SwePub - sökning: WFRF:(Pospieszczyk A.)

Numrering	Referens	Omslagsbild	Hitta
1.	Overview of the JET results 2015 Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 55:10 Tidskriftsartikel (refereegranskat)
2.	Abel, I, et al. (författare) Overview of the JET results with the ITER-like wall 2013 Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 53:10, s. 104002- Tidskriftsartikel (refereegranskat)abstract Following the completion in May 2011 of the shutdown for the installation of the beryllium wall and the tungsten divertor, the first set of JET campaigns have addressed the investigation of the retention properties and the development of operational scenarios with the new plasma-facing materials. The large reduction in the carbon content (more than a factor ten) led to a much lower Z(eff) (1.2-1.4) during L- and H-mode plasmas, and radiation during the burn-through phase of the plasma initiation with the consequence that breakdown failures are almost absent. Gas balance experiments have shown that the fuel retention rate with the new wall is substantially reduced with respect to the C wall. The re-establishment of the baseline H-mode and hybrid scenarios compatible with the new wall has required an optimization of the control of metallic impurity sources and heat loads. Stable type-I ELMy H-mode regimes with H-98,H-y2 close to 1 and beta(N) similar to 1.6 have been achieved using gas injection. ELM frequency is a key factor for the control of the metallic impurity accumulation. Pedestal temperatures tend to be lower with the new wall, leading to reduced confinement, but nitrogen seeding restores high pedestal temperatures and confinement. Compared with the carbon wall, major disruptions with the new wall show a lower radiated power and a slower current quench. The higher heat loads on Be wall plasma-facing components due to lower radiation made the routine use of massive gas injection for disruption mitigation essential.
3.	Romanelli, F, et al. (författare) Overview of the JET results 2011 Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 51:9 Tidskriftsartikel (refereegranskat)abstract Since the last IAEA Conference JET has been in operation for one year with a programmatic focus on the qualification of ITER operating scenarios, the consolidation of ITER design choices and preparation for plasma operation with the ITER-like wall presently being installed in JET. Good progress has been achieved, including stationary ELMy H-mode operation at 4.5 MA. The high confinement hybrid scenario has been extended to high triangularity, lower ρand to pulse lengths comparable to the resistive time. The steady-state scenario has also been extended to lower ρand ν*and optimized to simultaneously achieve, under stationary conditions, ITER-like values of all other relevant normalized parameters. A dedicated helium campaign has allowed key aspects of plasma control and H-mode operation for the ITER non-activated phase to be evaluated. Effective sawtooth control by fast ions has been demonstrated with3He minority ICRH, a scenario with negligible minority current drive. Edge localized mode (ELM) control studies using external n = 1 and n = 2 perturbation fields have found a resonance effect in ELM frequency for specific q95values. Complete ELM suppression has, however, not been observed, even with an edge Chirikov parameter larger than 1. Pellet ELM pacing has been demonstrated and the minimum pellet size needed to trigger an ELM has been estimated. For both natural and mitigated ELMs a broadening of the divertor ELM-wetted area with increasing ELM size has been found. In disruption studies with massive gas injection up to 50% of the thermal energy could be radiated before, and 20% during, the thermal quench. Halo currents could be reduced by 60% and, using argon/deuterium and neon/deuterium gas mixtures, runaway electron generation could be avoided. Most objectives of the ITER-like ICRH antenna have been demonstrated; matching with closely packed straps, ELM resilience, scattering matrix arc detection and operation at high power density (6.2 MW m-2) and antenna strap voltages (42 kV). Coupling measurements are in very good agreement with TOPICA modelling. © 2011 IAEA, Vienna.
4.	Litnovsky, A., et al. (författare) Overview of material migration and mixing, fuel retention and cleaning of ITER-like castellated structures in TEXTOR 2011 Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 415:1, s. S289-S292 Tidskriftsartikel (refereegranskat)abstract 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.
5.	Rubel, Marek, et al. (författare) Overview of wall probes for erosion and deposition studies in the TEXTOR tokamak 2017 Ingår i: Matter and Radiation at Extremes. - : Elsevier B.V.. - 2468-2047 .- 2468-080X. ; 2:3, s. 87-104 Tidskriftsartikel (refereegranskat)abstract 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.
6.	Sergienko, G., et al. (författare) Experience with bulk tungsten test-limiters under high heat loads : melting and melt layer propagation 2007 Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T128, s. 81-86 Tidskriftsartikel (refereegranskat)abstract The paper provides an overview of processes and underlying physics governing tungsten melt erosion in the fusion plasma environment. Experiments with three different bulk tungsten test-limiters were performed in TEXTOR: (i) thermally insulated solid plate fixed on a graphite roof-like limiter heated up by the plasma to the melting point, (ii) macro-brush of the ITER-relevant castellated structure and (iii) lamellae structure developed for the JET divertor. The main objectives were to determine the metal surface damage, the formation of the melt layer and its motion in the magnetic field. PHEMOBRID-3D and MEMOS-1.5D numerical codes were used to simulate the experiment with the roof-like test-limiter. Both experiments and simulation showed that the melting of tungsten can lead to a large material redistribution due to thermo-electron emission currents without ejection of molten material to the plasma.
7.	Weckmann, Armin, et al. (författare) Review on global migration, fuel retention and modelling after TEXTOR decommission 2018 Ingår i: NUCLEAR MATERIALS AND ENERGY. - : ELSEVIER SCIENCE BV. - 2352-1791. ; 17, s. 83-112 Forskningsöversikt (refereegranskat)abstract 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.
8.	Huber, A., et al. (författare) Comparison of impurity production, recycling and power deposition on carbon and tungsten limiters in TEXTOR-94 2001 Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 290, s. 276-280 Tidskriftsartikel (refereegranskat)abstract Impurity production, hydrogen recycling and power deposition on carbon and tungsten limiters have been investigated in TEXTOR-94 using a C-W twin test limiter. Considerable differences have been observed on W and C surfaces, which can be explained by the different particle and energy reflection coefficients of hydrogen on these surfaces. The measurements show in addition that the majority of the carbon release is from recycled carbon and that only a small part (below 10%) is due to net-erosion from the bulk carbon material. The heat deposition on C and W sides differs under the same plasma conditions significantly and is typically about 30% larger on the cal bon surface. The behaviour of the impurity production: recycling and power deposition for various discharge conditions is presented.
9.	Laux, M., et al. (författare) Arcing at B4C-covered limiters exposed to a SOL-plasma 2003 Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 313, s. 62-66 Tidskriftsartikel (refereegranskat)abstract 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.
10.	Litnovsky, A., et al. (författare) Dust investigations in TEXTOR : Impact of dust on plasma-wall interactions and on plasma performance 2013 Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 438:Suppl., s. S126-S132 Tidskriftsartikel (refereegranskat)abstract Dust will have severe impact on ITER performance since the accumulation of tritium in dust represents a safety issue, a possible reaction of dust with air and steam imposes an explosion hazard and the penetration of dust in core plasmas may degrade plasma performance by increasing radiative losses. Investigations were performed in TEXTOR where known amounts of pre-characterized carbon, diamond and tungsten dust were mobilized into plasmas using special dust holders. Mobilization of dust changed a balance between plasma-surface interactions processes, significantly increasing net deposition. Immediately after launch dust was dominating both core and edge plasma parameters. Remarkably, in about 100 ms after the launch, the effect of dust on edge and core plasma parameters was vanished: no increase of carbon and tungsten concentrations in the core plasmas was detected suggesting a prompt transport of dust to the nearby plasma-facing components without further residence in the plasma.
11.	Ohya, K., et al. (författare) Modeling of erosion and deposition patterns on C-W and W-Ta twin limiters exposed to the TEXTOR edge plasmas 2004 Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 329-33, s. 732-736 Tidskriftsartikel (refereegranskat)abstract The erosion and deposition patterns on tungsten and tantalum test limiters exposed to the TEXTOR deuterium plasma containing a small amount of C impurity are simulated with the modified EDDY code. At the very top of the W and Ta limiters, there occurs neither erosion nor deposition, but the erosion proceeds slowly along the surface. When approaching the edge, the surface is covered by a thick C layer, which shows a very sharp boundary similar to the observation in surface measurements. In the erosion zone, the re-deposited carbon forms a W (Ta)-C mixed layer with small C concentration. Assumptions for chemical erosion yields of similar to0.01 for W and
12.	Pospieszczyk, A., et al. (författare) B4C-limiter experiments at TEXTOR 2003 Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 313, s. 223-229 Tidskriftsartikel (refereegranskat)abstract In the TEXTOR tokamak the five top and five bottom poloidal carbon limiter blocks have been replaced by inertially cooled copper blocks coated with a 170 mum VPS-B4C layer. Similar limiter blocks have been inserted through lock systems, extensively diagnosed in situ as well as ex situ. During the thermal load by the plasma, the surface temperature rose and decayed extremely fast which can be explained by a different thermal conductivity and heat capacity of the coating. For heat loads below 8 MW m(-2) no severe cracking or delamination of the B4C-coating were observed. Due to the insulating behaviour of the layer, distinct craters developed on both limiter types, which reached down to the copper surface and are assumed to be caused by electrical arcs. An oscillation of the evolution of the surface temperature has been observed under certain conditions, which is clearly correlated to the use of the coated test limiter. Particle fluxes as well as hydrogen inventory turned out to be very similar to those from a low-Z surface in a carbon surrounding. No significant impact of the plasma on the coating and vice versa was observed.
13.	Pospieszczyk, A., et al. (författare) Operation of TEXTOR-94 with tungsten poloidal main limiters 2001 Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 290, s. 947-952 Tidskriftsartikel (refereegranskat)abstract In TEXTOR-93, experiments have been performed with the upper and lower poloidal limiter blocks made of vapour sprayed (VSP) tungsten (about 0.5 mm) deposited on graphite with a rhenium interlayer. A series of discharge conditions have been performed (density scan, scan of the auxiliary heating power, radius scan). There has been found no restriction for operation at any density with auxiliary heating. For Ohmic conditions the same density with testlimiters could be reached. Under siliconized conditions no severe accumulation of tungsten in the plasma centre could be detected. The blocks could in general stand surface temperatures below 1700 K. Most of them survived also temperatures above 3000 K without exfoliation. However, some blocks showed severe damage by melting or exfoliation probably due to insufficient contact of the tungsten layer with the graphite.
14.	Sergienko, G., et al. (författare) High temperature erosion of tungsten exposed to the TEXTOR edge plasma 2005 Ingår i: 32nd EPS Conference on Plasma Physics 2005, EPS 2005, Held with the 8th International Workshop on Fast Ignition of Fusion Targets. - 9781622763320 ; , s. 373-376 Konferensbidrag (refereegranskat)
15.	Weckmann, Armin, et al. (författare) Local Migration Studies of High-Z Metals in the TEXTOR Tokamak 2016 Ingår i: Physica Scripta. - : Institute of Physics Publishing (IOPP). - 0031-8949 .- 1402-4896. ; T167 Tidskriftsartikel (refereegranskat)abstract 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.
16.	Weckmann, Armin, et al. (författare) Whole-machine material migration studies in the TEXTOR tokamak with molybdenum 2017 Ingår i: NUCLEAR MATERIALS AND ENERGY. - : Elsevier. - 2352-1791. ; 12, s. 518-523 Tidskriftsartikel (refereegranskat)abstract 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.
17.	Hirai, T., et al. (författare) Performance and erosion of a tungsten brush limiter exposed at the TEXTOR tokamak 2003 Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 313, s. 67-71 Tidskriftsartikel (refereegranskat)abstract To examine the performance of a castellated structure under plasma loading, a hemispherical solid tungsten brush limiter was exposed to the plasma in the TEXTOR-94 tokamak. Due to the thermally isolated column of W segments, IR camera showed a non-uniform temperature distribution. The maximum incident power density was calculated to be about 35-40 MW/m(2). Concerning impurity generation, the structure did not show any particular effects. During plasma exposure, only some minor cracks developed in one of the columns, however, the crack propagation was interrupted by a groove. It can be concluded that the W brush limiter had comparable performance and superior mechanical behaviour compared to a solid W limiter. To study erosion and long-range transport of W atoms, a graphite limiter was exposed simultaneously with the brush limiter. As a result, the deposited W atoms via long-range transportation were estimated to be 10(15) cm(-2) shot(-1) at 46.5 cm from the plasma centre of TEXTOR.
18.	Hirai, T., et al. (författare) Testing of tungsten and tantalum limiters at the TEXTOR tokamak : Material performance and deuterium retention 2003 Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T103, s. 59-62 Tidskriftsartikel (refereegranskat)abstract 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.
19.	Huber, A., et al. (författare) In-situ measurement of trapped hydrogen by laser desorption in TEXTOR-94 2001 Ingår i: Physica scripta. T. - 0281-1847. ; T94, s. 102-105 Tidskriftsartikel (refereegranskat)abstract First measurements on laser induced desorption of deuterium incorporated in a boron layer formed by plasma chemical vapor deposition on tungsten and graphite limiter surfaces have been performed. A ruby laser (lambda = 694 nm) with maximum energy of 50 J and a pulse length of about 0.5 ms was used as a heat source. The desorbed deuterium was detected by mass spectroscopy and total pressure analysis in the residual gas. The amount of desorbed deuterium is about 10(17) D atoms cm(-2). The majority of the deuterium is released during the first laser pulse. The limiter heads were investigated post-mortem by means of ion beam analysis to determine the spatial distribution of boron and deuterium and to investigate the effect of the laser pulse on the release of deuterium and sublimation of boron in the laser spot. All the deuterium has been released by the laser pulse. The boron is sublimated partly from the graphite and removed nearly completely from the tungsten surface.
20.	Kantor, M., et al. (författare) Characterization of dust particles in the TEXTOR tokamak with Thomson scattering diagnostic 2013 Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 438:Suppl., s. S711-S714 Tidskriftsartikel (refereegranskat)abstract The presence of dust particles in a fusion plasma is recognized as a serious issue for safe and efficient operation of the ITER tokamak. The paper presents an in situ laser assisted method for characterization of dust from thermal emission of the particles. The method was developed in the TEXTOR tokamak with the use of Thomson scattering (TS). The diagnostic is capable to detect single particles and measure the dust density profile along the laser probing axis, velocity distribution of dust particles along this axis as well as surface temperature and size of the detected particles.
21.	Kreter, A., et al. (författare) Fuel retention in carbon materials under ITER-relevant mixed species plasma conditions 2009 Ingår i: Physica scripta. T. - : Institute of Physics Publishing (IOPP). - 0281-1847 .- 0031-8949 .- 1402-4896. ; T138, s. 014012- Tidskriftsartikel (refereegranskat)abstract Samples of CFC NB41 and fine-grain graphite ATJ have been exposed to PISCES plasmas containing (i) pure deuterium, (ii) deuterium and beryllium, (iii) deuterium, beryllium and helium, and (iv) deuterium, beryllium and argon. Thermal desorption spectrometry (TDS) and nuclear reaction analysis (NRA) have been used to measure the amount and distribution of retained deuterium in the samples. For the case of pure deuterium plasma, parametric studies of deuterium retention in NB41 have been done with variations of the incident deuterium fluence (Phi = 1 x 10(25)-5 x 10(26) m(-2)), ion energy (E-i = 20-120 eV) and sample surface temperature (T-s = 370-820 K). It has been found, that for T-s = 470K the retention scales as Phi(0.35). For T-s = 820K the retention saturates at a level of similar to 10(21) Dm(-2). The retention increases with E-i and drops with higher T-s. At T-s = 720 K, the beryllium seeding results in a building of a protective beryllium carbide layer, which appears to prevent the in-bulk diffusion of deuterium, thus reducing the retention. Admixture of Ar and, in the case of low E-i, He leads to a significant reduction of the retention.
22.	Ohya, K., et al. (författare) Simulation calculations of mutual contamination between tungsten and carbon and its impact on plasma surface interactions 2001 Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 290, s. 303-307 Tidskriftsartikel (refereegranskat)abstract Mutual contamination between C and W, resulting from the simultaneous use of these materials as plasma facing components, is simulated by means of a computer simulation code, Erosion and Deposition based on Dynamic model (EDDY). W deposition on C rapidly increases the reflection coefficient for D and C impurity. In comparison between the calculation and a C-W twin test limiter experiment in TEXTOR-94, C release from the C side of the limiter is dominated by reflection of C impurity from the W deposits, in addition to physical sputtering of C; chemical erosion is strongly suppressed. Due to the dynamic effect which makes C-W mixed layer, C deposition on W gradually changes the reflection coefficient and sputter yields. Formation of a sharp boundary between erosion and C deposition zones on the W side of the limiter is well reproduced by simulation. Local redeposition patterns of C and W on the limiter surface are also calculated.
23.	Ohya, K., et al. (författare) Simulation study of carbon and tungsten deposition on W/C twin test limiter in TEXTOR-94 2000 Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 283, s. 1182-1186 Tidskriftsartikel (refereegranskat)abstract In order to investigate the impurity release and surface modification on a W/C twin test limiter, made of a half of W and the other half of C, exposed to the edge plasma of TEXTOR-94, simulation calculations of ion-surface interaction are conducted by a Monte Carlo code. According to the calculations, experimentally observed spatial distributions of WI and CII line intensities around the W side of the limiter can be explained by physical sputtering of W, reflection of bombarding C ions and physical sputtering of implanted C. The CII line emission, resulting from thermal C atoms, around the C side of the limiter is suppressed by deposition of W, and the reflection of C ions from W deposited on C causes the CII intensity to decay more slowly than that from C without the deposition. Bombardment with deuterium edge plasmas, containing impurity W, produces a thick W layer on the C side of the limiter, whereas C implanted in the W side is strongly sputtered due to impact of most constituent D ions.
24.	Rubel, Marek, et al. (författare) Fuel deposition and material mixing in a castellated tungsten limiter 2007 Ingår i: 34th EPS Conference on Plasma Physics 2007, EPS 2007 - Europhysics Conference Abstracts. - 9781622763344 ; , s. 303-306 Konferensbidrag (refereegranskat)
25.	Rubel, Marek J., et al. (författare) An overview of fuel retention and morphology in a castellated tungsten limiter 2008 Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 83:7-9, s. 1049-1053 Tidskriftsartikel (refereegranskat)abstract A castellated tungsten test limiter composed of detachable segments was exposed to plasma discharges in the TEXTOR to kamak operated with graphite main limiters. Dismantling of the limiter enabled the analysis Of Surfaces located inside the castellation, The emphasis was on the determination of: (i) deposition and fuel retention; (ii) material mixing and new Compound formation on plasma-facing Surfaces and in the grooves of castellation. The investigation performed by means of accelerator-based ion beam analysis methods, microscopy and X-ray diffraction has brought several essential results: (i) deuterium retention oil plasma-facing Surfaces and in the castellation of metal PFC is strongly related to the co-deposition with carbon; (ii) both carbon and deuterium are detected only in narrow belts, a few millimetre broad, clown the gap with the decay length of around 1.2-1.8 mm; (iii) the presence of copper droplets and tungsten oxide (WO(2)) has been identified in the gaps. Different pathways leading to the oxide formation are considered.

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