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AbstractTungsten (W) specimens previously exposed to deuterium (D) plasmas both in the TEXTOR tokamak and high flux linear plasma generator (LPG) were subsequently loaded with tritium at 573 K for 3 h. Retention of tritium in the near-surface W layer was examined by imaging plate technique. On the TEXTOR-plasma-exposed W surface, tritium was mainly trapped in carbon deposits. For LPG-plasma-exposed W specimens, tritium was trapped in defects created in the near-surface layer during the course of D plasma exposure.

This contribution gives an overview of different simulation methods for transient events and damages induced in tungsten. The investigations were focussed on the resulting crack networks and special attention was paid to crack distance, width and depth. The results indicate that the different techniques show, in general, similar damage behaviours and the same damage thresholds.

AbstractThermo-oxidation is controlled exposure in an oxygen-containing atmosphere at elevated temperature and is being considered as a technique for the de-tritiation of carbon-based codeposits in ITER. In addition, unplanned oxidation may also occur during accidental air ingress. The impact of thermo-oxidation on ITER diagnostic mirrors causes concerns. A dedicated study was performed in DIII-D, where molybdenum and copper mirrors were installed in the main chamber, in the divertor, and at a location remote from the plasma and exposed for ~2 hours to a mixture containing 80% helium and 20% oxygen at a total pressure of 1.27 kPa. Mirrors in the main chamber and in the divertor were exposed at 350°C to 360°C whereas the temperature of mirrors in the remote area was ~160°C.Reflectivity of all mirrors was degraded after thermo-oxidation showing a decrease in the UV range from 60% to 10% for molybdenum mirrors and a 90% drop for copper mirrors at the wavelength 250 nm. The reflectivity of mirrors exposed at lower temperature was less degraded. Surface analyses revealed formation of oxides on all mirrors.In ITER, shutters planned for mirror protection are ineffective against thermo-oxidation. Nevertheless, in-situ cleaning systems planned for ITER mirrors may efficiently remove oxide layers.

AbstractThe chosen materials for plasma facing components for the deuterium/tritium phase of ITER are beryllium and tungsten. These materials have already been widely investigated in various devices like ion beam or electron beam tests. However, the operation of this material combination in a large tokamak including plasma wall interaction, material degradation, erosion and material mixing has not been proven yet.The ITER-like Wall, which has been recently installed in JET, consists of a combination of bulk tungsten and tungsten coated CFC divertor tiles as well as bulk beryllium and beryllium coated INCONEL in the main chamber. The experiments in JET will provide the first fully representative test of the ITER material choice under relevant conditions.This paper concentrates on material research and developments for the materials of the JET ITER-like Wall with respect to mechanical and thermal properties. The impact of these materials and components on the JET operating limits with the ITER-like Wall and implications for the ongoing scientific program will be summarised.

AbstractBehavior and characteristics of tungsten materials under impinging high heat fluxes are investigated. Experiments with inertially - not actively - cooled samples have been carried out in the plasma edge of the TEXTOR tokamak to study the changes of material properties such as grain size and abundance of voids or bubbles. In addition, the effects of electron beam impact regarding subsequent W power handling have been studied in view of future devices.The parallel heat flux at the radial position in TEXTOR impinging on the plasma-facing components (PFCs) ranges around q∥˜ 45 MW/m2allowing samples to be exposed at an impact angle of 35 deg to 20 to 30 MW/m2. Melt layer motion perpendicular to the magnetic field is observed following a Lorentz force originating from thermoelectric emission of the hot W sample. Up to 3 g of molten W are redistributed forming hill-like structures at the plasma-connected edge of the sample. The typical melt layer thickness is 1.0 to 1.5 mm. Those hills are, due to the changes in the local geometry, particularly susceptible to even higher heat fluxes of up to the full q∥; hence, locally the temperature of W can reach up to 6000 K, and thus boiling can occur.In terms of material degradation, several aspects are considered: formation of leading edges by redistributed melt, bubble formation, and recrystallization. Bubbles are occurring in sizes between 1 and 200 μm while recrystallization increases the grain size up to 1.5 mm. The power-handling capabilities are severely degraded by all those aspects. Melting of tungsten in future devices is highly unfavorable and needs to be avoided especially in light of uncontrolled transients and possible unshaped PFCs.Predamaged samples from the TEXTOR exposures have also been exposed in the JUDITH 1 facility under transient heat loads (up to ˜1 GW/m2, energy impact: 36 MW · m−2· s1/2). The samples show an unfavorable increase in the ductile-to-brittle transition temperature. In addition, surface cracks lose their directionality recrystallizing toward a more isotropic state from the manufactured monodirectional state. The increased grain size leads to a more brittle behavior under transient thermal loads with respect to crack progression.

Mechanisms underlying the retention of fuel species in tokamaks with carbon plasma-facing components are presented, together with estimates for the corresponding retention of tritium in ITER. The consequential requirement for new and improved schemes to reduce the tritium inventory is highlighted and the results of ongoing studies into a range of techniques are presented, together with estimates of the tritium removal rate in ITER in each case. Finally, an approach involving the integration of many tritium removal techniques into the ITER operational schedule is proposed as a means to extend the period of operations before major intervention is required.

13CH4was injected through graphite and tungsten spherical limiters in reproducible TEXTOR discharges. These materials were chosen, as they represent the actual compromise for the plasma facing components in the ITER divertor. 13C was used in order to distinguish injected and intrinsic carbon in the layer deposited on the limiter surface. Shot-by-shot video recordings show a continuous growth of the deposit near the injection hole. A pronounced difference in the 13C deposition pattern on the graphite and tungsten limiters was observed. Post-mortemsurface analysis showed that the ratios of the locally deposited to the injected amount of carbon are 4% for graphite and 0.3% for tungsten. The margins of the carbon layer deposited on tungsten are significantly steeper in comparison with the graphite limiter case. The large difference in the 13C deposition efficiency can be explained by direct reflection of carbon from tungsten and the enhanced sputtering of carbon on the tungsten substrate. Nucleation is suggested to be an important mechanism for carbon deposition on tungsten. Monte Carlo impurity transport calculations by the ERO code reproduce reasonably the experimental results for the graphite limiter.

Plasma operation with high-Z plasma facing components is investigated with regard to sputtering, core impurity contamination and scenario restrictions. A simple model based on dimensionless quantities for fuel and high-Z ion sources and transport to describe the high-Z concentration in the plasma core is introduced. The impurity release and further transport is factorized into the sputtering yield, the relative pedestal penetration probability and a core confinement enhancement factor. Since there are quite large uncertainties, in particular, in the sputtering source and the edge transport of high-Z impurities, very different scenarios covering a wide parameter range are taken into account in order to resolve the experimental trends. Sputtering of tungsten by charge exchange neutrals in the energy range 0.5–2 keV is comparable to the effect of impurity ion sputtering, while the impact of thermal fuel ions is negligible. Fast ions produced by neutral beam injection as well as sheath acceleration during ICR heating may cause considerable high-Z sources if the limiters on the low-field side have high-Z surfaces. The critical behaviour of the central high-Z concentration in some experimental scenarios could be attributed to edge and core transport parameters in the concentration model. The improved H-mode with off-central heating turns out to be the most critical one, since a hot edge is combined with peaked density profiles.

Plasma operation with high-Zplasma facing components is investigated with regard to sputtering, core impurity contamination and scenario restrictions. A simple model based on dimensionless quantities for fuel and high-Zion sources and transport to describe the high-Zconcentration in the plasma core is introduced. The impurity release and further transport is factorized into the sputtering yield, the relative pedestal penetration probability and a core confinement enhancement factor. Since there are quite large uncertainties, in particular, in the sputtering source and the edge transport of high-Zimpurities, very different scenarios covering a wide parameter range are taken into account in order to resolve the experimental trends. Sputtering of tungsten by charge exchange neutrals in the energy range 0.5-2?keV is comparable to the effect of impurity ion sputtering, while the impact of thermal fuel ions is negligible. Fast ions produced by neutral beam injection as well as sheath acceleration during ICR heating may cause considerable high-Zsources if the limiters on the low-field side have high-Zsurfaces. The critical behaviour of the central high-Zconcentration in some experimental scenarios could be attributed to edge and core transport parameters in the concentration model. The improved H-mode with off-central heating turns out to be the most critical one, since a hot edge is combined with peaked density profiles.

A series of 39 consecutive and repetitive discharges (Ip= 2MA, BT= 2.4T, = 3.8 × 1019m-3, gas rate ~1.5 × 1022Ds-1and with 2.8 MW of ICRH over a duration of 11s) has been performed in JET for a full day in order to study the particle retention behaviour as a function of the wall inventory and the global balance for a significant number of discharges associated to a high gas injection. Since the active pumping was achieved using the divertor cryopump only, its regeneration has allowed a direct calibration of the value of the pumped particle flux to be used in the particle balance analysis during the plasma operations for the “DOC-L” configuration. Taking into account the outgased flux between the discharges, the resulting wall inventory over the full day of operation is zero. During, the 11 sec of the ICRH power, about 8 % of the particles injected are retained in the machine equilibrated by a particle recovery between of 8% of the quantity injected. This shows that the gas released between pulses has been overestimated in previous JET gas balance analysis and that the particles trapped in the machine are localised in areas which are outgasing between the discharges.

The 11th EFPW took place in December 2003 at Heraklion in Crete, hosted by the Association EURATOM-Greece and the FORTH Institute, Heraklion and sponsored by the European Commission. Within the overall theme of 'plasma-wall interactions (PWI) and their implications for impurity generation and transport', four topics of importance to the future development of magnetically confined fusion were discussed in detail. Key PWI issues for ITER were also reviewed, the programmes of the two European physics task forces, on PWI and on integrated tokamak modelling, were discussed, and several topical reviews on key physics R&D issues for ITER were presented. The main issues discussed and the areas identified as requiring further study are summarized here.

AbstractThis paper gives an overview of the research activities at TEXTOR on impurity production, impurity transport through the plasma, and then deposition. First, laboratory experiments on chemical erosion by hydrogen and oxygen and radiation-enhanced sublimation are described, followed by the main part, which concentrates on the TEXTOR data of impurity release, impurity transport, and redeposition. The differences between the behavior of high-Z and low-Z materials are discussed. Many of the TEXTOR experiments are carried out using special limiter locks, but the overall carbon balance of net erosion sources and net deposition zones are also shown. Finally, modeling of erosion and dedicated transport experiments are addressed.

AbstractLimiter lock systems on the top and the bottom of the TEXTOR vessel are essential elements for experimental investigations of plasma-wall interaction in a tokamak. The lock systems are designed as user facilities that allow the insertion of wall elements (limiter) and tools for diagnostic (electrical probes, gas injection) without breaking the TEXTOR vacuum. The specially designed holder on top of the central carrier and a powerful vacuum pump system permit the exchange of components within ~1 h. Up to ten electrical signals, four thermocouples, and a gas supply can be connected at the holder interface. Between discharges, the inserted component can be positioned radially and turned with respect to the toroidal magnetic field. Additionally, the central carrier is electrically isolated to apply bias voltages and currents up to 1 kV and 1 kA, respectively.An important feature of the lock system is the good access for optical spectroscopic observation of the inserted components in the vicinity of the edge plasma. The whole spectrum from ultraviolet to infrared is covered by spectrometers and filters combined with cameras. Toroidally and poloidally resolved measurements are obtained from the view on top of the probes while the tangential poloidal view delivers radially and toroidally resolved information.A programmable logic controller (Simatic S5) that is operated inside the TEXTOR bunker and from remote locations outside the concrete wall drives all possible features of the lock system.

AbstractProper wall conditioning has been a major element in the development of fusion energy on the way to achieve high fusion plasma performance. Various of these techniques have been pioneered in the TEXTOR tokamak and later applied successfully in various devices worldwide. The main issues are to clean the surface from surface-bounded impurities, to remove hydrogen, and to coat the entire wall surface with a thin film of a proper first-wall material. The main benefits of wall conditioning are to control the oxygen impurity content of the plasma and to offer a suitable first-wall material. Entire coating of the first wall has allowed one to control to some extent the recycling hydrogenic fluxes but in particular to study the complex coupling between the choice of wall materials and the behavior of the plasma edge. This paper presents a review of the different wall-conditioning methods used in TEXTOR and their effects on the plasma behavior. Also, new wall-conditioning concepts, compatible with steady-state magnetic fields, are outlined briefly.

Recently, the dynamic ergodic divertor (DED) of TEXTOR has been studied in an m/n= 3/1 set-up which is characterized by a relatively deep penetration of the perturbation field. The perturbation field creates (a) a helical divertor, (b) an ergodic pattern and/or (c) excitation of tearing modes, depending on whether the DED current is static, rotating in the co-current direction or in the counter-current direction. Characteristic divertor properties such as the high recycling regime or enhanced shielding have been studied. A strong effect of the ergodization is spin up of the plasma rotation, possibly due to the electric field at the plasma edge. Tearing modes are excited in a rather reproducible way and their excitation threshold value, their motion and their reduction due to the ECRH/ECCD have been studied. The different scenarios are characterized by strong modifications of the toroidal velocity profile and by a reduced or enhanced radial transport.

Recently, the dynamic ergodic divertor (DED) of TEXTOR has been studied in an m/n = 3/1 set-up which is characterized by a relatively deep penetration of the perturbation field. The perturbation field creates (a) a helical divertor, (b) an ergodic pattern and/or (c) excitation of tearing modes, depending on whether the DED current is static, rotating in the co-current direction or in the counter-current direction. Characteristic divertor properties such as the high recycling regime or enhanced shielding have been studied. A strong effect of the ergodization is spin up of the plasma rotation, possibly due to the electric field at the plasma edge. Tearing modes are excited in a rather reproducible way and their excitation threshold value, their motion and their reduction due to the ECRH/ECCD have been studied. The different scenarios are characterized by strong modifications of the toroidal velocity profile and by a reduced or enhanced radial transport.

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

The main objective of this paper is investigation of methods for reduction of divertor heat loads in order to increase the lifetime of divertor tiles in future fusion reactors. Special emphasis is given to studies of reduction of transient heat loads due to edge localized modes (ELMs). Two methods are compared: argon seeded type-I ELMy H-modes and nitrogen seeded type-III ELMy H-modes. In both scenarios, the impurity seeding leads to a reduction in the pedestal energy and hence a reduction in the energy released by the ELM. This consequentially reduces the power load to the divertor targets. At high radiative power fractions in type-III ELMy H-modes, part of that released ELM energy (small ELMs, below 20 kJ) is dissipated by radiation in the scrape off layer (SOL). Modelling of the ELM mitigation supports the experimental findings. This ELM mitigation by radiative dissipation is not effective for larger ELMs. In between ELMs, the plasma is detached and radiates strongly from the X-point region. During an ELM, the nitrogen in the X-point and divertor region becomes ionized into more weakly radiating higher charge states and the plasma re-attaches for large ELMs. At JET, argon radiates predominantly in the main plasma and not so much in the cold divertor region. Hence, the effect of radiative dissipation of ELM heat fluxes by argon is very low due to the limited argon density in the divertor region. Nevertheless, both scenarios might be compatible with an integrated ITER scenario, with respect to acceptable divertor lifetime and acceptable confinement.

This contribution presents two-dimensional Monte-Carlo simulations of the local transport of hydrocarbons that are chemically eroded in the JET MkIIa divertor. The effect of a given background carbon flux flowing from the main plasma down to the divertor is also taken into account. The influence of local plasma temperatures and densities and the influence of different assumptions for the sticking of re-deposited particles hitting the tiles is analysed. Even under the assumption that the sticking of re-deposited hydrocarbon fragments is zero, a large amount (about 75&percent;) of the eroded particles is re-deposited in the form of ionized carbon on the tiles. A reasonable agreement between the simulation and the experimental observation of the carbon deposition at the inner louvers in the MkIIa JET divertor is achieved, if high chemical erosion yields of about 20&percent; together with a negligible effective sticking of hydrocarbons are assumed. Although there are experimental observations indicating such high erosion yields, such yields cannot be applied as a stationary value for the effective erosion since it would turn the inner divertor into a net erosion area in contrast to the experimental findings. The possibility of applying different erosion rates for (re)-deposited layers and the substrate material is discussed. To further improve the modelling a better knowledge about the formation and erosion of re-deposited layers is required. A first conclusion concerning the erosion and re-deposition behaviour in ITER will be drawn.

Results of a novel technique of measuring the deposited power profile based on thermocouples embedded in the divertor plates are presented and discussed. Power profiles were obtained for a series of 2.5 MA/2.4 T discharges involving a scan in the NBI heating power (4-18 MW, where P_L-H~5 MW) and a scan in D₂ fuelling rate ((0-3)×10²² s^-1). In H-mode plasmas, a narrow feature was observed in the power profile, which scaled inversely with power entering the scrape-off-layer, λ_q,nw∝P_SOL^-0.4±0.1, while the base profile showed virtually no power dependence. At constant input power, the peak heat flux was strongly reduced by D₂ puffing (from ~20 to 4 MW m^-2), with the narrow feature in the power profile effectively suppressed. Based on numerical analysis using a guiding-centre Monte-Carlo code ASCOT, coupled to a fluid-like plasma-neutral code package OSM2/EIRENE, the above results are interpreted as a consequence of inter-ELM ion orbit loss from the pedestal region.

One of the most severe problems for fusion reactors is the power load on the divertor target plates. Technically only power loads of less than 10 MW m^-2 are acceptable. However, strong edge localized mode (ELM) activity can lead to power loads in excess of 800 MW m^-2. In order to reduce the steady-state heat flux and the transient heat flux due to ELMs, radiation cooling experiments were performed at JET. Nitrogen was puffed into the divertor up to a radiative power fraction of 90%. This was achieved at a density of 0.85 times the Greenwald density (Greenwald M 1988 Nucl. Fusion 28 2199), while maintaining an H-factor of f_H98 = 0.82. The Z_eff in all those discharges stayed around 2.0. At approximately 55% radiative power fraction, the ELM characteristic changes from type I to type III, resulting in a loss of confinement of about 25% due to a degradation of the edge pedestal and hence a reduction of the ELM power load to the divertor tiles. By increasing the radiative power fraction to values of about 90%, the heat flux is reduced to 2 MW m^-2.

Results of a novel technique of measuring the deposited power profile based on thermocouples embedded in the divertor plates are presented and discussed. Power profiles were obtained for a series of 2.5 MA/2.4 T discharges involving a scan in the NBI heating power (4-18 MW, where P_L-H~5 MW) and a scan in D₂ fuelling rate ((0-3)×10²² s^-1). In H-mode plasmas, a narrow feature was observed in the power profile, which scaled inversely with power entering the scrape-off-layer, λ_q,nw∝P_SOL^-0.4±0.1, while the base profile showed virtually no power dependence. At constant input power, the peak heat flux was strongly reduced by D₂ puffing (from ~20 to 4 MW m^-2), with the narrow feature in the power profile effectively suppressed. Based on numerical analysis using a guiding-centre Monte-Carlo code ASCOT, coupled to a fluid-like plasma-neutral code package OSM2/EIRENE, the above results are interpreted as a consequence of inter-ELM ion orbit loss from the pedestal region.

ABSTRACTIn order to investigate tritium behavior in tokamak, we have measured surface distributions of deuterium and tritium on graphite limiter tiles used in TEXTOR under D-D operation by means of an ion beam analysis and tritium imaging plate technique, respectively.It was found that both distributions were quite different, i.e. deuterium retention was higher at the deposited area, whereas tritium retention was higher at the erosion dominated area. This is because tritium produced by the D-D reaction, initially having 1 MeV, did not fully lose its energy in the TEXTOR plasma and implanted into the plasma facing materials nearly homogeneously, whereas deuterium was codeposited with carbon and boron, the main impurities in the TEXTOR plasma. This is also confirmed by the finding that high level of tritium was detected beneath the deposited layer. Tritium distribution, however, was modified by the temperature increase due to plasma heat load.Thus the comparison of tritium profiles with the deuterium profile gives a large amount of important and new information on PMI, and may be used as a new diagnostic technique for PMI.

The increase in pulse duration and cumulative run-time, together with the increase of the plasma energy content, will represent the largest changes in operation conditions in future fusion devices such as the International Thermonuclear Experimental Reactor (ITER) compared to today's experimental facilities. These will give rise to important plasma-physics effects and plasma–material interactions (PMIs) which are only partially observed and accessible in present-day experiments and will open new design, operation and safety issues. For the first time in fusion research, erosion and its consequences over many pulses (e.g., co-deposition and dust) may determine the operational schedule of a fusion device. This paper identifies the most critical issues arising from PMIs which represent key elements in the selection of materials, the design, and the optimisation of plasma-facing components (PFCs) for the first-wall and divertor. Significant advances in the knowledge base have been made recently, as part of the R&D supporting the engineering design activities (EDA) of ITER, and some of the most relevant data are reviewed here together with areas where further R&D work is urgently needed.