| 1. |
-
Overview of the JET results
- 2015
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 55:10
-
Tidskriftsartikel (refereegranskat)
|
|
| 2. |
- 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 ρ*and to pulse lengths comparable to the resistive time. The steady-state scenario has also been extended to lower ρ*and 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.
|
|
| 3. |
- 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.
|
|
| 4. |
|
|
| 5. |
- Emmoth, Birger, et al.
(författare)
-
In-situ measurements of carbon and deuterium deposition using the fast reciprocating probe in TEXTOR
- 2009
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 390-91, s. 179-182
-
Tidskriftsartikel (refereegranskat)abstract
- 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.
|
|
| 6. |
- 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.
|
|
| 7. |
- 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.
|
|
| 8. |
- Sergienko, G., et al.
(författare)
-
Erosion of a tungsten limiter under high heat flux in TEXTOR
- 2007
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 363, s. 96-100
-
Tidskriftsartikel (refereegranskat)abstract
- Erosion characteristics of a tungsten plate heated up in TEXTOR by the plasma load have been investigated at temperatures extending to the melting point. No enhancement of atomic release exceeding physical sputtering and normal thermal sublimation for temperatures below 3700 K was observed. The liquid tungsten moved fast along the plate in the direction perpendicular to the magnetic field lines. The motion is caused by the Lorentz force due to the thermoelectron current emitted from the hot tungsten surface. The motion of liquid tungsten caused a material loss of 2.85 g during two discharges. The material redistribution due to the melt layer motion is compared with a MEMOS-1.5D simulation.
|
|
| 9. |
- 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.
|
|
| 10. |
|
|
| 11. |
- Emmoth, Birger, et al.
(författare)
-
Particle collection at the plasma edge by a fast reciprocating probe at the TEXTOR tokamak
- 2003
-
Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 313, s. 729-733
-
Tidskriftsartikel (refereegranskat)abstract
- 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.
|
|
| 12. |
- Gasior, P., et al.
(författare)
-
Laser-induced removal of co-deposits from graphitic plasma-facing components : Characterization of irradiated surfaces and dust particles
- 2009
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 390-91, s. 585-588
-
Tidskriftsartikel (refereegranskat)abstract
- Laser-induced fuel desorption and ablation of co-deposited layers on limiter plates from the TEXTOR tokamak have been studied. Gas phase composition was monitored in situ, whereas the ex situ studies have been focused on the examination of irradiated surfaces and broad analysis of dust generated by ablation of co-deposits. The size of the dust grains is in the range of few nanometers to hundreds of micrometers. These are fuel-rich dust particles, as determined by nuclear reaction analysis. The presence of deuterium in dust indicates that not all fuel species are transferred to the gas phase during irradiation. This also suggests that photonic removal of fuel and the ablation of co-deposit from plasma-facing components may lead to the redistribution of fuel-containing dust to surrounding areas.
|
|
| 13. |
- 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.
|
|
| 14. |
- 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.
|
|
| 15. |
- Ivanova, Darya, et al.
(författare)
-
Dust Particles in Controlled Fusion Devices: Generation Mechanism and Analysis
- 2010
-
Konferensbidrag (refereegranskat)abstract
- Generation and in-vessel accumulation of carbon and metal dust are perceived to be serious safety andeconomy issues for a steady-state operation of a fusion reactor, e.g. ITER. This contribution provides acomprehensive account on: (a) properties of carbon and metal dust formed in the TEXTOR tokamak; (b) dustgeneration associated with removal of fuel and co-deposit from carbon PFC from TEXTOR and Tore Supra; (c)surface morphology of wall components after different cleaning treatments. The amount of loose dust found on thefloor of the TEXTOR liner does not exceed 2 grams with particle size range 0.1 m – 1 mm. The presence of fine(up to 1 m) crystalline graphite in the collected matter suggests that brittle destruction of carbon PFC could takeplace during off-normal events. Carbon is the main component, but there are also magnetic and non-magnetic metalagglomerates. The fuel content in dust and co-deposits varies from 10% on the main limiters to 0.03% on theneutralizer plates. Fuel removal by oxidative methods or by annealing in vacuum disintegrates co-deposits and, in thecase of thick layers, makes them brittle thus reducing the adherence to the target. Also photonic cleaning by laserpulses produces debris, especially under ablation conditions. The results obtained strongly indicate that in a carbonwall machine the disintegration of flaking co-deposits on PFC is the main source of dust.
|
|
| 16. |
- Ivanova, Darya, et al.
(författare)
-
Laser-based and thermal methods for fuel removal and cleaning of plasma-facing components
- 2011
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 415:1, s. S801-S804
-
Tidskriftsartikel (refereegranskat)abstract
- The efficiency of two methods for in-situ fuel removal has been tested on carbon and tungsten limiters retrieved from the TEXTOR and Tore Supra tokamaks: laser-inducedablation of co-deposits and annealing in vacuum at elevated temperature. The analyses of gas phase and surfaces performed with thermal desorption spectrometry, optical spectroscopy, ion beam analysis, surface profilometry and microscopy methods have shown: (i) the ablation leads to the generation of dust particles of 50 nm – 2μm; (ii) volatile products of ablation undergo condensation on surrounding surfaces; (iii) D/C ratio in such condensate is in the range 0.02-0.03; (iv) long-term annealing of 623 K for 70 hours results in release of not more ~10 % of deuterium accumulated in plasma-facing components; (v) effective removal is reached by heating to 900-1300 K.
|
|
| 17. |
- Ivanova, Darya, et al.
(författare)
-
Survey of dust formed in the TEXTOR tokamak : structure and fuel retention
- 2009
-
Ingår i: Physica scripta. T. - : Institute of Physics Publishing (IOPP). - 0281-1847. ; T138, s. 014025-
-
Tidskriftsartikel (refereegranskat)abstract
- A detailed survey of erosion and deposition on plasma-facing components was performed in the TEXTOR tokamak. Co-deposits and dust particles were collected from graphite limiters and from several locations on the Inconel liner. The total amount of dust (loose material), originating mainly from carbon-rich co-deposits detached from the limiters and the liner, was around 2 g, with sizes from 0.1 mu m to 1 mm. The morphology and fuel retention was determined using microscopy methods, ion beam analysis and thermal desorption spectrometry. The study revealed differences in structure and fuel content between deposits from the toroidal and main poloidal limiters. There were also splashes, up to 1 mm in diameter, of molten metal (mainly nickel) on the toroidal limiters. Issues of the dust conversion factor (erosion-to-dust) are addressed and a comparison with results of previous dust surveys at TEXTOR is also briefly presented.
|
|
| 18. |
- 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.
|
|
| 19. |
- Ongena, J, et al.
(författare)
-
Study and design of the ion cyclotron resonance heating system for the stellarator Wendelstein 7-X
- 2014
-
Ingår i: Physics of Plasmas. - : AIP Publishing. - 1089-7674 .- 1070-664X. ; 21:6, s. 061514-
-
Tidskriftsartikel (refereegranskat)abstract
- The current status of the mechanical and electromagnetic design for the ICRF antenna system for W7-X is presented. Two antenna plugins are discussed: one consisting of a pair of straps with pre-matching to cover the first frequency band, 25–38 MHz, and a second one consisting of two short strap triplets to cover a frequency band around 76 MHz. This paper focusses on the two strap antenna for the lower frequency band. Power coupling of the antenna to a reference plasma profile is studied with the help of the codes TOPICA and Microwave Studio that deliver the scattering matrix needed for the optimization of the geometric parameters of the straps and antenna box. Radiation power spectra for different phasings of the two straps are obtained using the code ANTITER II and different heating scenario are discussed. The potential for heating, fast particle generation, and current drive is discussed. The problem of RF coupling through the plasma edge and of edge power deposition is summarized. Important elements of the complete ion cyclotron resonance heating system are discussed: a resonator circuit with tap feed to limit the maximum voltage in the system, and a decoupler to counterbalance the large mutual coupling between the 2 straps. The mechanical design highlights the challenges encountered with this antenna: adaptation to a large variety of plasma configurations, the limited space within the port to accommodate the necessary matching components and the watercooling needed for long pulse operation.
|
|
| 20. |
- Philipps, V., et al.
(författare)
-
Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices
- 2013
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 53:9, s. 093002-
-
Tidskriftsartikel (refereegranskat)abstract
- Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the understanding and validate the models of the in vessel build-up of the T inventory in ITER and future D-T devices. So far, research in these areas is largely supported by post-mortem analysis of wall tiles. However, access to samples will be very much restricted in the next-generation devices (such as ITER, JT-60SA, W7-X, etc) with actively cooled plasma-facing components (PFC) and increasing duty cycle. This has motivated the development of methods to measure the deposition of material and retention of plasma fuel on the walls of fusion devices in situ, without removal of PFC samples. For this purpose, laser-based methods are the most promising candidates. Their feasibility has been assessed in a cooperative undertaking in various European associations under EFDA coordination. Different laser techniques have been explored both under laboratory and tokamak conditions with the emphasis to develop a conceptual design for a laser-based wall diagnostic which is integrated into an ITER port plug, aiming to characterize in situ relevant parts of the inner wall, the upper region of the inner divertor, part of the dome and the upper X-point region.
|
|
| 21. |
- 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.
|
|
| 22. |
- Rubel, Marek, et al.
(författare)
-
Efficiency of fuel removal techniques tested on plasma-facing components from the TEXTOR tokamak
- 2012
-
Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 87:5-6, s. 935-940
-
Tidskriftsartikel (refereegranskat)abstract
- An overview of several techniques considered for fuel and co-deposits removal is given. The methods were tested both on plasma-facing components from the TEXTOR tokamak and on laboratory-prepared layers: (a) chemical approach based on oxidative or nitrogen-assisted plasma; (b) photonic methods with laser-induced fuel desorption or ablation of co-deposits; (c) thermal desorption in vacuum or under oxidative conditions at a broad range of temperatures. The emphasis is on outstanding issues associated with every technique aiming at the reduction of fuel content: the efficiency of fuel and co-deposit removal, the surface state of PFC following the treatment and dust generation.
|
|
| 23. |
|
|
| 24. |
|
|
| 25. |
- 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.
|
|
| 26. |
- Rubel, Marek, et al.
(författare)
-
Tungsten migration studies by controlled injection of volatile compounds
- 2013
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 438, s. S170-S174
-
Tidskriftsartikel (refereegranskat)abstract
- Volatile tungsten hexa-fluoride was locally injected into the TEXTOR tokamak as a marker for material migration studies. The injection was accompanied by puffing N-15 rare isotope as a nitrogen tracer in discharges with edge cooling by impurity seeding. The objective was to assess material balance by qualitative and quantitative determination of a global and local deposition pattern, material mixing effects and fluorine residence in plasma-facing components. Spectroscopy and ex situ ion beam analysis techniques were used. Tungsten was detected on all types of limiter tiles and short-term probes retrieved from the vessel. Over 80% of the injected W was identified. The largest tungsten concentration, 1 x 10(18) cm(-2), was in the vicinity of the gas inlet. Co-deposits contained tungsten and a mix of light isotopes: H, D, He-4, B-10, B-11, C-12, C-13, N-14, N-15, O-16 and small quantities of F-19 thus showing that both He and nitrogen are trapped following wall conditioning (He glow) and edge cooling.
|
|
| 27. |
|
|
| 28. |
|
|
| 29. |
|
|
| 30. |
|
|
| 31. |
- Tanabe, T., et al.
(författare)
-
Application of tungsten for plasma limiters in TEXTOR
- 2000
-
Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 283, s. 1128-1133
-
Tidskriftsartikel (refereegranskat)abstract
- Three different types of W limiters were exposed in the TEXTOR plasma and the response of the plasma and materials performance of the limiters were investigated. 1. A W bulk limiter operated with preheating above 800 K withstood a plasma heat load of about similar to 20 MW/m(2) for a few seconds with some slight surface melting during the highest heat load shot. However, it was severely damaged when operated at around 500 K. 2. A C/W twin test limiter, half made of bulk W and the other half of graphite (EK-98) gave very useful information on how low- and high-Z materials behave under conditions of simultaneous utilization as PFM such as cross-contamination and the influence of a large mass difference on hydrogen reflection and deposition. 3. Two sets of main poloidal W limiters made of vacuum vapor sprayed (VPS)-W deposited on graphite (IG-430U) with a Re interlayer could absorb about 60% of the total convection heat and the ohmic plasma with a density as high as 5 x 10(13) cm(-3) was sustained. Most of the VPS-W coated limiters tolerated a heal load of similar to 20 MW/m(2). This series of W limiters experiments in TEXTOR has shown that W is applicable as a PFM, if its central accumulation is avoided by NBI and/or ICRH heating. Nevertheless, some concerns still remain, including difficulty of plasma startup, W behavior in higher temperature plasmas, and materials' selection.
|
|
| 32. |
- Tanabe, T., et al.
(författare)
-
Material mixing on W/C twin limiter in TEXTOR-94
- 2000
-
Ingår i: Fusion engineering and design. - 0920-3796 .- 1873-7196. ; 49, s. 355-362
-
Tidskriftsartikel (refereegranskat)abstract
- In order to investigate the effect of mutual contamination between tungsten (W) and carbon
|
|
| 33. |
- Wienhold, P., et al.
(författare)
-
Exposure of metal mirrors in the scrape-off layer of TEXTOR
- 2005
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 337-39:03-jan, s. 1116-1120
-
Tidskriftsartikel (refereegranskat)abstract
- 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.
|
|
