•13CD4 co-deposits from EAST D- and He-plasma tracer injection.•1.4 MeV deuteron ion beam analysis in ≈ 3000 points on about 8500 mm2.•24% (D plasma) and 16% (He plasma) deposition efficiency.•98% of 13C found within 20 mm around the injection hole.•D-NRA yields additional physics via D/C ratios, ExB correlation and fall-off length.

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 This review incorporates both published and unpublished results from this TEXTOR study and combines findings with analytical methods as well as modelling results from two codes, ERO and ASCOT.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 radial electric field and source form have a high impact on global deposition patterns, while toroidal flow has little influence. Some of the experimental findings could be reproduced. Still, qualitative differences between simulated and experimental global deposition patterns remain. The review closes with lessons learnt during this extensive TEXTOR study which might be helpful for future scientific exploitation of other tokamaks to be decommissioned.

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.

As the last experiment before the final shutdown of the TEXTOR tokamak, equipped with all-graphite plasma-facing components (PFCs), MoF6 had been injected into the vacuum vessel. During decommissio ...

Tracer injection experiments in TEXTOR with MoF6 and WF6 lead to local deposition of about 6% for Mo and about 1% for W relative to the injected amount of Mo and W atoms. Modelling of these experiments has been done with ERO applying updated data for physical sputtering. The dissociation of the injected molecules has been treated in a simplified manner due to the lack of dissociation rate coefficients. However, with this it was possible to reproduce the observed radial penetration of Mo and W atoms into the plasma. The modelled local deposition efficiencies are about 50% for Mo and 60% for W assuming typical plasma parameters for the experimental conditions used. To reproduce the measured deposition efficiencies an enhancement factor for the erosion of deposited Mo and W has to be assumed (similar to 10 for Mo and similar to 25 for W). Due to the rather low electron temperature T e of these plasma conditions (T-e similar to 15 eV at the location of injection), Mo and W are mostly sputtered by impurities whereas sputtering due to deuterium is negligible. A parameter study applying larger electron temperature leads to increased sputtering and thus to reduced local deposition efficiencies of about 30% for Mo and 5% for W. Though, even under these conditions enhanced erosion, albeit with reduced enhancement factors, is needed in the modelling to obtain the small measured deposition efficiencies. (C) 2016 The Authors. Published by Elsevier Ltd.

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 to major shut-downs, while the MoF6 injection was followed by the final shutdown in connection with TEXTOR decommissioning. In all cases a set of surface probes and limiter tiles was 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 show that only near the gas inlet significant are amounts of Mo deposited along the scrape-off layer flow and E. x. B drift directions, which could be reproduced by ERO-code modelling. Correlation between the injection scenario and the deposition patterns is presented.

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. (C) 2016 Published by Elsevier Ltd.

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. (C) 2014 Elsevier B.V. All rights reserved.

In optical diagnostic systems of ITER, mirrors will be used to guide the light from plasma towards detectors and cameras. The mirrors will be subjected to erosion due to fast particles and to deposition of impurities from the plasma which will affect adversely the mirror reflectivity and therefore must be suppressed or mitigated at the maximum possible extent. Predictive modeling envisages the successful suppression of deposition in the diagnostic ducts with fins trapping the impurities on their way towards mirrors located in the end of these ducts. To validate modeling predictions, cylindrical and cone-shaped diagnostic ducts were exposed in TEXTOR for 3960 s of plasma operation. After exposure, no drastic suppression of deposition was observed in the cylindrical ducts with fins. At the same time, no detectable deposition was found on the mirrors located at the end of cone-shaped ducts outlining the advantages of the cone geometry. Analyses of exposure provide evidence that the contamination of exposed mirrors was due to wall conditioning discharges and not due to working plasma exposure. Cleaning by plasma sputtering was performed on molybdenum mirrors pre-coated with a 100 nm thick aluminum film. Aluminum was used as a proxy of beryllium. During exposure in electron cyclotron resonance-generated helium plasma, the entire coating was sputtered within nine hours, leaving no trace of aluminum and leading to the full recovery of the specular reflectivity without detrimental effects on the mirror surface.

Resonant Magnetic Perturbations (RMPs) are applied with the Dynamic Ergodic Divertor (DED) at TEXTOR to control the plasma edge transport and the plasma surface interaction. This leads to the formation of a three-dimensional (3D) topology of the scrape-off layer (SOL). To quantify the erosion/deposition balance and the material migration in this 3D boundary, spherical test limiters were exposed to plasmas with and without RMP fields applied. Methane doped with 13C as tracer element was injected through a gas inlet in the test limiter. The local gas source was monitored by spatially resolving spectroscopy and the resulting deposition patterns on the limiters were analysed with colourimetry and nuclear reaction analysis. These measurements were compared to simulations of the magnetic field topology simulations. The data provide evidence of a particle migration dominated by an ExB drift within stochastic zones of the 3D plasma boundary.

The 1.1–1.5 mm wide gaps between tiles of the main toroidal belt limiter in TEXTOR were utilized to study the long-term impurity deposition and fuel retention in gaps. The tiles were exposed during a full tokamak campaign of 9365 s of plasma to various discharge conditions and wall conditioning, accumulating of up to 30 μm thick layers at the gap entrance. It was found that (i) gaps trap impurities twice as efficient as the top surface, (ii) the deposition in the toroidal gaps is twice as high as in the poloidal, (iii) carbon deposition decays with a fall-off length of about 0.7 mm towards the gap bottom, (iv) deposition on the bottom is significantly higher than on the adjacent side walls of gaps, and (v) the amount of deuterium scales with the amount of carbon with D/C varying from 3% to 30% depending on the surface temperature.

Results of a new dedicated experiment addressing the problem of impurity deposition at the bottom in gaps are presented along with modelling. A test limiter with an isolated gap was exposed to the scrape-off layer plasma in TEXTOR. The exposure was accompanied by injection of C-13-marked methane in the vicinity of the gap. Deposition at the bottom of the gap was monitored in situ with Quartz-Microbalance diagnostics. The C-13 deposition efficiency of about 2.6 x 10(-5) was measured. Post mortem analysis of resulting deposited layers performed with SIMS and EPMA techniques yields about a factor 2 smaller value corresponding to approximately 10% contribution of the gap bottom to the total C-13 deposition in the gap. This measured contribution is effectively much smaller than observed earlier in TEXTOR, taking the difference in geometry into account, and is in reasonable agreement with modelling performed with ERO and 3D-GAPS codes. (C) 2013 Elsevier B. V. All rights reserved.

In ITER, mirrors will be used as plasma-viewing elements in all optical and laser diagnostics. In the harsh environment mirror performance will degrade hampering the operation of associated diagnostics. The most adverse effect on mirror performance is caused by the deposition of impurities. It is expected that the most challenging situation will occur in the divertor. With the envisaged changes to all-metal plasma-facing components (PFCs) in ITER, an assessment of mirror performance in an existing divertor tokamak with allmetal PFCs is urgently needed. Molybdenum and copper mirrors were exposed for nearly nine months in ASDEX Upgrade which has alltungsten PFCs. Mirrors were located at the inner wall, under the dome and in the pump-duct. During exposure, the mirrors were heated to temperature in the range 145C-165C. This was made to approach the expected level of heating due to absorption of neutrons and gammas on mirrors in ITER divertor. After exposure, degradation of the reflectivity was detected on all mirrors. The highest reflectivity drop was measured on mirrors under the dome facing the outer strike point, reaching -55% at 500 nm. The least degradation was detected on mirrors in the pump duct, where the reflectivity was preserved in the range 500-2500 nm and the largest decrease was about -8% at 250 nm. On all contaminated mirrors carbon fraction did not exceed 50 at.% while the major contaminants were metals and oxygen. The degradation of exposed mirrors underlines the necessity for urgent R&D on deposition mitigation and in-situ mirror cleaning in ITER.

In ITER, mirrors will be used as plasma-viewing elements in all optical and laser diagnostics. In the harsh environment, mirror performance will degrade hampering the operation of associated diagnostics. The most adverse effect on mirror performance is caused by the deposition of impurities. It is expected that the most challenging situation will occur in the divertor. With the envisaged changes to all-metal plasma-facing components (PFCs) in ITER, an assessment of mirror performance in an existing divertor tokamak with all-metal PFCs is urgently needed.Molybdenum and copper mirrors were exposed for nearly nine months in ASDEX Upgrade which has all-tungsten PFCs. Mirrors were located at the inner wall, under the dome and in the pump duct. During exposure, the mirrors were heated to temperature in the range 145-165 degrees C. This was made to approach the expected level of heating due to absorption of neutrons and gammas on mirrors in the ITER divertor. After exposure, degradation of the reflectivity was detected on all mirrors. The highest reflectivity drop was measured on mirrors under the dome facing the outer strike point, reaching -55% at 500 nm. The least degradation was detected on mirrors in the pump duct, where the reflectivity was preserved in the range 500-2500 nm and the largest decrease was about -8% at 250 nm. On all contaminated mirrors carbon fraction did not exceed 50 at% while the major contaminants were metals and oxygen. The degradation of exposed mirrors underlines the necessity for urgent R&D on deposition mitigation and in situ mirror cleaning in ITER.

Tracer experiments have been carried out by injection C-13 marked methane through test limiters exposed to the scrape-off-layer in TEXTOR. The influence of impact energy and flux on depositing C-13 species has been studied. One experiment has been performed with biased test limiter (-300 V) in order to increase energy of positively charged ions and the other one with 10 times reduced (CH4)-C-13 injection rate compared to previously used injection rate. Biasing of the test limiter increases the resulting C-13 deposition by a factor of similar to 6 - post-mortem analysis yields a C-13 deposition efficiency of similar to 1.7% compared to similar to 0.3% without biasing. Reducing the injection rate increases C-13 deposition efficiency to similar to 0.7%, which is more than two times larger compared to experiments with previously used injection rate. ERO modelling shows that enhanced re-erosion of redeposits is still necessary to reproduce measured C-13 deposition efficiencies. (C) 2013 Elsevier B. V. All rights reserved.

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 future fusion devices like ITER deposition of impurities will likely occur in areas, remote from plasma and on the sensitive components of optical diagnostics, like mirrors and windows. Deposition in remote areas may lead to the tritium retention and therefore represent a safety issue. Deposition on optical components will severely deteriorate their optical properties potentially leading to the shutdown of the respective diagnostic systems. An active control over deposition is therefore highly desirable. The paper contains an overview of experiments on active control over carbon deposition undertaken in TEXTOR tokamak. A prototype of diagnostic duct was exposed in the scrape-off layer plasmas of TEXTOR where several techniques were applied to mitigate carbon deposition on diagnostic mirrors located inside this prototype. The complete suppression of carbon deposition on the surface of diagnostic mirror was achieved by feeding the deuterium gas inside the diagnostic duct.

To analyse the impurity transport in plasma-shadowed, remote areas, methane CD4 has been injected into the far scrape-off layer of TEXTOR through a cylinder equipped with a quartz micro balance (QMB). CD4 transport including break-up and resulting deposition on the QMB (shot-resolved) and on the cylinder top surface (shot-integrated) has been modelled with the codes ERO and 3D-GAPS. The modelling shows good agreement with the observations if reflection coefficients based on molecular dynamics simulations are used. In contrast to plasma-wetted areas, no enhanced erosion has to be applied.

Deposition of impurities can drastically change optical properties of metallic mirrors used for diagnostics of fusion plasmas. Specifically in ITER, this can become critical for the reactor operation by affecting the quality and reliability of detected signals. A series of experiments was performed in TEXTOR where local gas feeding was used for mitigation of carbon deposition on the mirror surface. A prototype of a diagnostic duct with molybdenum mirrors equipped with a gas feeding system was exposed in scrape-off layer plasma under deposition-dominated conditions. In the case of helium feeding, suppression of deposition was observed, while feeding of deuterium not only demonstrated full suppression of deposition but even favored complete removal of initially pre-deposited a-C:D layers. This paper presents results of experiments along with dedicated modeling of plasma–gas interaction inside the diagnostic duct.