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Surface topography of and deuterium (D) retention in reduced activation ferritic-martensitic Eurofer’97 and ferritic oxide dispersion strengthening ODS-16Cr steels have been studied after exposure at 600 K to low-energy (70 and 200 eV), high-flux (∼1022 D/m2s) pure D and D-10%He plasmas with D fluence of 2 × 1025 D/m2. The methods used were scanning electron microscopy, energy-scanning D(3He,p)4He nuclear reaction, and thermal desorption spectroscopy. As a result of the plasma exposures, nano-sized structures are formed on the steel surfaces. After exposure to pure D plasmas, a significant fraction of D is accumulated in the bulk, at depths larger than 8 μm. After exposures to D-He plasmas, D is retained mainly in the near-surface layers. In spite of the fact that the He fluence was lower than the D fluence, the He retention in the steels is one order of magnitude higher than the D retention.

The thermal evolution of vacancies and vacancy clusters in tungsten (W) has been studied. W (100) single crystals were irradiated with 200 keV hydrogen (H) ions to a low damage level (5.8 x 10(-3) dpa) at 290 K and then annealed at temperatures in the range of 500-1800 K. The resulting defects were characterized by positron annihilation lifetime spectroscopy (PALS) and positron annihilation Doppler broadening spectroscopy (DBS). Annealing at 700 K resulted in the formation of clusters containing 10-15 vacancies, while at 800 K and higher temperatures clusters containing about 20 vacancies or more were formed. Reduction of the defect concentration likely accompanied by further coarsening of the clusters started at 1300 K and ended at 1800 K with the complete defect recovery. The determined cluster sizes at 700 K and 800 K were larger than the estimated minimum cluster sizes that are thermally stable at these temperatures, indicating that the migration and ensuing coalescence of small clusters plays an important role in cluster growth. (C) 2020 Max-Planck-Institut fur Plasmaphysik. Published by Elsevier B.V. All rights reserved.

Hydrocarbon film deposition was studied with cavity samples in remote areas of the inner and outer JET divertor and below the divertor septum during the 1999–2001 and 2005–2009 campaigns. Thick hydrocarbon films were formed inside the cavities. These deposited hydrocarbon layers have high D/C ratios close to 1. The formation of these films is mainly due to sticking of hydrocarbon particles with high surface loss probabilities >0.6. The observed surface loss probabilities depend on the position in the divertor and vary during different campaigns. The particles responsible for hydrocarbon layer formation originate from the divertor strike points. Except for the septum cavity the deposition of beryllium was very low and showed a very different distribution from that of deuterium and carbon.

Samples made of tungsten doped either with titanium carbide (W–1.1TiC) or tantalum carbide (W–3.3TaC) were either exposed to D 2 gas at a pressure of 100 kPa at 800–963 K or irradiated by 38 eV/D ions at 800 K. The deuterium (D) inventory in the samples was examined by nuclear reaction analysis and thermal desorption spectroscopy. The D bulk concentration and total retention in W–3.3TaC were comparable in all cases to that in pure polycrystalline W. The D bulk concentration in W–1.1TiC was more than one order of magnitude higher than that in pure W after exposure to D 2 gas, and was also several times higher than that in W–1.1TiC after irradiation at 800 K. It is suggested that D trapping inside the carbide precipitates in W–1.1TiC becomes essential at high temperatures.

Selected Ion Beam Analysis techniques applicable for detecting deuterium and heavier impurities have been used in the post-mortem analyses of tiles removed after the first JET ITER-Like Wall (JET-ILW) campaign. Over half of the retained fuel was measured in the divertor region. The highest figures for fuel retention were obtained from regions with the thickest deposited layers, i.e. in the inner divertor on top of tile 1 and on the High Field Gap Closure tile, which resides deep in the plasma scrape-off layer. Least retention was found in the main chamber high erosion regions, i.e. in the mid-plane of Inner Wall Guard Limiter. The fuel retention values found typically varied with deposition layer thicknesses. The reported retention values support the observed decrease in fuel retention obtained with gas balance experiments of JET-ILW.

The erosion of Be and W marker layers was investigated using long-term samples during the first ITER-like wall discharge campaign 2011-2012. The markers were mounted in Be coated Inconel tiles between the inner wall guard limiters (IWGL). They were analyzed using Rutherford backscattering (RBS) before and after exposure. All samples showed strong erosion. The results were compared to the data for Be and W erosion rates for the 2005-2009 and the 2001-2004 campaigns, respectively, when JET was operated with a carbon wall. In 2005-2009 Be and C samples were used, and W samples were used in 2001-2004. The mean W erosion rates and the toroidal and poloidal distributions of the W erosion were the same for the 2001-2004 and the 2011-2012 campaigns. The mean erosion rate of Be during the 2011-2012 campaign was smaller by a factor of about two compared to the 2005-2009 campaign and showed a different poloidal distribution. The mean erosion rate of the inner JET ITER-like wall was about 4-5 times smaller than the mean erosion rate of the carbon wall.

The mapping of hydrogen distributions in 3 dimensions and its correlation with structural features allow further insight into mechanisms of hydrogen trapping in tungsten. We studied hydrogen distributions in 25 μm thick polycrystalline tungsten foils by 3D hydrogen microscopy using a proton–proton-scattering method. Two types of tungsten samples were prepared: (i) at 1200 K annealed foils and using 1.8 MeV implantation energy (ii) at 2000 K annealed foils using 200 eV implantation energy. It has been found that large variations of surface hydrogen contamination occur within different samples. Nevertheless, a statistically significant variation of the hydrogen content across grain boundaries has been observed.

Deuterium retention in and gas-driven permeation through the new ferritic–martensitic steel Rusfer EK-181 (Fe–12Cr–2W–V–Ta–B–C) developed in the Russian Federation is studied. The retention of deuterium was investigated in the temperature range of RT-773 K at pressures 10 and 10 4 Pa. The amount of deuterium retained strongly increases with increasing temperature. The permeation of deuterium through tubes of EK-181 was investigated at gas loading in the pressure range of 5 × 10 −2 –100 Pa and in the temperature range of 573–923 K. In order to study the influence of surface impurities the front surfaces of the tube were cleaned by an argon plasma discharge, as a result the permeating flux was increased by a factor of 3.

We have studied the migration of 13C in ASDEX Upgrade after a global impurity injection experiment in 2011. The main chamber was observed to be the largest deposition region for carbon: almost 35% of the injected atoms end up there. Moreover, gaps between wall tiles account for surface densities which are comparable to those on the plasma-facing surfaces. SOLPS modeling of the experiment produced a set of background plasmas and poloidal flow profiles for simulating the transport of 13C with ASCOT; a match with measured deposition, however, required using an imposed flow profile. ASCOT reproduced the observed localized deposition at the outer midplane but work is needed to explain the measured deposition at the inner side of the torus and at the top of the vessel.

Deuterium trapping in graphite under high flux plasma exposure at high temperatures was studied. A target of MPG-8 fine grain graphite was exposed to plasma beam in LENTA device at zero potential. The average ion flux was of about 10 21 D/m 2 s with a maximum in the center. The temperature was 1473 K in the center and 873 K outside the beam spot. After irradiation, the target was sectioned along and across the surface and studied by SEM, NRA, and TDS. Deuterium profiles measured by NRA showed a peak close to the surface and long tails in the bulk. TDS of samples taken for the front and back sides of the target were very similar. It is suggested, that D retention at the plasma irradiation is mainly controlled by trapping from the gas phase, locally created in the graphite pores by plasma ions injection and subsequent flow of gas to the back side through internal porosity.

At FOM Rijnhuizen, linear plasma generators are used to investigate plasma-material interactions under high-density (≤1021 m -3), low-temperature (≤5 eV) plasma bombardment. Research into carbon-based materials has been focused on chemical erosion by hydrogen plasmas. Results from plasma exposure to high-flux (>1023 H +/m2 s) and low-temperature hydrogen plasma indicate silicon carbide has a lower relative rate of gross erosion than other carbon-based materials (e.g. graphite, diamond, carbon-fiber composites) by a factor of 7-10. Hydrogenic retention is the focus of research on tungsten and molybdenum. For target temperatures of 700-1600 K, the temperature dependence of hydrogenic retention is the dominant factor. Damage to the surface by heavy ion irradiation has shown to enhance retention by a factor of 2.5-4.1. Thermal stressing of W via. e-beam thermal cycling also enhances hydrogenic retention by a factor of 2.1 ± 0.2, likely due to the introduction of thermal defects, which act as trapping sites for implanted hydrogenic isotopes.

Experiments on deuterium ion-driven permeation through tungsten foils with thicknesses of 50 and 360 μm in the temperature range 820–920 K are described and discussed. The lag time for permeation was four orders of magnitude higher than estimated by extrapolation of Frauenfelder’s diffusion values in undamaged material. The experiments can be described by assuming trapping sites with a detrapping energy of 2.05 ± 0.15 eV in tungsten bulk. We assume that trapping is due chemisorption on the surface of voids in the bulk of tungsten, which, according to literature, have a similar detrapping energy. Two kinds of voids (pores and cracks) were observed by scanning electron microscopy of sample cross-sections prepared by a focused ion beam. The recombination coefficients on the inlet side were also deduced from the permeation experiments and compared with theoretical and experimental data from literature.

Hydrogen gas-driven permeation through graphite coated with different types of tungsten coatings with thicknesses up to 200 lm has been investigated. The substrate material was fine-grain graphite R5710 and R6710, which properties with respect to hydrogen transport are well known [1]. Magnetron-sputtered W coatings of thicknesses 1 and 3 lm and two coatings of ASDEX Upgrade were investigated: A 3 lm thick layer of tungsten deposited by physical vapor deposition (PVD-W) and 200 lm thick layers of tungsten deposited by plasma-spaying in vacuum (VPS-W). The gas-driven permeation was investigated at a pressure gradient of 10 � 2

In order to benchmark predictions for the in vessel tritium inventory in ITER, a survey of fuel retention measured in 4 carbon dominated tokamaks (TEXTOR, ASDEX Upgrade in the 2002–2003 carbon configuration, Tore Supra and JET) was performed, showing retention rates from ∼1 g D/h in TEXTOR (L mode, limiter machine) up to ∼6–12 g D/h in AUG (H mode, divertor machine). A simple scaling used for ITER predictions is applied for comparison with experimental values: (1) estimate of wall fluxes, (2) estimate of the gross carbon erosion, (3) estimate of the net erosion/redeposition assuming a redeposition fraction and (4) estimate of the retention rate using D/C ratio scalings. The validity of each step is discussed, showing that this approach yields the right order of magnitude, but tends to underestimate the experimental values unless a high wall flux, a low local redeposition fraction and/or a high D/C ratio are used.

In investigations of hydrogen retention in first wall components the influence of the conditions of the implanted target storage prior to analysis and the storage time is often neglected. Therefore we have performed a dedicated set of experiments. The release of hydrogen from samples exposed to ambient air after irradiation was compared to samples kept in vacuum. For air exposed samples significant amounts of HDO and D 2 O are detected during TDS. Additional experiments have shown that heavy water is formed by recombination of releasing D and H atoms with O on the W surface. This water formation can alter hydrogen retention results significantly, in particular – for low retention cases. In addition to the influence of ambient air exposure also the influence of storage time in vacuum was investigated. After implantation at 300 K the samples were stored in vacuum for up to 1 week during which the retained amount decreased significantly. The subsequently measured TDS spectra showed that D was lost from both the high and low energy peaks during storage at ambient temperature of ∼300 K. An attempt to simulate this release from both peaks during room temperature storage by TMAP 7 calculations showed that this effect cannot be explained by conventional diffusion/trapping models.

A specific experiment was performed in Tore Supra, dedicated to the search for consistency between wall inventories estimated from gas balance and post-mortem analysis and to the characterization of the D-retention mechanism. More than 160 similar discharges were performed, representing a cumulative time of 18 000 s of plasma with no intermediate conditioning procedure. The only significant operational issue was linked to the ejection of flakes from the plasma facing components, whose frequency increased dramatically during the campaign, triggering a plasma detachment phase followed by a disruption in number of cases. In-vessel inventory was increased by ∼3.1 × 1024 D, and constant retention rate (∼2.3 × 1020 D/s averaged over discharge duration) was measured, with no indication of wall saturation. First measurements on dismantled tiles of the Toroidal Pump Limiter (TPL) show [D]/[C] ratios ranging from 0.04 to 0.20, on a typical thickness of 15–20 μm. Extrapolated to the whole TPL, this yields an amount of ∼(1.5 ± 0.3) × 1024 D, i.e. ∼50% of the estimated in-vessel inventory.

The main objectives of the JET ITER-like Wall Project are to provide a beryllium main wall and tungsten divertor with at least a 4 year lifetime to allow full evaluation of the materials and related plasma scenarios for ITER. Tungsten coatings will be used over most of the divertor area and this paper describes the latest developments in the coating technology and an analysis of the implications for the coating lifetime and machine operation. Both steady state and transient heat loads are assessed. (C) 2009 Published by Elsevier B.V.

Deuterium depth profiles in graphite tiles of JT-60U have been determined by means of the D( 3 He, p) 4 He nuclear reaction. Tiles with net carbon deposition showed high deuterium retention. A D/C ratio of 0.07 was found in the outer dome wing tile. In the inner divertor tile covered with a thick deposition layer, the depth profile shows a broad peak in a deep region. This indicates that deuterium trapped in the shallow region can be replaced by hydrogen with hydrogen discharges due to isotope exchange, while deuterium trapped in the deeper region can not. Deuterium in the redeposited layers can be more easily removed than that in the substrate. The deuterium concentration in the outer divertor target tile was very low (D/C < 0.003). The highest D/C ratio, 0.09, was found in the upper region of the first wall.

H and D retention characteristics in the near top surface of the JT-60U plasma facing carbon tiles were investigated by means of elastic recoil detection analysis. From the result, H was dominantly retained within 1 μm depth because of the H discharges for the wall cleaning carried out before the removal of sample tiles. (H + D)/C ratios were small: it was ∼0.07 at a maximum in the divertor samples except for the deposition on the dome outer wing tile which was ∼0.13, and it was ∼0.15 at a maximum in the first wall samples. The hydrogen concentration in the redeposited carbon layers is strongly influenced by the temperature. Consequently, (H + D)/C ratios at the observed low level are a result of the operation of JT-60U at wall temperatures of 300 °C or higher.

Carbon-13 labelled methane was injected into the outer divertor during a series of H-mode discharges on the last day of operations with the JET MkII-SRP divertor. Tiles from around the vessel were removed during the subsequent shutdown and surface deposits were analysed by IBA techniques and SIMS. First attempts to model the pattern of 13 C deposition using EDGE2D are reported. Erosion of W markers at the outer divertor was observed, with implications for the ITER-like wall experiment planned for JET, whilst thin film growth in the same region has been followed by the effect on infrared measurements. The composition of thick films deposited at the inner divertor during the MkII-SRP campaign, and the migration to the inner corner of the divertor observed by a quartz micro-balance, provide further information on divertor transport. Crown Copyright (c) 2007 Published by Elsevier B.V. All rights reserved.

Depth profiles of deuterium trapped in W single crystals and polycrystalline W irradiated with 200 eV D ions at 300–323 K have been measured up to a depth of 7 μm using the D( 3 He,p) 4 He nuclear reaction in a resonance-like technique. The proton yield as a function of incident 3 He energy was measured and the D depth profile was obtained by deconvolution of the measured proton yields using the program simnra . The depth at which deuterium is retained can be tentatively divided into three zones: (i) the near-surface layer (up to a depth of ∼0.2 μm), (ii) the sub-surface layer (from ∼0.5 to ∼2 μm), and (iii) the bulk (>5 μm). Low-energy D ion irradiation modifies the W structure to depths of up to about 5 μm, both for W single crystals and polycrystalline W. Up to this depth trapping sites are created at a fluence of 5 × 10 22 /m 2 for the retention of deuterium at about 0.1 at.%.

Numerous traces of arcs have been found on W-covered graphite tiles of ASDEX Upgrade after exposure. The distributions of number density, lengths and orientation are calculated and compared to pure graphite tiles at comparable locations. It was established that arcs perforate a 1 μm tungsten layer down to the carbon substrate. The amount of removal should rise with arc current, but a surface fraction of about 8% is eroded at 10 A already. At tiles of the divertor baffle the layer is continuously removed along the entire track pointing to higher currents. The carbon of the stripped parts is subject to subsequent erosion processes. The distribution of materials in and around arc tracks was investigated by sputter depth profiling (SIMS and AES) and the characteristic geometry was studied using an electron microscope. Observations are interpreted using results from laboratory vacuum arcs on the same material.

This paper presents the results of a modelling study conducted to estimate the deposition of hydrocarbon radicals in the regions of the ITER divertor hidden from the plasma (e.g., underneath the private flux region dome and in the pumping ducts). A Monte-Carlo code based on the approach described in Ref. [M. Mayer et al., J. Nucl. Mater. 313–316 (2003) 429] has been developed and applied for the actual ITER geometry. It allows for a full 3-dimensional treatment of particle trajectories and follows individual hydrocarbon particles until they either stick on the surfaces or leave the system. The cases analysed clarify the influence of several parameters such as the sticking coefficient of the hydrocarbon species, the pressure of the background gas, and the geometry. Consistent with experimental findings from tokamaks, this assessment shows that particles with relatively high sticking probability (>0.01) deposit primarily underneath the divertor structure, and only species with very small sticking probability (⩽10 −3 ) may reach and stick along the inner surfaces of the pumping ducts.

To investigate material transport in scrape-off layer plasma and in/out asymmetries in divertor, 13CH4 was puffed at the end of 2002/2003 experimental campaign into ASDEX Upgrade from the outer mid-plane. Ex-situ analyses of the tiles were performed by secondary ion mass spectrometry and Rutherford backscattering spectrometry. The highest amount of 13C was detected at the top of inner divertor, and below the inner and outer strike points towards the roof baffle. The 13C peaks below the strike points correlate with thick long term deposition layers detected in ion beam analysis experiments.

Carbon deposition and erosion were measured on ASDEX Upgrade divertor tiles and below the roof baffle during the operation period 2002/2003. The inner divertor is a net carbon deposition area, while a large fraction of the outer divertor is erosion dominated and the roof baffle tiles show a complicated distribution of erosion and deposition areas. In total, 43.7 g B + C were redeposited, of which 88% were deposited on tiles and 9% in remote areas (below roof baffle, on vessel wall structures). 0.6 g C was pumped out as volatile hydrocarbon molecules. Carbon sources in the main chamber are too low by a factor of more than ten to explain the observed carbon divertor deposition. Carbon erosion is observed at the outer divertor strike point tiles, but it is arguable if material can be transported from the outer strike point to the inner divertor.

To minimize radiation losses of the plasma during long pulse operation, the first wall protection of WENDELSTEIN 7-X demands low-Z plasma facing materials. In addition to carbon materials on high heat flux loaded components, 300–500 μm thick vacuum plasma-sprayed (VPS) layers of boron carbide (B4C) are considered as coatings on large actively cooled stainless steel panels. In order to evaluate the behaviour relevant to the expected plasma wall interactions, an extensive material characterization and test programme was executed. Also included is the examination of the industrial manufacturing and the suitability of these coatings as 70 m2 first wall plasma facing material for W7-X. To improve the adhesion of thick B4C layers on stainless steel, different interlayers have been investigated. Based on the results, a VPS-based coating technique was identified which is suitable to manufacture the B4C protection layers on stainless steel wall panels of W7-X.

Plasma sprayed B4C-layers considered as wall coatings for the W7X stellarator have been studied during and after exposure to TEXTOR and after arcing experiments in vacuum. Arcing through the B4C layer occurred favoured by high power fluxes and not restricted to less stable phases. But this arcing implies an especially noisy scrape-off layer (SOL). Instead of moving retrograde in the external magnetic field, the arc spot on the B4C-layer sticks to the same location for its whole lifetime. Consequently, the arc erodes the entire B4C layer, finally burning down to the Cu substrate. In the neighbourhood of craters the surface contains Cu originating from those craters. This material, hauled to the surface by the arc, is subject to subsequent erosion, transport, and redeposition by the SOL-plasma. The behaviour of arcs on B4C is most probably caused by the peculiar temperature dependences of the electrical and heat conductivity of B4C. 2003 Elsevier Science B.V. All rights reserved.

The key problem for the use of carbon in a future fusion device is the formation of tritium containing a-C:H layers. ASDEX Upgrade offers the possibility to investigate these layers at ITER relevant divertor conditions. Long term probes show, that the layer growth under the structure of the new divertor (Div IIb) is very similar to Div II. Additional quartz microbalance monitors offers measurements on the layers growth on a shot-to-shot base. The layers are found to grow continuously during the campaign. Using similar shots, a proportionality of the growth rate to the divertor neutral pressure is found. A Langmuir probe, installed below the divertor structure, measured a strongly variable plasma with electron densities up to 1×1018 m−3. The density of this parasitic plasma is correlated to the divertor radiation and neutral density, which point to photo ionisation or photo emission as the origin of the plasma.

A new method for the determination of massive erosion and deposition on plasma facing components has been developed and was tested successfully on a graphite tile of the ALT-II limiter of TEXTOR. The surface profile of the tile was measured before and after exposure to plasma discharges with an optical profiler, erosion or deposition is determined from the difference of the two profiles. The profiles were determined relative to specially machined holes, which provide stable reference points. An accuracy of about 1 μm can be achieved. After exposure for 7625 plasma seconds a maximum erosion of 28 μm carbon is observed in erosion dominated areas, while a maximum deposition of about 40 μm is observed in net redeposition areas. The composition and structure of the redeposited layers were investigated with secondary ion mass spectroscopy and scanning electron microscopy.

Carbon sources and the sinks have been quantified in TEXTOR and are discussed in terms of short and long range transport. The major source (22 g/h) is the graphite belt limiter, but part (10 g/h) of the carbon is directly re-deposited after short range transport. Long range transport causes flake formation on obstacles and neutralisers, but little and deuterium rich (D/C≈0.7) deposition in remote areas. The rest is leaving via the pumps in gaseous form. This behaviour is different from that in JET where large amounts of deuterium rich deposits were found in the louvers. Tungsten is favoured for the ITER divertors because of its low sputtering yield for hydrogen, but melting and erosion by carbon may be an additional concern. The short range transport of tungsten has been investigated in a well defined experiment and quantitatively re-constructed by means of the ERO-TEXTOR code. Code validation is necessary in order to increase the confidence and the applicability to JET and ITER.

The paper reviews hydrogen isotope retention and migration in tungsten (W). Due to a large scatter of the deuterium (D) retention database, new measurements of ion-driven D retention in polycrystalline W foil have been performed to clarify the mechanism of hydrogen isotope inventory in W. Deuterium retention has been investigated as a function of ion fluence, implantation temperature, incident energy and surface conditions. Special attention has been given on the investigation of D retention in thin films of tungsten carbide and tungsten oxide which can be formed on W surface in a fusion device. Such kinds of films increase the D retention in W. Several points are reviewed: (i) inventory in pure W, (ii) inventory in W pre-implanted by carbon ions and (iii) inventory in tungsten oxide.

A set of Be and W tiles removed after the first ITER-like wall campaigns (JET-ILW) from 2011 to 2012 has been analysed. The results indicate that the primary erosion site is in the main chamber (Be) as in previous carbon campaigns (JET-C). In particular the limiters tiles near the mid-plane are eroded probably during the limiter phases of discharges. W is found at low concentrations on all plasma-facing surfaces of the vessel indicating deposition via plasma transport initially from the W divertor and from main chamber W-coated tiles; there are also traces of Mo (used as an interlayer for these coatings). Deposited films in the inner divertor have a layered structure, and every layer is dominated by Be with some W and O content.

The recovery of radiation defects in recrystallized tungsten was studied in the temperature range 400–1150 K. The defects were created by 20 MeV tungsten ions irradiation at room temperature to a maximum damage level of 0.9 dpa. The samples were then annealed to temperatures of 400–1150 K for 1 hour in order to anneal the created defects. After annealing the remaining defects were decorated with D by exposing them to a low-temperature ECR-plasma at 400 K to a fluence of 1 × 10 25 D/m 2 . Deuterium was detected by nuclear reaction analysis and temperature programmed desorption. Annealing of defects is observed already at slightly elevated temperatures. At 820 K about 50% of the initial defects were annealed, while still about 30% of the initial defects were present after annealing at 1150 K.

The composition of co-deposited hydrocarbon-silicon layers (a-C:Si:D) with varying Si concentrations and their removal by heating in air were investigated using MeV ion beam techniques. The amount of trapped D per re-deposited target atom depends weakly on the Si concentration. For pure C and Si, the D concentrations are about 0.45 and 0.5 D atoms per re-deposited target atom at room temperature, respectively. A maximum of about 0.7 D/(Si+C) was found at Si/C≈1. For increasing deposition temperature the D concentration does not decrease significantly until about 600 K. At about 1000 K the D concentration for pure C layers is still about 30% of the concentration at room temperature. The removal rates of D and C by heating in air increase strongly at temperatures around 550 K for a-C:D layers. With increasing Si content, these temperatures rise to above 650 K for layers with Si concentrations higher than 0.2 Si/(Si+C). The C removal rate is always lower than the D removal rate. Si is not removed by this method. For comparison, the composition of co-deposited stainless steel layers and Ti–C mixtures were investigated.

Two different sets of long term samples (LTS) were exposed in TEXTOR-94 at the liner with boronized and siliconized wall conditions. Measured erosion/deposition rates of the coating constituents were compared with sputtering rates calculated for pure elements. D0-fluxes at the LTS locations were determined by the B2-EIRENE code. Over the whole liner, erosion of initial coatings by D0 is observed. Boron erodes at a rate of 8.4×10 13 B cm −2 s −1 , while silicon erodes at a rate of 5.6×10 13 Si cm −2 s −1 , a factor of 1.5 lower than that of boron. With boronized walls, carbon atoms initially present in the a-C/B:D layers are eroded at a rate of 7.6×10 12 C cm −2 s −1 , at the same time 1.7×1012 (Fe + Cr + Ni) cm −2 s −1 are deposited. Compared to boronized conditions where carbon erodes, deposition of carbon (1.7×10 13 cm −2 s −1 ) and of metals (3.8×10 12 cm −2 s −1 ) occurred during the Si campaign. Due to long time exposure at high wall temperature (350°C), hydrogenic isotopes desorb; (D + H)/(B + C) and (D + H)/(Si + C) ratios degrade from originally ∼0.4 to ∼0.2.

Hydrogen retention in tokamaks is due to implantation into plasma-facing materials and trapping in deposited layers. In the limiter tokamak TEXTOR-94 hydrogen-rich deposited layers with thicknesses up to 1 mm are observed on recessed parts of the limiters, areas perpendicular to the magnetic field in the scrape-off layer (SOL), neutralizer plates of the pumped limiter and inside the pumping ducts. In the divertor tokamak JET the main deposition is observed in the divertor, additional deposits are observed in the main chamber on the sides of the guard limiters. Codeposition of carbon ions with hydrogen is the major mechanism of layer growth at areas with direct plasma contact. At remote areas without direct plasma contact, sticking of neutral hydrocarbon radicals seems to play an important role for hydrogen trapping.

13 CH 4 was injected into 18 consecutive ohmic discharges (108 s) through a hole in a graphite block equipped with an aluminium plate. CH and CII radical emission adjacent the hole was observed by spectroscopy, while the incremental grow of the deposit was measured in situ by colorimetry. SIMS depth profiling yielded the fractions of 13 C and of the 12 C out of the background. Ratios up to 13 C / C ≈0.42 were found although the deposition efficiency for 13 C was only about 0.2%. H and D were co-deposited up to ratios of 0.4 and 0.2, respectively, as measured by NRA and ERDA. The results can be simulated quantitatively by means of the ERO-TEXTOR code if high re-erosion is assumed for the hydrogen rich radicals CH x + . Within the interaction depth the 13 C/C ratio seems to become quasi-stationary. 13 C transported to obstacles creates a ratio of 13 C / C ≈0.27 after 108 s.

The composition of co-deposited carbon–silicon layers (a-C:Si:D) with varying Si concentrations and their removal by heating in air were investigated using MeV ion beam techniques. The ion-induced release of D due to the analysing beam (1.2 MeV 3He) was determined. The removal rates of D and C by heating in air increase strongly at temperatures around 550 K for a-C:D layers. With increasing Si content, these temperatures rise to above 650 K for layers with Si concentrations larger than 0.2 Si/(Si+C). The C removal rate is always lower than the D removal rate. Si is not removed by this method. The observed properties of the layers are compared with those of hard and soft a-C:D films.

The co-deposition of deuterium with silicon doped carbon for silicon concentrations between 0–100 at.% in the temperature range from room temperature to 1000 K has been investigated. The eroded material from various different targets was caught on collectors together with the reflected D to build up the co-deposited layers, which were analysed with MeV ion beam techniques. The amount of trapped D per re-deposited target atom depends weakly on the Si concentration. The maximum of about 0.7 D/(Si + C) was found at Si/C ≈ 1. For pure C and pure Si the D concentration is about 0.45 and 0.5 D atoms per re-deposited target atom at room temperature, respectively. For increasing deposition temperature the D concentration does not decrease significantly until about 600 K. At about 1000 K the D concentration for pure carbon layers is still about 30% of the concentration at room temperature. Also, co-deposited layers of stainless steel and of titanium–carbon mixtures were investigated.

Erosion and deposition at the vessel walls of the main chamber of JET were measured with long term samples during the whole operation period of the Mark I carbon divertor from April 1994 until March 1995. Assuming toroidal symmetry, about 70 g Ni + Cr + Fe was found to be eroded from the inner torus wall and about 55 g Be from the outer torus wall due to sputtering by energetic charge-exchange neutrals. Deposition has been measured on a poloidal section of the Mark I C divertor and on tiles from the inner and outer wall limiters. Eroded material is redeposited in the divertor and in the main chamber where it is found predominantly at the sides of the poloidal limiters and at the outer vessel wall. At these deposition-dominated areas in total about 25 g Ni + Cr + Fe, 28 g Be and 370 g C are found. The carbon originates mostly from the carbon limiters; additionally some may be originating from the divertor plates.

The retention of 1 keV and 8 keV deuterium implanted in B, Si, and Ti doped graphites and in the carbides B 4 C, SiC and TiC up to 6.7 × 10 19 D atoms/cm 2 has been studied at room temperature with thermal desorption spectroscopy (TDS) and ion beam analysis. Compared to pyrolytic graphite all doped graphites showed a tendency to increased retention of deuterium, especially the materials with high porosity. For the carbides the saturation concentration in the implantation zone in B 4 C is about 0.4 D per target atom and therefore comparable to the saturation concentration of D in graphite. SiC exhibits a significant higher saturation concentration of 0.6 D per target atom, whereas TiC shows a significant lower saturation concentration of 0.15 D per target atom.

In experiments with magnetically confined hot plasmas in respect to controlled thermonuclear fusion, such as tokamaks or stellerators, the surface layers of the vessel walls are modified by the plasma by erosion, redeposition, hydrogen isotope implantation and heating due to the power load from the plasma. Thus, the composition and structure of the surface layers are finally different to those of the material initially installed. This new material at the surface layers of tokamak experiments is sometimes also named ‘Tokamakium’. Generally, all elements ever introduced into the vessel can be found in the surface layers of the plasma-facing wall tiles. At all areas of the plasma-facing components both erosion and deposition have been observed including the deposition of metal droplets. Some areas are erosion dominated, while at others deposition dominates with atomic depositions of up to several μm, which partly flake. The elements of dust particles introduced into the vessel by in-vessel works get mostly incorporated as impurities into the surface layers of the plasma facing material.

The trapping of energetic deuterium codeposited with beryllium, carbon and tungsten has been measured on a silicon collector probe at room temperature. The subsequent release of the trapped deuterium at elevated temperatures was determined by thermal desorption spectroscopy. At room temperature, deuterium codeposits both with carbon and BeO with a ration of 0.41 D-atorns/C-atom and 0.38 D-atoms/BeO, respectively. No codeposition of deuterium with tungsten is observed. The thermal release of codeposited hydrogen from BeO begins at 400 K. All hydrogen is released at temperatures above 800 K.

Investigations of impurity accumulation phenomena in ASDEX are reviewed. There are four different operating regimes where pronounced accumulation is observed and these regimes are also characterized by improved energy confinement. In particular, medium-Z metallic ions are involved in accumulation processes whereas low-Z ions appear almost unaffected. The rapid accumulation observed in the case of metallic ions may be explained by neoclassical inward drifts if we assume that the anomalous diffusion is sufficiently suppressed, some indication of this being found from laser blow-off studies. The present results, however, can only be partly explained by neoclassical theory, according to which accumulation of low-Z impurities should also occur. The temporal behaviour of accumulation and the retarding effect of proton dilution for collision dominated transport are also discussed. Finally, we conclude that the full benefits of improved energy confinement can be achieved only if the impurity influxes are kept to a sufficiently low level. Expressed in terms of concentrations under low confinement conditions we have to postulate, for ASDEX, concentrations ≲ 10−4 for metals and ≲ 2% for all light impurities.

A toroidal large area limiter was installed in the divertor tokamak ASDEX. The limiter basically consists of a plane annular stainless steel sheet which is tangential to the plasma and which is placed in the bottom of the main plasma chamber. In a second series of experiments this limiter was covered with graphite plates. With the steel limiter clean discharges could not be obtained. Energy deposition on the limiter was only 12% of the input. With the graphite limiter Z eff was below 3, energy deposition on the limiter was 25% (up to 40% in low density operation). Energy deposition was toroidally symmetrical. Results, including some experiments with a single null divertor are presented.

Discharges fuelled by stationary pellet injection (PI), gas puffing (GP) or a combination of the two methods are compared with respect to recycling in the divertor and particle confinement. Fuelling by PI yields much better global particle confinement than by GP. This has been found for both low and high recycling. In the low-recycling case this improvement is due to the deeper particle deposition for PI than for GP since the transport in the inner plasma is not reduced. For high recycling the improvement results from both the deeper deposition and a reduction in the transport. The best global particle confinement was found for phases with low or no GP. This, however, can be reached for short times only. Since with PI alone it is impossible to keep the recycling on a high level, GP is unavoidable for sustaining the favourable high-recycling condition.

Neutral particle fluxes onto the walls of ASDEX have been investigated using a time-of-flight (TOF) method. The energy distributions of the neutrals could be determined in the range of 10–1000 eV/amu. Ohmic divertor and limiter discharges with equal plasma currents and densities have been compared for He and D2. The He0 outflux at ∼2000 eV from He discharges is 110 of the corresponding D0 flux in D2 discharges. At lower energies this difference is much smaller. In all cases many more He neutrals were observed than was anticipated from the CX rate-coefficients for He2+. The impurity fluxes due to sputtering by the CX-neutrals show no significant difference for He and D2 discharges. For divertor discharges CX-sputtering can fully account for the Fe impurity content determined spectroscopically.

The fueling efficiency for gas puffing, i.e. the fraction of the external gas flux that is ionized inside the separatrix, is reduced in divertor discharges since part of it is ionized in the scrape-off layer and pumped off by the divertor. The fueling efficiency is determined by switching-off the gas feed during the stationary phase of a discharge and dividing the time derivative of the total number of particles inside the separatrix by the external gas flux. The determination of this time derivative must take into account profile changes. In ASDEX the fueling efficiency ranges from close to 1.0 for discharges with a stainless steel poloidal limiter and decreases to about 0.2 at high densities ( 6 × 10 13 cm −3 line average) for diverted discharges. These results are compared with estimates of the fueling efficiency which include molecular disintegration, plasma albedo for neutral atoms and imperfect wall reflection.

In the first experimental phase the divertor tokamak ASDEX was run with a closed SS-limiter. The cleaning procedure for the vessel consisted of baking to 120°C, carbon removal by glow discharge in hydrogen, and oxygen removal with a continuous low power 50 Hz AC discharge. Wall contact of the plasma was reduced by carefully positioning the plasma with a feedback system. Discharges with plasma currents up to 280 kA and a disruption-free duration of up to 1 s were reliably produced with a filling pressure of 5 × 10 −5 mbar and a programmed current rise. No change in start-up conditions and discharge behaviour was observed in material limiter discharges with superimposed divertor field.