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Before decommissioning of the TEXTOR tokamak in 2013, the machine was conditioned with a comprehensive migration experiment where MoF6 and N-15(2) were injected on the very last operation day. Thereafter, all plasmafacing components (PFCs) were available for extensive studies of both local and global migration of impurities - Mo, W, Inconel alloy constituents, 15 N, F - and fuel retention studies. Measurements were performed on 140 limiter tiles out of 864 throughout the whole machine to map global transport. One fifth of the introduced molybdenum could be found. Wherever possible, the findings are compared to results obtained previously in other machines. This review incorporates both published and unpublished results from this TEXTOR study and combines findings with analytical methods as well as modelling results from two codes, ERO and ASCOT. The main findings are: Both local and global molybdenum transport can be explained by toroidal plasma flow and (sic) x (sic) drift. The suggested transport scheme for molybdenum holds also for other analysed species, namely tungsten from previous experiments and medium-Z metals (Cr-Cu) introduced on various occasions. Analytical interpretation of several deposition profile features is possible with basic geometrical and plasma physics considerations. These are deposition profiles on the collector probe, the lower part of the inner bumper limiter, the poloidal cross-section of the inner bumper limiter, and the poloidal limiter. Any deposition pattern found in this TEXTOR study, including fuel retention, has neither poloidal nor toroidal symmetry, which is often assumed when determining deposition profiles on global scale. Fuel retention is highly inhomogeneous due to local variation of plasma parameters - by auxiliary heating systems and impurity injection - and PFC temperature. Local modelling with ERO yields good qualitative agreement but too high local deposition efficiency. Global modelling with ASCOT shows that the radia, QC 20190115

Before decommissioning of the TEXTOR tokamak in 2013, the machine was conditioned with a comprehensive migration experiment where MoF6 and N-15(2) were injected on the very last operation day. Thereafter, all plasmafacing components (PFCs) were available for extensive studies of both local and global migration of impurities - Mo, W, Inconel alloy constituents, 15 N, F - and fuel retention studies. Measurements were performed on 140 limiter tiles out of 864 throughout the whole machine to map global transport. One fifth of the introduced molybdenum could be found. Wherever possible, the findings are compared to results obtained previously in other machines. This review incorporates both published and unpublished results from this TEXTOR study and combines findings with analytical methods as well as modelling results from two codes, ERO and ASCOT. The main findings are: Both local and global molybdenum transport can be explained by toroidal plasma flow and (sic) x (sic) drift. The suggested transport scheme for molybdenum holds also for other analysed species, namely tungsten from previous experiments and medium-Z metals (Cr-Cu) introduced on various occasions. Analytical interpretation of several deposition profile features is possible with basic geometrical and plasma physics considerations. These are deposition profiles on the collector probe, the lower part of the inner bumper limiter, the poloidal cross-section of the inner bumper limiter, and the poloidal limiter. Any deposition pattern found in this TEXTOR study, including fuel retention, has neither poloidal nor toroidal symmetry, which is often assumed when determining deposition profiles on global scale. Fuel retention is highly inhomogeneous due to local variation of plasma parameters - by auxiliary heating systems and impurity injection - and PFC temperature. Local modelling with ERO yields good qualitative agreement but too high local deposition efficiency. Global modelling with ASCOT shows that the radia, QC 20190115

An overview of diagnostic tools – test limiters and collector probes – used over the years for material migration studies in the TEXTOR tokamak is presented. Probe transfer systems are shown and their technical capabilities are described. This is accompanied by a brief presentation of selected results and conclusions from the research on material erosion – deposition processes including tests of candidate materials (e.g. W, Mo, carbon-based composites) for plasma-facing components in controlled fusion devices. The use of tracer techniques and methods for analysis of materials retrieved from the tokamak are summarized. The impact of research on the reactor wall technology is addressed., QC 20201002

MoF6 injection from a localised source into plasma edge in the TEXTOR tokamak was the last experiment before the final shut-down of the TEXTOR machine. During decommissioning all plasma-facing components (PFCs) became available for surface studies. Detailed mapping of Mo deposition was performed in order to determine its migration on global scale. The concentration of Mo on PFC decays exponentially with distance from the source. The decay length is of the order of 0.1 m on the main PFC and 1 m on the receded components. Also the decay lengths modelled with the ERO code are between 0.15-1.3 m, depending on the anomalous cross-field diffusion coefficient. The inner bumper limiter is found to be the major repository for Mo. Material balance measurements show that only up to 22% of the injected Mo was detected on all the PFCs thus indicating that a large fraction of injected Mo may have been pumped out before being deposited., QC 20180112

Volatile compounds of tungsten (WF6) and molybdenum (MoF6) were used as tracers of high-Z metal migration in the TEXTOR tokamak in several gas injection experiments when puffing was done through a test limiter. The experiments with W were performed prior major shut-downs, while the MoF6 was followed by the final shutdown in connection with TEXTOR decommissioning. In all cases a set of various surface probes and limiter tiles were retrieved and analysed with electron and ion beam techniques. The focus was on the local deposition in the vicinity of the gas inlet and in the inlet system. Depth profiles in the deposits and metal distribution maps clearly shown that only near the gas inlet significant amounts of Mo are deposited along the scrape-off layer flow and E×B drift directions, which could be reproduced by ERO-code modelling. Correlation between the plasma operation scenario and the deposition patterns is presented., QC 20170630

Tracer techniques were used in the TEXTOR tokamak to determine high-Z metal migration and the deposition of species used for plasma edge cooling or wall conditioning under different types of operation conditions. Volatile molybdenum hexa-fluoride, nitrogen-15 and oxygen-18 were used as markers in tokamak or ion cyclotron wall conditioning discharges (ICWC). The objective was to obtain qualitative and quantitative of a global and local deposition pattern and material mixing effects. The deposition and retention was studied on plasma-facing components, collector probes and test limiters. Optical spectroscopy and ex-situ analysis techniques were used to determine the plasma response to tracer injection and the modification of surface composition. Molybdenum and light isotopes were detected on all types of limiters and short-term probes retrieved from the vessel showing that both helium and nitrogen are trapped following wall conditioning and edge cooling. Only small amounts below 1 x 10(19) m(-2) of O-18 were detected on surfaces treated by oxygen-assisted ICWC., QC 20150901

Tracer techniques were used in the TEXTOR tokamak to determine high-Z metal migration and the deposition of species used for plasma edge cooling or wall conditioning under different types of operation conditions. Volatile molybdenum hexa-fluoride, nitrogen-15 and oxygen-18 were used as markers in tokamak or ion cyclotron wall conditioning discharges (ICWC). The objective was to obtain qualitative and quantitative of a global and local deposition pattern and material mixing effects. The deposition and retention was studied on plasma-facing components, collector probes and test limiters. Optical spectroscopy and ex-situ analysis techniques were used to determine the plasma response to tracer injection and the modification of surface composition. Molybdenum and light isotopes were detected on all types of limiters and short-term probes retrieved from the vessel showing that both helium and nitrogen are trapped following wall conditioning and edge cooling. Only small amounts below 1 x 10(19) m(-2) of O-18 were detected on surfaces treated by oxygen-assisted ICWC., QC 20150901

Plasma-facing components (PFCs) in ITER will be castellated by splitting them into small-size blocks to maintain the thermo-mechanical stability. However, there are concerns in particular on retention of codeposited radioactive fuel in the gaps. An R&D program is underway in TEXTOR addressing this acute issue of castellation. Material migration and fuel inventory are investigated using long- and short-term discharge-resolved experiments with castellated structures in TEXTOR. Significant impurity transport to the gaps was detected and results were in part quantitatively reproduced with 3D-GAPS code. Deposits containing up to 70 at.% of tungsten on the gap areas closest to the plasma were detected in recent experiments. Deposition in the gaps accompanied by metal mixing demand for development of effective cleaning techniques. In experiments with ITER-like castellation, the gaps were cleaned from carbonaceous deposits using oxygen plasmas at 350 degrees C. This contribution contains an overview of experimental and modeling results along with recommendations for PFCs in ITER., 19th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI). QC 20120227

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 similar to 1 g D/h in TEXTOR (L mode, limiter machine) up to similar 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., QC 20120227

Silicon samples were exposed in the scrape-off layer of the TEXTOR plasma using a fast reciprocating probe, with the aim of studying carbon deposition and deuterium retention during Dynamic Ergodic Divertor (DED) operation. Separate samples were exposed for 300 ms at the flat-top phase of neutral beam heated discharges. The exposure conditions were varied on a shot-to-shot basis by external magnetic perturbations generated by the DED in the m/n = 3/1, DC regime, base configuration. Nuclear Reaction Analysis (NRA) was used to characterise collector sample surfaces after their exposure. Enhanced concentrations of both carbon and deuterium (C 3-10 x 10(16) at./cm(2), D 8-60 x 10(15) at./cm(2)) were found. The D/C ratio was less than unity which indicates that most of the carbon and deuterium were co-deposited. Carbon e-folding lengths of about 2 cm were found on both toroidal sides of the probe independent of DED perturbations.

Silicon samples were exposed in the scrape-off layer of the TEXTOR plasma using a fast reciprocating probe, with the aim of studying carbon deposition and deuterium retention during Dynamic Ergodic Divertor (DED) operation. Separate samples were exposed for 300 ms at the flat-top phase of neutral beam heated discharges. The exposure conditions were varied on a shot-to-shot basis by external magnetic perturbations generated by the DED in the m/n = 3/1, DC regime, base configuration. Nuclear Reaction Analysis (NRA) was used to characterise collector sample surfaces after their exposure. Enhanced concentrations of both carbon and deuterium (C 3-10 x 10(16) at./cm(2), D 8-60 x 10(15) at./cm(2)) were found. The D/C ratio was less than unity which indicates that most of the carbon and deuterium were co-deposited. Carbon e-folding lengths of about 2 cm were found on both toroidal sides of the probe independent of DED perturbations., QC 20101118

Metal mirrors will be implemented in about half of the ITER diagnostics. Mirrors in ITER will have to withstand radiation loads, erosion by charge-exchange neutrals, deposition of impurities, particle implantation and neutron irradiation. It is believed that the optical properties of diagnostic mirrors will be primarily influenced by erosion and deposition. A solution is needed for optimal performance of mirrors in ITER throughout the entire lifetime of the machine. A multi-machine research on diagnostic mirrors is currently underway in fusion facilities at several institutions and laboratories worldwide. Among others, dedicated investigations of ITER-candidate mirror materials are ongoing in Tore-Supra, TEXTOR, DIII-D, TCV, T-10 and JET. Laboratory studies are underway at IPP Kharkov (Ukraine), Kurchatov Institute (Russia) and the University of Basel (Switzerland). An overview of current research on diagnostic mirrors along with an outlook on future investigations is the subject of this paper., QC 20100525

Samples of a limiter tile from the TEXTOR tokamak were investigated by scanning electron microscopy and nuclear reaction analysis both before and after laser heating. SEM images showed spheres and thin flakes covering the surface which are the areas modified by plasma particles striking under grazing angles. Due to roughness of the surface there are shadowed regions between the 'flakes'. Laser pulses did not lead to expected common ablation of the surface. Features that looked like 'melting' of thin surface layers were rather observed. The initial deuterium content in the surface layer of tiles was of the order of 10(18) D atoms per cm(2). After the laser light impact the content decreased with 60-70%; by reducing the deposited power by a factor four, the deuterium content was decreased by 40-50%. We make the interpretation that we approach a threshold of the laser detritiation method in fusion devices., QC 20100525 QC 20111012. 16th International Conference on Plasma Surface Interactions in Controlled Fusion Devices. Portland, ME. MAY 24-28, 2004

To insure the thermo-mechanical durability of ITER it is planned to manufacture the castellated armour of the divertor i.e. to split the armour into cells [W. Daener et al., Fusion Eng. Des. 61T62 (2002) 61]. This will cause an increase of the surface area and may lead to carbon deposition and tritium accumulation in the gaps in between cells. To investigate the processes of deposition and fuel accumulation in gaps, a castellated test-limiter was exposed to the SOL plasma of TEXTOR. The geometry of castellation used was the same as proposed for the vertical divertor target in ITER [W. Daener et al., Fusion Eng. Des. 61T62 (2002) 61]. After exposure the limiter was investigated with various surface diagnostic techniques. Deposited layers containing carbon, hydrogen, deuterium and boron were found both on top plasma-facing surfaces and in the gaps. The amount of deuterium in the gaps was at least 30% of that found on the top surfaces., QC 20100525 QC 20111012. 16th International Conference on Plasma Surface Interactions in Controlled Fusion Devices. Portland, ME. MAY 24-28, 2004

Large molybdenum mirrors have been exposed in the SOL of TEXTOR in order to simulate conditions relevant for ITER optical components. Distortions of the reflectivity - increase as well as decrease - are found in the erosion and deposition dominated areas, respectively. The changes are most pronounced in the near UV and level off in the IR and can partly be attributed to observed surface changes. A novel periscope system was installed and mirrors exposed in a pilot experiment to simulate the transmission of light to distant sensors in ITER., QC 20100525 QC 20111012. 16th International Conference on Plasma Surface Interactions in Controlled Fusion Devices. Portland, ME. MAY 24-28, 2004

Castellated limiter with ITER-like geometry was exposed under the erosion-dominated conditions in the SOL of TEXTOR. After the exposure, deposits were found in the gaps, with a negligible amount of deposited material that was detected on top surfaces open to the plasma. The thickest deposits are located on plasma shadowed sides of the gaps with maximum thickness up to 0.5 μm on the plasma closest locations as it was measured with SIMS diagnostic. The deposit thickness decrease exponentially with the depth of gaps with e-folding length of 1.7-2 mm as inferred from EPMA measurements. Material intermixing was found to occur in the deposited layers. Deposits are mixed Mo:C:D:O:B:H layers. Gaps contain at least (5÷10)×1016 of D fuel atoms per cm2. This corresponds roughly to 0.02÷0.04 % of total averaged D fluence impinging on the castellation., QC 20140819

First direct comparative test of single crystal and polycrystalline diagnostic mirror materials under erosion conditions has been made in TEXTOR. Before exposure in TEXTOR, glow discharge cleaning has efficiently restored the reflectivity of initially oxidized mirrors. After the exposure, no significant changes in total reflectivity were observed. Drastic increase of diffuse reflectivity was measured for polycrystalline molybdenum mirror, but not for the single crystal. Thus, specular reflectivity of single crystal is significantly higher than of polycrystalline one. The most affected wavelength range is 250-1000 nm, no significant changes of reflectivity was noticed in the range 1000-2000 nm. No or negligible effect of erosion on polarization characteristics of mirrors was measured., QC 20140825

C-13 labeled methane and rhenium-boron coated plates were used at the JET tokamak as tracers for studies of the material transport, its erosion and re-deposition. Experimental procedures are described. The results are discussed in terms of processes underlying the material transport and the change of morphology of targets exposed to the plasma: physical sputtering, chemical erosion, prompt re-deposition. The influence of wall materials on fuel inventory is also addressed. C-14 measurements in the TEXTOR tokamak are presented and possibilities of using C-14 in carbon migration studies are considered., QC 20101027 QC 20110926. Conference: 11th International Conference on Fusion Reactor Materials (ICFRM). Kyoto, JAPAN. DEC 07-12, 2003

Plasma facing components from JET and TEXTOR were studied. The emphasis was on the comparison of co-deposition, material mixing and fuel inventory on plasma facing and side surfaces of tiles, i.e. in gaps separating the tiles. Integrated fuel content in gaps of the Mk-I JET divertor floor was approximately two times greater than detected on the plasma facing surfaces. Taking into account similarities between the Mk-I structure and the castellation in the ITER divertor, the impact of the tile shaping on the tritium inventory is addressed. Deposition on the side of limiter tiles in TEXTOR was around 3-5% of that on the plasma facing surfaces. Experiments aiming at a deeper insight into the deposition on ITER-relevant components are also proposed., QC 20101119 QC 20111031

Supporting pilot experiments and activities which are currently considered or already performed for the development of the International Thermonuclear Experiment Reactor active beam spectroscopy diagnostic are addressed in this article. Four key issues are presented including optimization of spectral instrumentation, feasibility of a motional Stark effect (MSE) evaluation based on line ratios, "first-mirror" test-bed experiments at the tokamak TEXTOR, and finally the role of integrated data analysis for the conceptual layout of the change exchange recombination spectroscopy and MSE diagnostic., QC 20110916. 15th Topical Conference on High-Temperature Plasma Diagnostics. San Diego, CA. APR 19-22, 2004

A fast reciprocating probe system capable of transferring different types of heads has been constructed and implemented at the TEXTOR tokamak for diagnosing the plasma edge. It gives the possibility of using a particle collector technique to extend studies of material transport from the scrape-off layer to the near plasma edge. For the first time, the system was used for exposures of graphite samples (pure and coated with a-C:H or W) at positions both within and outside the last closed flux surface. Various surface analysis methods were applied to investigate the probe morphology and, by this, to determine radial deposition profiles of boron impurities and deuterium. The profiles for boron are remarkably flat whilst those for deuterium are characterised by a steep decay with the e-folding length of approximately 15 mm. On tungsten-coated samples almost no deuterium was found, most likely because of little carbon co-deposition, shallow implantation and low trapping coefficient of deuterons in the tungsten layer. Reconstruction of experimental results by means of a multifluid TECXY code helped to identify the contribution of impurity sources (limiters, wall) to the observed radial distribution of species., QC 20100525

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., QC 20100525

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 approximate to 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., QC 20100525

Application of ion beam analysis techniques in the studies of material transport and fuel inventory in the controlled fusion devices is exemplified. Enhanced proton scattering on the carbon isotopes C-12(p,p)C-12, C-13(p,p)C-13 and secondary ion mass spectrometry allowed for determination of carbon erosion and re-deposition on the wall components following the experiments with a tracer ((CH4)-C-13) injection into the plasma edge at the TEXTOR tokamak. For the assessment of the deuterium fuel accumulation in the plasma facing components depth profiling by means of nuclear reaction analysis, He-3(d,p)He-4, was performed. Advantages and limitations of those nuclear methods in solving experimental problems are addressed., QC 20100525

Recent results regarding the formation of co-deposits, fuel accumulation and overall material transport at the TEXTOR tokamak are described. Two categories of brittle flaking co-deposits were identified: (i) smooth stratified layers of a thickness of up to 50 mum and a fuel content of up to 16 at.%. (ii) granular and columnar structures reaching 1 mm in thickness and containing around 0.5 at.% of fuel species. They were formed on the blades of the toroidal belt pump limiter (similar to 15000 s of plasma operation) and on the neutraliser plates of this limiter (similar to 90000 s), respectively. A comparison is made to the fuel inventory measured in other controlled fusion devices with carbon walls., QC 20100525

Tungsten and tantaium were examined in the TEXTOR tokamak as test limiters in order to compare their performance under plasma operation and to recognise fuel recycling on endothermic (W) and exothermic (Ta) hydrogen absorbers. Differences have been noticed in the distribution and microstructure of co-deposits. in the fuel inventory in the bulk of metals and, in the deuterium release mechanism (ratio of molecules to atoms). As a result of poor thermal conductivity. the surface temperature of Ta during the power deposition was higher than that of W and it increased shot-by-shot because of the degradation of thermal properties due to surface modification. Results on thermal response, fuel recycling and inventory show that. as a candidate material for plasma facing components. tungsten is superior to tantalum., QC 20100525

Processes of erosion and redeposition and their impact on plasma facing materials in devices for magnetically confined fusion are discussed. Volatile molecules formed in the erosion process are partly pumped out but the majority of species released from the wall components returns to the surface causing the modification of its morphology. Prompt redeposition and redeposition after global transport reduce the gross erosion at any surface. Detailed analysis shows that erosion and deposition can coexist on otherwise erosion-dominated surfaces due to local inhomogenities. The erosion yield of redeposited material of sub-monolayer thickness significantly differs from that observed for thick targets. On deposition dominated areas one observes the formation of thick co-deposits containing a mixture of trapped fuel atoms (hydrogen isotopes) and species removed originally from the wall. This leads to a large and long-term fuel accumulation (tritium inventory) in a device. Reduced mechanical integrity of such layers stimulates their flaking and peeling-off if a critical thickness has been reached. This, in turn, results in the formation of hydrogen-rich dust particles., QC 20100525

The work is focused on the determination of the distribution and the total content of deuterium in co-deposits formed in the TEXTOR tokamak on a toroidal belt limiter which is the machine's major plasma facing component (PFC). Limiter tiles in use for 14 100 s of plasma operation were dismounted for examination with surface analysis and microscopy methods. Mapping of the deuterium distribution by means of nuclear reaction analysis (NRA) revealed the presence of deposition zones covering about 35% of the tiles'surface area. Besides C and D, other constituents of the layers were boron, silicon and inconel components. The co-deposit, with a stratified structure and a thickness of up to 50 mum, could be detached from the tiles. Deuterium depth profiling on both sides of the detached flakes and in the underlying graphite substrate enabled the D content in the deposition zones to be estimated at a level of 3.5 x 10(19) cm(-2). Adding the fuel content found in the erosion zone (3-7 x 10(17) cm(-2)) and on the back side of the tile (0.9-1.8 x 10(17) cm(-2)), the total amount of D atoms trapped in all the limiter tiles was assessed to be about 2 x 10(23) atoms. D content in the co-deposits accounted for approximately 10 at.% (C-D/C-C similar to 0.1) which was considered to be low in comparison to much greater values observed in other devices. The results of the ion beam analyses (IBA) agree well with the determination by thermal desorption spectrometry (TDS)., QC 20100525

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., QC 20100525

(CH4)-C-13 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 C-13 and of the C-12 out of the background. Ratios up to C-13/C approximate to 0.42 were found although the deposition efficiency for C-13 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 CHx+. Within the interaction depth the C-13/C ratio seems to become quasi-stationary. C-13 transported to obstacles creates a ratio of C-13/C approximate to 0.27 after 108 s, QC 20100525

Carbon erosion and deposition were investigated on the surface of a flat target covered with an a-C:H film and exposed for 197 s in the SOL of TEXTOR-94. The target was declined by 20 degrees with respect to the toroidal direction and partly protected by an aluminum (3 mm) plate which created an 8 mm wide local shadow. Thickness changes were measured by colorimetry after each plasma discharge. Carbon is eroded from surface areas near the plasma edge (LCFS +1 cm) and transported into the local shadow regions. Accumulation rates up to approximate to 7nm/s were found. The erosion in the local shadow regions (about - 0.1 nm/s) is due to charge exchange neutrals. The observations are confirmed by ion beam analyses and by preliminary calculations with the B2-EIRENE and ERO-TEXTOR code., QC 20100525

Graphite-tungsten twin limiters have been used at the TEXTOR tokamak for testing of high-Z metals as plasma facing materials and, in particular, for studies of the local and global transport of tungsten. The emphasis was on the change in surface morphology of limiters: the formation and properties of multicomponent co-deposits and the trapping characteristics of fuel on carbon and high-Z substrates exposed to the plasma under various operation conditions, i.e., heating scenarios, configuration of limiters, etc. Vast quantities of tungsten have been found to be locally transported to the adjacent graphite surfaces. Ion beam analysis also indicated strong intermixing of carbon, tungsten and boron on the hottest parts of the limiters. The results are discussed in terms of various mechanisms involving the transport of tungsten-containing species, possibilities of oxide production and formation of mixed (W-C-B) compounds., QC 20100525

This final report concerns a feasibility study of the increased frangibility of Microwave Landing System equipment mounted on the extended runway centerline and as assessment of the FAA requirements in regard to frangibility. As described herein, a conceptual design known as the inverted AZ array was developed and fabricated as a mechanical prototype. The primary structural components of the antenna are vertical cantilever waveguides and frangible joints incorporated into their base. Impact and static load tests proved that several joint designs meet the FAA requirements for frangibility. (Author)