| 1. |
- Hirai, T., et al.
(författare)
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Characterization and heat flux testing of beryllium coatings on Inconel for JET ITER-like wall project
- 2007
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Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T128, s. 166-170
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Tidskriftsartikel (refereegranskat)abstract
- In order to perform a fully integrated material test, JET has launched the ITER-like wall project with the aim of installing a full metal wall during the next major shutdown. The material foreseen for the main chamber wall is bulk Be at the limiters and Be coatings on inconel tiles elsewhere. R&D process comprises global characterization ( structure, purity etc) of the evaporated films and testing of their performance under heat loads. The major results are (i) the layers have survived energy loads of 20 MJ m(-2) which is significantly above the required level of 5 - 10 MJ m(-2), (ii) melting limit of beryllium coating would be at the energy level of 30 MJ m(-2), (iii) cyclic thermal load of 10 MJ m(-2) for up to 50 cycles have not induced any noticeable damage such as flaking or detachment.
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| 2. |
- Ivanova, Darya, et al.
(författare)
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Survey of dust formed in the TEXTOR tokamak : structure and fuel retention
- 2009
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Ingår i: Physica scripta. T. - : Institute of Physics Publishing (IOPP). - 0281-1847. ; T138, s. 014025-
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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.
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| 3. |
- Psoda, M., et al.
(författare)
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Material mixing on plasma-facing components : Compound formation
- 2009
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Ingår i: Journal of Nuclear Materials. - : Elsevier. - 0022-3115 .- 1873-4820. ; 386-388, s. 740-743
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Tidskriftsartikel (refereegranskat)abstract
- Two different tungsten limiters (castellated bulk metal block and W-coated graphite), subjected to high power loads in the TEXTOR tokamak, were examined in order to determine chemical composition of deposits inside the castellated grooves and on side surfaces of the coated limiter. Comprehensive analyses carried out by X-ray diffraction, ion beam analysis and other methods revealed: (i) the formation of tungsten oxide (WO2) inside the castellated grooves: (ii) the formation of tungsten carbides (WC main phase and traces of W2C) on side surfaces of the coated limiter. Elemental tungsten was found in deposits on side surfaces only in trace quantities thus indicating that tungsten eroded from the limiter top and transported to the scrape-off layer reacted with carbon. Based on thermodynamic data, the pathways leading to the formation of compounds are discussed.
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| 4. |
- Rubel, Marek J., et al.
(författare)
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An overview of a comprehensive First Mirror Test for ITER at JET
- 2009
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Ingår i: Journal of Nuclear Materials. - Amsterdam : ELSEVIER SCIENCE BV. - 0022-3115 .- 1873-4820. ; 390-91, s. 1066-1069
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Tidskriftsartikel (refereegranskat)abstract
- The test was performed with 32 stainless steel and molybdenum mirrors placed in pan-pipe shaped cassettes and exposed in JET in the divertor and on the main chamber wall for 127000 s including 97000 s of X-point operation. Surface composition and total reflectivity were determined afterwards All mirrors. from the divertor were coated with deuterated carbon deposits causing the reflectivity loss by a factor of 6-10 in the visible range. Flaking and exfoliation of deposits were observed in some cases On the main. chamber wall the deposition occurred mainly on mirrors located deep in cassette channels whereas mirrors close to the channels entrances were free from deposits and retained fair reflectivity (similar to 90% of initial value) especially in the infra-red range. No significant differences in behaviour of steel and molybdenum were noted. The need for development of methods for mirror cleaning and/or protection in a reactor-class device is addressed.
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| 5. |
- Rubel, Marek J., et al.
(författare)
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Beryllium plasma-facing components for the ITER-Like Wall Project at JET
- 2008
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Ingår i: PROCEEDINGS OF THE 17TH INTERNATIONAL VACUUM CONGRESS/13TH INTERNATIONAL CONFERENCE ON SURFACE SCIENCE/INTERNATIONAL CONFERENCE ON NANOSCIENCE AND TECHNOLOGY. - : IOP Publishing.
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Konferensbidrag (refereegranskat)abstract
- ITER-Like Wall Project has been launched at the JET tokamak in order to study a tokamak operation with beryllium components on the main chamber wall and tungsten in the divertor. To perform this first comprehensive test of both materials in a thermonuclear fusion environment, a broad program has been undertaken to develop plasma-facing components and assess their performance under high power loads. The paper provides a concise report on scientific and technical issues in the development of a beryllium first wall at JET.
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| 6. |
- Sergienko, G., et al.
(författare)
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Experience with bulk tungsten test-limiters under high heat loads : melting and melt layer propagation
- 2007
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Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T128, s. 81-86
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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.
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| 7. |
- Sundelin, Per, et al.
(författare)
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A test of nitrogen-assisted plasma discharges for fuel removal from plasma-facing components in tokamaks
- 2008
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Ingår i: PROCEEDINGS OF THE 17TH INTERNATIONAL VACUUM CONGRESS/13TH INTERNATIONAL CONFERENCE ON SURFACE SCIENCE/INTERNATIONAL CONFERENCE ON NANOSCIENCE AND TECHNOLOGY. - Bristol : IOP PUBLISHING LTD. ; , s. 062027-
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Konferensbidrag (refereegranskat)abstract
- Safety regulations limit the amount of tritium accumulated in wall components of a fusion reactor to 350g. Because of this, reduction of long-term fuel inventory is one of the most urgent tasks to be resolved to ensure the safe and economic operation of a reactor-class fusion device. Several methods have been suggested and tested. The aim of this paper is to evaluate the cleaning efficiency of plasma-facing components by ICRH-assisted plasma discharges with in nitrogen-hydrogen in the TEXTOR tokamak. Three types of probes were investigated: laboratory prepared a-C: D layers on silicon; boron layers on silicon obtained by pre-boronisation in TEXTOR and not coated Inconel substrates. The main results are following: (i) laboratory prepared a-C: D layers are not affected: deuterium and carbon contents did not decrease (ii) the morphology of layers pre-boronised in TEXTOR is not affected (iii) no significant effects were noticed on Inconel probes. A comparison of cleaning methods with nitrogen and oxygen is also presented.
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| 8. |
- Sundelin, Per, et al.
(författare)
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Nitrogen-assisted removal of deuterated carbon layers
- 2009
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 390-91, s. 647-650
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Tidskriftsartikel (refereegranskat)abstract
- Deuterated carbon films prepared in laboratory and boronised films prepared in the TEXTOR tokamak were exposed to hydrogen-nitrogen plasmas in order to determine erosion characteristics and fuel removal efficiency. Exposures were performed in: (i) TEXTOR tokamak during ion cyclotron heated wall conditioning discharges (ICWC) and (ii) TOMAS magnetic plasma facility in radio frequency-assisted glow discharges. The essential results are: (i) films exposed in TEXTOR are not affected: deuterium and carbon content does not decrease and the morphology is unchanged, and (ii) deuterium and carbon contents in films exposed in TOMAS is reduced by 30-60% after 2 h of cleaning and topographical changes are noted. The study shows that while exposure to H-2-N-2 laboratory plasma removes a-C:D films, no effect is seen at the position of the sample exposure during tokamak ICWC plasmas. It also indicates that the removal efficiency is only weakly related to nitrogen, since the highest removal efficiency is seen with pure hydrogen plasma. A comparison to oxygen-assisted fuel removal is given.
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