosion zone of the limiter midplane, in which the dominant fuel retention mechanism is via implantation: 15 h annealing resulted in retained deuterium higher than 90. TDS results from the divertor were simulated with TMAP7 calculations. The spectra were modelled with three deuterium activation energies resulting in good agreement with the experiments. [ABSTRACT FROM AUTHOR]

Plasma-wall interactions have been studied in the full-W ASDEX Upgrade during its dedicated helium campaign. Relatively clean plasmas with a He content of  >80% could be obtained by applying ion cyclotron wall conditioning (ICWC) discharges upon changeover from D to He. However, co-deposited layers with significant amounts of He and D were measured on W samples exposed to ICWC plasmas at the low-field side (outer) midplane. This is a sign of local migration and accumulation of materials and residual fuel in regions shadowed from direct plasma exposure albeit globally D was removed from the vessel. When exposing W samples to ELMy H-mode helium plasmas in the outer strike-point region, no net erosion was observed but the surfaces had been covered with co-deposited layers mainly consisting of W, B, C, and D and being the thickest on rough and modified surfaces. This is different from the typical erosion-deposition patterns in D plasmas, where usually sharp net-erosion peaks surrounded by prominent net-deposition maxima for W are observed close to the strike point. Moreover, no clear signs of W nanostructure growth or destruction could be seen. The growth of deposited layers may impact the operation of future fusion reactors and is attributed to strong sources in the main chamber that under suitable conditions may switch the balance from net erosion into net deposition, even close to the strike points. In addition, the absence of noticeable chemical erosion in helium plasmas may have affected the thickness of the deposited layers. Retention of He, for its part, remained small and uniform throughout the strike-point region although our results indicate that samples with smooth surfaces can contain an order of magnitude less He than their rough counterparts. [ABSTRACT FROM AUTHOR]

Surface morphology and deuterium retention in tungsten exposed at surface temperature of ~550 K to mixed deuterium–neon plasmas of different neon concentrations are investigated. It is found that the addition of neon up to 20% mitigates blistering on the surface. Cross-section view of the surface shows the formation of pores near the surface in the depth less than 100 nm. Deuterium depth profile is featured by an enhanced deuterium concentration within a depth of 16 nm but a mitigated penetration in depth larger than 1 µm. Deuterium retention is reduced by up to a factor of four. It is suggested the open pores formed in the surface serves as an escaping channel, mitigates deuterium penetration towards bulk and retention in the bulk. [ABSTRACT FROM AUTHOR]

Helium retention is measured in tungsten samples exposed to mixed H/He plasma in the Magnum-PSI linear plasma device. It is observed that there is very little He retention below helium ion impact energies of eV, indicating the existence of a potential barrier which must be overcome for implantation to occur. The helium retention in samples exposed to plasma at temperatures  >1000 K is strongly correlated with nano-bubble formation measured using grazing incidence small-angle x-ray scattering. The diameters of nano-bubbles were not found to increase with increasing helium concentration, indicating that additional helium must be accommodated by increasing the bubble concentration or an increase in bubble pressure. For some samples pre-irradiation with heavy ions of 2.0 MeV energy is investigated to simulate the effects of neutron damage. It is observed that nano-bubble sizes are comparable between samples pre-irradiated with heavy-ions, and those without heavy-ion pre-irradiation. [ABSTRACT FROM AUTHOR]

The goal of this work was to investigate coated first mirrors under very harsh erosion conditions. Mock-up mirrors were exposed to high-flux hydrogen/argon plasma in the linear plasma facility Magnum-PSI. Rhodium (Rh) and molybdenum (Mo) coated mirrors of different coating thicknesses, with or without water cooling, exhibited different responses to this exposure. Failures of Rh films were demonstrated for 5 micron thick film, 1 micron film revealed 10% decrease in the specular reflectivity only in the exposed area. In comparison, water cooled Mo mock-ups showed a significant diffuse reflectivity on the entire surface leading to more than 50% specular reflectivity losses in the visible range. The losses for non-cooled Mo samples did not exceed 7% in the whole studied wavelength range of 250–2500 nm. Three phenomena were proposed to explain these results. First the mechanical properties of the films as characterized by scratch and hardness measurements as well as residual stress analysis measured by x-ray diffraction. Rh films showed a high compressive stress value of 2.5  ±  0.4 GPa leading to poor adhesion of the thick films deposited on stainless steel substrate due to the high amount of available energy per area stored in the unbuckled film i.e. J m−2. It was confirmed by ANSYS simulation that the von Mises stress for the Rh coating was twice as high as that for the Mo coating due to different mechanical properties. Moreover, the maximum stress for thick Rh film (261 MPa) was higher than the critical buckling stress calculated with a buckle clamped Euler column model demonstrating the failure mode of the film. The second phenomenon was roughening of the mirror surface which was flux and temperature dependent, i.e. at low temperatures the surface would roughen randomly without any oriented surface morphology and at higher temperatures the surface diffusion constants would dominate the process and smoothen the surface. The last phenomenon was the significant oxidation and carbidization of the Mo surface even on the non-exposed area, as detected by x-ray photoelectron spectroscopy, leading also to a decrease in the reflectivity in the entire measured range, which was not observed for Rh film. [ABSTRACT FROM AUTHOR]

The first study of the interaction of metallic dust (tungsten, aluminum) adhered on tungsten substrates with transient plasma heat loads is presented. Experiments were carried out in the Pilot-PSI linear device with transient heat fluxes up to 550 MW m−2 and in the DIII-D divertor tokamak. The central role of the dust-substrate contact area in heat conduction is highlighted and confirmed by heat transfer simulations. The experiments provide evidence of the occurrence of wetting-induced coagulation, a novel growth mechanism where cluster melting accompanied by droplet wetting leads to the formation of larger grains. The physical processes behind this mechanism are elucidated. The remobilization activity of the newly formed dust and the survivability of tungsten dust on hot surfaces are documented and discussed in the light of implications for ITER. [ABSTRACT FROM AUTHOR]

Linear plasma generators are plasma devices designed to study fusion-relevant plasma-surface interactions. The first requirement for such devices is to operate with adjustable and well characterized plasma parameters. In the linear plasma device Magnum-PSI, the distribution of the charged particle flux across the target surface can be tailored by the target bias. The process is based on the radial inhomogeneity of the plasma column and it is evidenced by electrical measurements via a 2D multi-probe system installed as target. Typical results are reported for a hydrogen discharge operated at 125 A and confined by a magnetic field strength of 0.95 T in the middle of the coils. The probes were biased in the range of  −80 to  −25 V, while the floating potential of the target was about  −35 V. The results were obtained in steady-state regime of Magnum-PSI, being time-averaged over any type of fluctuations. Depending on the relative value of the target bias voltage with respect to the local floating potential in the plasma column, the entire target surface can be exposed to ion or electron dominated flux, respectively, or it can be divided into two adjacent zones: one exposed to electron flux and the other to ion flux. As a consequence of this effect, a floating conductive surface that interacts with an inhomogeneous plasma is exposed to non-zero local currents despite its overall null current and it is subjected to internal current flows. [ABSTRACT FROM AUTHOR]

The first combined experimental and theoretical studies of dust remobilization by plasma forces are reported. The main theoretical aspects of remobilization in fusion devices under steady state conditions are analyzed. In particular, the dominant role of adhesive forces is highlighted and generic remobilization conditions—direct lift-up, sliding, rolling—are formulated. A novel experimental technique is proposed, based on controlled adhesion of dust grains on tungsten samples combined with detailed mapping of the dust deposition profile prior and post plasma exposure. Proof-of-principle experiments in the TEXTOR tokamak and the EXTRAP-T2R reversed-field pinch are presented. The versatile environment of the linear device Pilot-PSI allowed for experiments with different magnetic field topologies and varying plasma conditions that were complemented with camera observations. [ABSTRACT FROM AUTHOR]

Lithium-coated high-Z substrates are planned for use in the NSTX-U divertor and are a candidate plasma facing component (PFC) for reactors, but it remains necessary to characterize the gross Li erosion rate under high plasma fluxes (>1023 m−2 s−1), typical for the divertor region. In this work, a realistic model for the compositional evolution of a Li/D layer is developed that incorporates first principles molecular dynamics (MD) simulations of D diffusion in liquid Li. Predictions of Li erosion from a mixed Li/D material are also developed that include formation of lithium deuteride (LiD). The erosion rate of Li from LiD is predicted to be significantly lower than from pure Li. This prediction is tested in the Magnum-PSI linear plasma device at ion fluxes of 1023–1024 m−2 s−1 and Li surface temperatures  ⩽800 °C. Li/LiD coatings ranging in thickness from 0.2 to 500 μm are studied. The dynamic D/Li concentrations are inferred via diffusion simulations. The pure Li erosion rate remains greater than Langmuir Law evaporation, as expected. For mixed-material Li/LiD surfaces, the erosion rates are reduced, in good agreement with modelling in almost all cases. These results imply that the temperature limit for a Li-coated PFC may be significantly higher than previously imagined. [ABSTRACT FROM AUTHOR]

The influence of recrystallization on thermal shock resistance has been identified as an issue that may influence the long term performance of ITER tungsten (W) divertor components. To investigate this issue a unique series of experiments has been performed on ITER divertor W monoblock mock-ups in three EU high heat flux facilities: GLADIS (neutral beam), JUDITH 2 (electron beam) and Magnum-PSI (plasma beam). To simulate ITER mitigated edge localised modes, heat fluxes between 0.11 and 0.6 GW m−2 were applied for Δt  <  1 ms. Two different base temperatures, Tbase  =  1200 °C and 1500 °C, were chosen on which ~18 000/100 000 transient events were superimposed representing several full ITER burning plasma discharges in terms of number of transients and particle fluence. An increase in roughening for both e-beam and plasma loaded surfaces was observed when loading during or after recrystallization and when loading at higher temperature. However, regarding the formation of cracks and microstructural modifications the response was different for e-beam and plasma loaded surfaces. The samples loaded in Magnum-PSI did not crack nor show any sign of recrystallization, even at Tbase  =  1500 °C. This could be a dynamic hydrogen flux effect, because pre-loading of samples with hydrogen neutrals (GLADIS) or without hydrogen (e-beam JUDITH 2) did not yield this result. These results show clearly that the loading method used when investigating and qualifying the thermal shock performance of materials for ITER and future fusion reactors can play an important role. This should be properly accounted for and in fact should be the subject of further R&D. [ABSTRACT FROM AUTHOR]

Experiments were performed in three different facilities in order to investigate the impact of combined steady state deuterium plasma exposure and ELM-like thermal shock events on the performance of ultra high purity tungsten. The electron beam facility JUDITH 1 was used to simulate pure thermal loads. In addition the linear plasma devices PSI-2 and Pilot-PSI have been used for successive as well as simultaneous exposure where the transient heat loads were applied by a high energy laser and the pulsed plasma operation, respectively. The results show that the damage behaviour strongly depends on the loading conditions and the sequence of the particle and heat flux exposure. This is due to hydrogen embrittlement and/or a higher defect concentration in the tungsten near surface region due to supersaturation of hydrogen. The different results in terms of damage formation from both linear plasma devices indicate that also the plasma parameters such as particle energy, flux and fluence, plasma impurities and the pulse shape have a strong influence on the damage performance. In addition, the different loading methods such as the scanning with the electron beam in contrast to the homogeneous exposure by the laser leads to an faster increase of the surface roughness due to plastic deformation. [ABSTRACT FROM AUTHOR]

The thermal shock behaviour of blister-covered W surfaces during combined steady-state/pulsed plasma loading was studied by scanning electron microscopy and electron backscatter diffraction. The W samples were first exposed to steady-state D plasma to induce blisters on the surface, and then the blistered surfaces were exposed to steady-state/pulsed plasma. Growth and cracking of blisters were observed after the exposure to the steady-state/pulsed plasma, while no obvious damage occurred on the surface area not covered with blisters. The results confirm that blisters induced by D plasma might represent weak spots on the W surface when exposed to transient heat load of ELMs. The cracks on blisters were different from the cracks due to the transient heat loads reported before, and they were assumed to be caused by stress and strain due to the gas expansion inside the blisters during the plasma pulses. Moreover, most of cracks were found to appear on the blisters formed on grains with surface orientation near [1 1 1]. [ABSTRACT FROM AUTHOR]

Dust trajectories have been recorded with an unprecedented, under fusion-relevant plasma conditions, spatial resolution of 9 μm/pixel in Pilot-PSI. The optical setup allowed the use of fast cameras as a basic microscope. It is demonstrated that such a resolution is essential for the correct interpretation of experiments on several aspects of dust-surface interactions. Highly resolved tungsten dust dynamics measurements are presented from dedicated experiments on dust collisions with plasma facing components, motion in the vicinity of castellated samples and remobilization from planar samples. [ABSTRACT FROM AUTHOR]