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Undeformed tungsten suffers from a brittleness that makes it unsuitable for applications at low temperatures. Cold-worked tungsten materials such as drawn wires or rolled plates can however show considerable ductility even at low temperatures. The reason for this behaviour is so far not understood. We investigated a series of potassium-doped tungsten wires that were subsequently drawn from one sintered ingot, making them chemically identical. Hence, the properties of the wires could be studied without the influence of different impurity levels. Using transient mechanical tests, namely repeated stress relaxation experiments and strain-rate jump tests, the effective activation volumes (Formula presented.) and strain-rate sensitivities m of the wires were determined at room-temperature. Based on the obtained results, it is deduced that the motion of (Formula presented.) screw dislocations by formation and dissociation of kink-pairs is controlling the rate of plastic deformation in all wires that show plasticity at room temperature. It is hence concluded that the ductility of drawn tungsten wires at low temperatures is not due to a change in the rate-controlling deformation mechanisms, but should be a consequence of the microstructural and textural changes during wire drawing.

Tungsten fiber-reinforced tungsten (Wf/W) composites are being developed to improve the intrinsic brittleness of tungsten. In these composites, engineered fiber/matrix interfaces are crucial in order to realize toughening mechanisms. For such a purpose, yttria (Y2O3), being one of the suitable interface materials, could be realized through different coating techniques. In this study, the deposition of thin films of yttria on a 150 µm tungsten wire by physical and chemical vapor deposition (PVD and CVD) techniques is comparatively investigated. Although fabrication of yttria is feasible through both CVD and PVD routes, certain coating conditions such as temperature, growth rate, oxidation of Wf, etc., decide the qualitative nature of a coating to a particular extent. In the case of PVD, the oxidation of Wf is highly reduced compared to the WO3 formation in high-temperature CVD coating processes. Yttria-coated tungsten fibers are examined comprehensively to characterize their microstructure, phase, and chemical composition using SEM, XRD, and Raman spectroscopy techniques, respectively.

Erosion of material by energetic ions, i.e., sputtering, is widely used in industry and research. Using experiments and simulations that, independently of each other, obtain the sputter yield of thousands of individual grains, we demonstrate here that the sputter yield for heavy keV ions on metals changes as a continuous function of the crystal direction. Moreover, we show that polycrystalline metals with randomly oriented grains do not sputter with the same yield as the amorphous material. The key reason for this is attributed to linear collision sequences rather than channeling.

Self-passivating tungsten based alloys for the first wall armor of future fusion reactors are expected to provide an important safety advantage compare to pure tungsten in case of a loss-of-coolant accident with simultaneous air ingress, due to the formation of a stable protective scale at high temperatures in presence of oxygen preventing the formation of volatile and radioactive WO3. In this work, Zr is added to self-passivating W-10Cr-0.5Y alloy, manufactured by mechanical alloying and HIP, in view of improving its mechanical strength and thus, its thermal shock resistance. The as-HIPed W-10Cr-0.5Y-0.5Zr exhibits a nanocrystalline microstructure with the presence of an extremely fine nanoparticle dispersion. After heat treatment at 1555 °C for 1.5 h, the grain size growths from less than 100 nm to 620 nm and nanoparticles are present both at the grain boundaries and inside the grains. Oxidation tests at 1000 °C revealed that the alloy with Zr exhibits also a strong oxidation reduction compared to pure W. The long-term oxidation rate is similar to that of the alloy without Zr. Under thermal shock loading simulating 1000 ELM-like pulses at the divertor, the heat treated Zr-containing alloy did not present any damage.

To overcome the intrinsic brittleness of tungsten (W), a tungsten fiber-reinforced tungsten-composite material (Wf/W) is a possible solution. The introduction of energy dissipation mechanisms like fiber bridging or fiber pull-out by means of an engineered interface between fiber and matrix mitigate the brittleness of tungsten and lead to a pseudo-ductile material behaviour. The push-out test of single-fiber samples is an experimental method to investigate the properties of the interface between fiber and matrix of composite materials. It is widely used for the investigation of ceramic composites. This method was also used to investigate the debonding and frictional properties of the Er2O3 interface region between fiber and matrix of Wf/W single-fiber samples made by CVDand HIP-processes. In this article finite element calculations are used to get a better understanding of the processes acting in the interface during a push-out test of Wf/W. A detailed overview of the debonding progress and of the corresponding stress states of the interface during the different stages of the test is presented. In addition the sensitivity of the push-out behaviour regarding the different interface properties and the plastic flow curve of the tungsten fiber are investigated.

In order to get first insight into net tungsten erosion in W7-X, tungsten (W) marker layers were exposed during the operational phase OP 1.2b at one position of the Test Divertor Unit (TDU), at 21 different positions of the inner heat shield, and at two scraper elements. The maximum tungsten erosion rate at the TDU strike line was 0.13 nm s−1 averaged over the whole campaign. The erosion was inhomogeneous on a microscopic scale, with higher erosion on ridges of the rough surface inclined towards the plasma and deposition of hydrocarbon layers in the recessed areas of the rough surface. The W erosion at the inner heat shield was below the detection limit of 3–6 × 1012 W-atoms/cm2s, and all inner heat shield tiles were covered with a thin B/C/O layer with thickness in the range 2 × 1017–1018 B + C atoms/cm2 (about 20–100 nm B + C). W-erosion of the marker layers on the scraper elements was also below the detection limit.

Plastic deformation of tungsten wire is an effective source of toughening tungsten fibre-reinforced tungsten composites (Wf/W) and other tungsten fibre-reinforced composites. To provide a reference for optimization of those composites, unconstrained pure tungsten wire is studied after various heat treatments in terms of microstructure, mechanical behaviour and fracture mode. Recrystallization is already observed at a relatively low temperature of 1273 K due to the large driving force caused by a high dislocation density. Annealing for 30 min at 1900 K also leads to recrystallization, but causes a rather different microstructure. As-fabricated wire and wire recrystallized at 1273 K for 3 h show fine grains with a high aspect ratio and a substantial plastic deformability: a clearly defined tensile strength, high plastic work, similar necking shape, and the characteristic knife-edge-necking of individual grains on the fracture surface. While the wire recrystallized at 1900 K displays large, almost equiaxed grains with low aspect ratios as well as distinct brittle properties. Therefore, it is suggested that a high aspect ratio of the grains is important for the ductile behaviour of tungsten wire and that embrittlement is caused by the loss of the preferable elongated grain structure rather than by recrystallization. In addition, a detailed evaluation of the plastic deformation behaviour during tensile test gives guidance to the design and optimization of tungsten fibre-reinforced composites.

We performed a systematic study of the deuterium retention in the tungsten heavy alloy W-2Ni-1Fe (HPM 1850 from HC Starck, Germany). Consisting of two different phases with generally two distinctly different heats of solution, such alloys represent a special class of materials with respect to hydrogen isotope transport and retention. Deuterium was implanted from a plasma with an energy of 38 eV D−1 at sample temperatures of 100 °C, 175 °C, and 250 °C in the fluence range 1023 D m−2 to 4 × 1025 D m−2. For each combination of implantation temperature and fluence, two alloy samples were exposed together with one sample of bulk hot-rolled tungsten (from Plansee SE). The amount of retained deuterium was determined by thermal desorption with heating up to 850 °C for the alloy and 1050 °C for the bulk tungsten. The outgassed deuterium was detected by mass spectroscopy at the mass channels 3, 4, 19, and 20 corresponding to HD, D2, HDO, and D2O, respectively. The near-surface deuterium concentration was determined by 3He nuclear reaction analysis. The retention in the alloy as compared to the bulk tungsten decreases systematically with increasing implantation fluence for each of the investigated implantation temperatures. For the higher fluences the ratio decreases systematically also, as a function of implantation temperature. This means that especially for high implantation fluences and temperatures the alloy HPM 1850 systematically retains less hydrogen isotopes than bulk tungsten. This property represents an important qualification criterion as a plasma-facing material for a future fusion reactor.

Advanced engineering materials (2020). doi:10.1002/adem.201901242, Published by Wiley-VCH Verl., Weinheim

Material issues pose a significant challenge for the design of future fusion reactors. Recently progress has been made towards fully dense multi short-fibre powder metallurgical production of tungsten-fibre reinforced tungsten (Wf /W) as well as optimizing the process understanding for the routes using chemical vapour deposition (CVD). For CVD-Wf /W weaves and textile preforms are being used to facilitate large scale production. Classically 150 μm tungsten fibres supplied by OSRAM GmbH have been used. In order to facilitate the better use of textile processes less stiff 16 μm filaments are being evaluated. The strength of the 16 μm filament is at 4500 MPa and thus significantly higher than the strength of the 150 μm fibre (~2500 MPa) (in the as-fabricated state). Better weavability allows a more flexible use of fibre preforms Two main yarn production routes have been investigated: covered yarns where a set of tungsten filaments is held together by a PVA (Polyvinyl alcohol) cover and braided yarns. In order to allow a comparison to the previously used single fibres, yarns with ~140 μm effective diameter were produced. Braided yarns with tensile strength of 2500 MPa and 6% strain at fracture and twisted yarns with tensile strength of 4500 MPa and 3% strain at fracture. For both yarns single fibre CVD samples have been produced to investigate the infiltration properties of the yarns and thus their applicability for the CVD route. A dense infiltration is observed for all yarns under investigation

This report summarizes the contributions presented at the 3rd IAEA technical meeting on divertor concepts, held in Vienna, Austria, 4–7 November 2019. The meeting brought together more than 70 experts from nuclear fusion research sites worldwide to discuss the different aspects that the divertor design and fusion machine operation involve, from ITER divertor developments to innovative technologies for future DEMO divertor. The main topics of the meeting were: divertor and confinement; radiative power exhaust; scrape-off layer (SOL) and divertor physics; steady state operation and transient heat loads; plasma facing components materials and heat exhaust for steady state operation; and divertors for DEMO and future power reactors.

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

In the stellarator Wendelstein 7-X with its twisted 3D magnetic field geometry, studies of material migration with respect to first wall components become very important in view of the envisioned long-pulse operation. A variety of erosion/deposition samples were installed on the plasma-facing components exposed at three different nominal heat load levels between 0.1 and 10 MW/m2. After the first successful operation phase in divertor configuration, all the probes at higher and lower load levels were removed, whereas at the intermediate load levels, 352 out of 30 000 screws have been exchanged at selected locations along the toroidal and poloidal directions. The exchanged probes have been analyzed by various measurement techniques. At the higher load levels where the probes were installed within the divertor, heavy erosion has been observed presumably at the strike line positions. Both, erosion and deposition phenomena have been found on the screw heads. The optical reflection measurement profile of the whole plasma vessel show the deposition patterns at similar locations in all the five modules. At the low load level, the Si-wafer probes are under investigation.

JT-60SAではタングステン壁で定常・高ベータプラズマを生成するため、炭素壁表面にタングステンを被覆する計画がある。本研究では、炭素繊維複合材料に被覆した厚さ0.4 mmのタングステンに対して熱負荷試験を行った。15 MW/m2で4秒間の熱負荷を50サイクル印加すると、被覆タングステンは1800℃まで温度上昇したが、剥離などの重大な損傷は形成されなかった。有限要素法で熱負荷試験時の被覆タングステンの温度分布を解析し、被覆界面の熱抵抗は無視できるほど小さいことがわかった。本被覆タングステンをJT-60SAのダイバータで使用する場合、被覆タングステンの再結晶化を防ぐためにはダイバータ形状の最適化が必要であることがわかった。, Plasma Conference 2017

JT-60SAではタングステン壁で定常・高ベータプラズマを生成するため、炭素壁表面にタングステンを被覆する計画がある。本研究では、炭素繊維複合材料に被覆した厚さ0.4 mmのタングステンに対して熱負荷試験を行った。15 MW/m2で4秒間の熱負荷を50サイクル印加すると、被覆タングステンは1800℃まで温度上昇したが、剥離などの重大な損傷は形成されなかった。, 核融合原型炉開発に向けたアクションプラン説明会

JT-60SAではタングステン壁で高密度・高圧力プラズマを生成するため、炭素壁表面にタングステンを被覆する計画がある。ダイバータターゲットに被覆する場合、15 MW/m2で定常の繰り返し熱負荷に耐える必要がある。本研究では、真空プラズマスプレー法で厚さ0.4 mmのタングステンを被覆した炭素繊維複合材料に15 MW/m2, 3秒間及び4秒間の熱負荷を50サイクルずつ印加した。熱サイクル印加後タングステン被膜を電子顕微鏡で観察し、剥離、溶融、割れがないことを確認し、本タングステン被覆材料が15 MW/m2の熱負荷に耐えられる可能性を示した。, 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications

Tungsten has established itself as the most suitable plasma-facing material for long-term operation in future magnetic-confinement fusion devices, but its properties make it a poor structural material and complicate the manufacturing of complex components. Recent advances in additive-manufacturing (AM) technology have begun to make the production of tungsten components with complex geometry more feasible. The design freedom afforded by AM could be leveraged to produce more resilient plasma-facing components (PFCs). A methodology to optimize the material distribution of composite PFCs was developed to reduce the maximum thermal stress caused by high heat fluxes. Its use was demonstrated for copper-infiltrated AM tungsten (Walt;subagt;AMalt;/subagt;/Cu) structures. Stress reductions of 50 - 85% are predicted under nominal load conditions. Optimized designs also reduce stress over a wide range of off-nominal conditions. The resulting optimized structures are composed of a spatially heterogeneous distribution of W and Cu comprising a broad range of composite mixtures. A sample manufacturable component was modelled based on optimization results.

The impact of lithium (Li) on plasma performance was investigated at the ASDEX Upgrade tokamak, which features a full tungsten wall. Li pellets containing 1.6 × 1020 Li atoms were launched with a speed of 600 m s−1 to achieve deep penetration into the plasma and minimize the impact on the first wall. Homogeneous transient Li concentrations in the plasma of up to 15% were established. The Li sustainment time in the plasma decreased with an increasing heating power from 150 to 40 ms. Due to the pellet rate being restricted to 2 Hz, no Li pile-up could take place. No significant positive impact on plasma properties, as reported from other tokamak devices, could be found; the Li pellets rather caused a small reduction in plasma energy, mainly due to enhanced radiation. Due to pellet injection, a short-lived Li layer was formed on the plasma-facing components, which lasted a few discharges and led to moderately beneficial effects during plasma start-up. Most pellets were found to trigger type-I ELMs, either by their direct local perturbation or indirectly by the altered edge conditions; however, reliability was less than 100%.

In the 2007 and early part of 2008 experimental campaigns, ASDEX Upgrade operated with full tungsten (W) wall without boronization. Use of ICRF power results in a significant increase of W source. Low temperature conditions at the plasma facing components, achieved by a large clearance between the separatrix and the antenna (>6 cm) and by elevated gas puff rates ( > 5 × 10 21 s - 1 ) help to lower W sputtering yield during ICRF. Operation of neighboring ICRF antennas at the phase difference close to −90° can lead to a reduction in the W source. However, a reduction of parallel near-fields by antenna design is needed to further minimize the W source. A relation has been established between the HFSS code calculations predicting a dominant role of box currents in the formation of parallel antenna near-fields and the experiment. The shapes of the measured vertical profile of effective sputtering yields and the calculated sheath driving voltages show a qualitative agreement. This confirms that the existing tools are a good basis to design an improved antenna.

Recent experiments in ASDEX Upgrade aimed at improving the physics base for ITER and DEMO to prepare operation and aid the design. In order to increase its exhaust capabilities and operational flexibility, a new bulk W divertor as well as an adjustable cryopump had been installed prior to the 2014 campaign. In experiments with high-field-side pellet injection, central electron densities twice as high as the Greenwald density limit could be achieved without strongly increasing the pedestal density and deleterious effect on confinement. Due to its large installed heating power, a large normalized heat flux $P_{\mathrm{ sep}}/R = 10$ MWm $^{-1}$ has been reached, representing two-thirds of the ITER value, under partially detached conditions with a peak target heat flux well below 10 MWm $^{-2}$ . The divertor load could be further reduced by increasing the core radiation, still keeping the confinement in the range of $H_{98y2}\approx 1$ . Suppression of edge-localized modes (ELMs) at low collisionality has been observed in a narrow spectral window in contrast to earlier results at high densities. The ITER $Q = 10$ baseline scenario has been investigated, matching as close as possible the triangularity, the plasma beta, $q_{95}$ , and the distance to the L-H threshold. It turned out that the ELM frequency is low and consequently the energy ejected by a single ELM is very high and ELM mitigation appears to be difficult. As a possible alternative, a scenario has been developed achieving a similar performance at a lower plasma current (and consequently higher $q_{95})$ . Experiments using electron cyclotron current drive (ECCD) with feedback-controlled deposition have allowed successfully testing several control strategies for ITER, including automated control of (3, 2) and (2, 1) neoclassical tearing modes during a single discharge. Concerning advanced scenarios, experiments with central ctr-ECCD have been performed in order to modify the q-profile. A strong reversal of the q-profile could be stationarily achieved and an internal transport barrier could be triggered. In disruption mitigation studies with massive gas injection (MGI), a runaway electron beam could be provoked and mitigated by a second MGI. Ongoing enhancements aim at strengthening the power supplies in order to allow full use of the installed heating power, the exchange of two ion cyclotron resonance heating (ICRH) antennas to reduce the W influx during ICRH, and the upgrading of the electron cyclotron resonance heating (ECRH) system to 7–8 MW for 10 s.

Experiments have been performed at ASDEX Upgrade, aiming to investigate the impact of lithium in an all-metal-wall tokamak and attempting to enhance the pedestal operational space. For this purpose, a lithium pellet injector has been developed, capable of injecting pellets carrying a particle content ranging from 1.82 × 10(19) atoms (0.21 mg) to 1.64 × 10(20) atoms (1.89 mg). The maximum repetition rate is about 2 Hz. Free flight launch from the torus outboard side without a guiding tube was realized. In such a configuration, angular dispersion and speed scatter are low, and a transfer efficiency exceeding 90% was achieved in the test bed. Pellets are accelerated in a gas gun; hence special care was taken to avoid deleterious effects by the propellant gas pulse. Therefore, the main plasma gas species was applied as propellant gas, leading to speeds ranging from 420 m/s to 700 m/s. In order to minimize the residual amount of gas to be introduced into the plasma vessel, a large expansion volume equipped with a cryopump was added into the flight path. In view of the experiments, an optimal propellant gas pressure of 50 bars was chosen for operation, since at this pressure maximum efficiency and low propellant gas flux coincide. This led to pellet speeds of 585 m/s ± 32 m/s. Lithium injection has been achieved at ASDEX Upgrade, showing deep pellet penetration into the plasma, though pedestal broadening has not been observed yet.

Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (Wf/W) has been developed incorporating extrinsic toughening mechanisms. Small Wf/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO3 compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

Experiments in the midsized tokamak ASDEX Upgrade (AUG) aim at improving the physics base for ITER and DEMO to prepare operation and aid the design. Being equipped with all tungsten plasma facing components ASDEX Upgrade could acquire important results concerning the fuel retention, tungsten sources and transport important for the operation of all metal devices. The availability of powerful heating systems as well as a comprehensive set of diagnostics allowed pushing the physics understanding in several areas such as pedestal and H-Mode physics, power exhaust and magneto hydro-dynamics (MHD) and development of integrated operational scenarios. Over the next 2 years the capabilities of the heating and current drive systems will be further improved, finally making available about 34 MW of auxiliary heating power. Additionally, existing diagnostic systems will be upgraded and new ones will be installed in order to provide even better insight into the physics of advanced and high power exhaust scenarios.

The observation in JET of co-deposition of tritium with carbon has led to a broad discussion on the replacement of graphite by a high-Z material for the first wall coverage. Moreover, due to the high erosion rate, carbon plasma facing components (PFCs) appear to be unacceptable for a commercial fusion reactor. Research in this area has subsequently gained increased attention. This paper describes the status of investigations on the use of tungsten as a first wall material. It discusses on the physical side the plasma wall interaction, the transport of tungsten in the plasma boundary and in the core. As an intermediate step on the technological side, graphite is often coated with tungsten layers. For highly loaded surfaces in a fusion reactor finally bulk tungsten components will have to be developed.

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10?20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITER's use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Recent experiments in ASDEX Upgrade aimed at improving the physics base for ITER and DEMO to aid the design and prepare operation. In order to increase its exhaust capabilities and operational flexibility a new bulk W divertor as well as an adjustable cryo-pump had been installed prior to the 2014 campaign. In experiments with high field side pellet injection central electron densities twice as high as the Greenwald density limit could be achieved without strongly increasing the pedestal density and deleterious effect on confinement. Due to its large installed heating power a large normalized heat flux Psep/R=10 MWm−1 has been reached, representing 2/3 of the ITER value, under partially detached conditions with a peak target heat flux well below 10 MWm−2. The divertor load could be further reduced by increasing the core radiation still keeping the confinement in the range of H98y≈1. ELM suppression at low collisionality has been observed in a narrow spectral window in contrast to earlier results at high densities. The ITER Q=10 baseline scenario has been investigated, matching as close as possible the triangularity, the plasma beta, q95 and the distance the L-H threshold. It turned out that the ELMs frequency is low and consequently the energy ejected by a single ELM is very high and ELM mitigation appears to be difficult. As a possible alternative an improved H-Mode scenario has been developed achieving a similar performance at lower plasma current (and consequently higher q95). Experiments using ECCD with feedback controlled deposition have allowed successfully testing several control strategies for ITER, including automated control of (3,2) and (2,1) NTMs during a single discharge. Concerning advanced scenarios, experiments with central ctr-ECCD have been performed in order to modify the q-profile. A strong reversal of the q-profile could be stationary achieved and an internal transport barrier could be triggered. In disruption mitigation studies with massive gas injection (MGI) a runaway electron beam could be provoked and mitigated by a second MGI. Ongoing enhancements aim at the strengthening of the power supplies in order to allow full use of the installed heating power, the exchange of two ICRH antennas to reduce the W influx during ICRH and the upgrading of the ECRH system to 7–8 MW for 10s.

Experiments dealing with plasma-wall interactions and first-wall materials comprise a significant part of the work program of ASDEX Upgrade. To elucidate carbon chemical erosion under reactor-relevant conditions, dedicated spectroscopic measurements were performed. These investigations are complemented with long-term erosion and deposition probes consisting of various materials, which are mounted at numerous locations inside the vacuum vessel. The codeposition of hydrogen with carbon below the divertor is studied in detail with long-term samples as well as with quartz microbalance measurements, which allow a discharge-resolved measurement of the layer growth. In parallel to the investigations on carbon, the behavior of tungsten plasma facing components (PFCs) and their influence on plasma performance is studied. In several experimental campaigns, the divertor as well as large parts of the PFCs in the main chamber were equipped with tungsten-coated tiles. Surface conditioning by applying a silicon layer (siliconization) was performed as a preexperiment of the tungsten program, and the results are compared to those of surface conditioning with boron (boronization)

Some of the remaining crucial plasma edge physics and plasma–material interaction issues of the ITER tokamak are discussed in this paper, using either modelling or projections of experimental results from existing tokamak operation or relevant laboratory simulations. The paper covers the following subject areas at issue in the design of the ITER device: (1) plasma thermal loads during Type I ELMs and disruptions, ensuing erosion effects and prospects for mitigating measures, (2) control of co-deposited tritium inventory when carbon is used even on small areas in the divertor near the strike points, (3) efficiency of edge and core fuelling for expected pedestal densities in ITER, and (4) erosion and impurity transport with a full tungsten divertor. Directions and priorities of future research are proposed to narrow remaining uncertainties in the above areas.

High heat flux tests have been performed on 0.4 mm thick W coatings produced by vacuum plasma spraying on carbon-fiber composites (CFCs) for the future actively cooled W-coated CFC divertor target in JT-60SA. The surface temperature of the W coating increased up to 2000 °C at 15 MW m−2 for 5 s. No severe morphological changes, for example, melting, cracking and delamination were observed although recrystallization and grain growth were found in the W coating. Cyclic loading at 15 MW m−2 for 3 s with 50 cycles and additional 50 cycles with longer pulse length of 4 s produced again no severe morphological changes. Heat transfer analysis of the W-coated CFC sample during the high heat flux test showed that thermal resistance between W coatings and CFCs was negligibly low. This analysis also showed that the surface temperature of the 1 mm thick W coating on the actively cooled CFC monoblock divertor target in JT-60SA will be higher than the recrystallization temperature, 1200 °C, at the steady state heat flux of 15 MW m−2. Therefore, a reduction of armor thickness of the future actively cooled W-coated CFC monoblock divertor target in JT-60SA could be necessary to decrease the surface temperature to less than the recrystallization temperature.

The former all-carbon wall on JET has been replaced with beryllium in the main torus and tungsten in the divertor to mimic the surface materials envisaged for ITER. Comparisons are presented between Type I H-mode characteristics in each design by examining respective scans over deuterium fuelling and impurity seeding, required to ameliorate exhaust loads both in JET at full capability and in ITER., 55 pages, 15 figures

Tungsten divertor plates have been used in ASDEX Upgrade for a full experimental campaign of approximately 800 discharges. Key issues investigated were the plasma performance, tungsten erosion, redeposition and migration, hydrogen isotope retention and the consequences of the simultaneous use of carbon and tungsten as plasma facing materials. Tungsten behavior was investigated by spectroscopic observation of line and quasi-continuum radiation and by surface analysis of material probes and of samples taken from plasma facing components after the experimental campaign. Code simulations were used to gain further understanding of the observed tungsten erosion properties and of tungsten plasma transport processes. The results obtained so far are presented and the implications with respect to the construction of future fusion devices are discussed.

The major aspects linked to the use of the JET auxiliary heating systems: NBI, ICRF and LHCD, in the new JET ITER-like wall (JET-ILW) are presented. We show that although there were issues related to the operation of each system, efficient and safe plasma heating was obtained with room for higher power. For the NBI up to 25.7MW was safely injected; issues that had to be tackled were mainly the beam shine-through and beam re-ionisation before its entrance into the plasma. For the ICRF system, 5MW were coupled in L-mode and 4MW in H-mode; the main areas of concern were RF-sheaths related heat loads and impurities production. For the LH, 2.5 MW were delivered without problems; arcing and generation of fast electron beams in front of the launcher that can lead to high heat loads were the keys issues. For each system, an overview will be given of: the main modifications implemented for safe use, their compatibility with the new metallic wall, the differences in behavior compared with the previous carbon wall, with emphasis on heat loads and impurity content in the plasma., 21 pages, 17 figures

Several experiments were conducted in ASDEX Upgrade to prove the suitability of tungsten as a divertor target material under the conditions of a high density and low temperature divertor. The observed fluxes from a tungsten tile into the plasma are low, in keeping with the extremely low sputtering yields. In addition, the very favourable effect of `prompt redeposition' (redeposition during the first gyration) could be confirmed by the experiments. Cooling of the edge region by neon injection seems permissible, i.e. neon impurity sputtering did not increase the eroded fluxes of tungsten. The transport and accumulation behaviour were investigated by means of the laser blow-off technique. No accumulation effects could be observed in ohmic discharges. In discharges with NBI heating but without ICRH, strong accumulation can occur. High heat flux tests were performed on graphite tiles coated with plasma sprayed tungsten, which withstood a thermal load of 15 MW/m2 lasting 2 s as well as 1000 cycles of 10 MW/m2 for 2 s without disabling damage. Owing to the encouraging results, an experiment using a tungsten divertor is planned in ASDEX Upgrade

JET underwent a transformation from a full carbon-dominated tokamak to a fully metallic device with beryllium in the main chamber and a tungsten divertor. This material combination is foreseen for the activated phase of ITER. The ITER-Like Wall (ILW) experiment at JET shall demonstrate the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments (Ip = 2.0 MA, Bt = 2.0-2.4 T, δ = 0.2-0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction in the long-term retention rate by more than a factor of 10 with respect to carbon-wall reference discharges. All experiments are executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system is reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges (Paux = 0.5 MW), type III (Paux = 5.0 MW) and type-I ELMy H-mode plasmas (Paux = 12.0 MW) a drop of the deuterium retention rate normalized to the operational time in divertor configuration is measured from 1.27 × 1021, 1.37 × 1021 and 1.97 × 1021 D s-1 down to 4.8 × 1019, 7.2 × 1019 and 16 × 1019 D s-1, respectively. The dynamic retention increases in the limiter phase in comparison with carbon-fibre composite, but also the outgassing after the discharge has risen in the same manner and overcompensates this transient retention. Overall an upper limit of the long-term retention rate of 1.5 × 1020 D s-1 is obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and is in line with identification of fuel co-deposition with Be as the main mechanism for the residual long-term retention. The reduction widens the operational space without active cleaning in the DT phase in comparison with a full carbon device. © 2013 IAEA, Vienna.

This paper reports the successful installation of the JET ITER-like wall and the realization of its technical objectives. It also presents an overview of the planned experimental programme which has been optimized to exploit the new wall and other JET enhancements in 2011/12.

The use of refractory metal plasma-facing components (PFCs) requires intensive research in all areas, i.e., in plasma-wall interaction, in the physics of the confined plasma, diagnostic, and in material development. Only a few present-day divertor tokamaks-mainly Alcator C-Mod (C-Mod) and ASDEX Upgrade (AUG)-gained experience with the refractory metals molybdenum and tungsten, respectively. AUG was stepwise converted from graphite to tungsten PFCs. In line with this transition, a reduction of the deuterium retention by almost a factor of ten has been observed due to the strong suppression of D codeposition with carbon. The deuterium retained in W is in line with laboratory results in contrast to C-Mod, where the D retention in Mo is more than a factor of ten larger than that in corresponding laboratory experiments. As expected from the sputtering threshold of Mo and W, negligible erosion by the thermal divertor background plasma is found in these experiments under low-temperature divertor conditions. However, erosion by fast particles and intrinsic impurities, which additionally might be accelerated in rectified electrical fields observed during ion cyclotron frequency heating, plays an important role. The Mo and W concentrations in the plasma center are strongly affected by plasma transport, and variations up to a factor of 50 are observed for similar influxes. However, it could be demonstrated that sawteeth and turbulent transport driven by central heating can suppress central accumulation. The inward transport of high-Z ions at the edge can be efficiently reduced by ?flushing? the pedestal region caused by frequent edge instabilities. Extrapolations to ITER and DEMO are difficult since the physics of the plasma transport is not yet completely understood, the particle and energy fluxes are orders of magnitude higher, and the technical boundary conditions in DEMO strongly differ from those of present-day devices.