| 1. |
- Hirai, T., et al.
(författare)
-
R&D on full tungsten divertor and beryllium wall for JET ITER-like wall project
- 2007
-
Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 82:15-24, s. 1839-1845
-
Tidskriftsartikel (refereegranskat)abstract
- The ITER reference materials have been tested separately in tokamaks, plasma simulators, ion beams and high heat flux test beds. 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. As a result of R&D projects in 2005-2006, bulk tungsten tiles are foreseen at the outer horizontal target and tungsten coating at the other divertor tiles. In some regions of the main chamber, beryllium coated Inconel tiles and bulk beryllium tiles are utilised which include marker tiles as erosion diagnostics. This paper gives an overview of the R&D carried out in the frame of the ITER-like Wall Project on the development of an inertially cooled bulk tungsten tile design and the characterization of tungsten and beryllium coating technologies.
|
|
| 2. |
|
|
| 3. |
- Kreter, A., et al.
(författare)
-
Deuterium retention in different carbon materials exposed in TEXTOR :
- 2008
-
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. 062024-
-
Konferensbidrag (refereegranskat)abstract
- CFC NB31, CFC DMS780 and fine-grain graphite EK98 were simultaneously exposed in the SOL of TEXTOR in order to measure the retention of deuterium in the material bulk. After exposure, the samples were analysed by thermal desorption spectrometry and nuclear reaction analysis with a conventional beam as well as with a microbeam. The deuterium retention amounts (2-4).10(21) D/m(2) for the incident fluence of similar to 2.10(25) D/m(2). The retention values are similar for both CFC materials and lower by similar to 20-40% for EK98. The retention in all three materials scales roughly with a square root of incident fluence without saturation for the range of fluences obtained. The majority of deuterium is stored in a surface layer of <8 mu m. However, in NB31 deuterium is detectable as deep as 80 mu m. The in-bulk retention estimated for a TEXTOR experimental campaign of approximate to 7500 s of plasma has a contribution of approximate to 10% to the total retention, which is dominated by deuterium-carbon co-deposition.
|
|
| 4. |
- Litnovsky, A., et al.
(författare)
-
Carbon deposition and fuel accumulation in a castellated limiter exposed under erosion-dominated conditions in the SOL of TEXTOR
- 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. 361-364
-
Konferensbidrag (refereegranskat)abstract
- Castellated limiter with ITER-like geometry was exposed under the erosion-dominated conditions in the SOL of TEXTOR. After the exposure, deposits were found in the gaps, with a negligible amount of deposited material that was detected on top surfaces open to the plasma. The thickest deposits are located on plasma shadowed sides of the gaps with maximum thickness up to 0.5 μm on the plasma closest locations as it was measured with SIMS diagnostic. The deposit thickness decrease exponentially with the depth of gaps with e-folding length of 1.7-2 mm as inferred from EPMA measurements. Material intermixing was found to occur in the deposited layers. Deposits are mixed Mo:C:D:O:B:H layers. Gaps contain at least (5÷10)×1016 of D fuel atoms per cm2. This corresponds roughly to 0.02÷0.04 % of total averaged D fluence impinging on the castellation.
|
|
| 5. |
- Litnovsky, Andrey, et al.
(författare)
-
Carbon transport, deposition and fuel accumulation in castellated structures exposed in TEXTOR
- 2007
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 367, s. 1481-1486
-
Tidskriftsartikel (refereegranskat)abstract
- In order to maintain the thermo-mechanical durability of ITER it is proposed to castellate the interior surface of the first wall and divertor by splitting them into small-size cells [W. Daener et a]., Fusion Eng. Des. 61&62 (2002) 61]. A concern is the accumulation of fuel in the gaps of the castellation. In TEXTOR, molybdenum limiters were exposed in the scrape-off layer (SOL) plasma to assess fuel accumulation. The first limiter was exposed under deposition-dominated conditions. Carbon deposits were formed both on top surfaces and in the gaps. About 0.12% of the impinging D-fluence was found in the gaps. Another castellated limiter was exposed under erosion-dominated conditions. Deposited layers were found only on the plasma shadowed areas of the gaps. A significant amount of molybdenum from the limiter was found intermixed in the deposit. The gaps contained similar to 0.03% of the impinging D-fluence. Modeling was performed to simulate carbon transport into the gaps.
|
|
| 6. |
- Litnovsky, A., et al.
(författare)
-
Diagnostic mirrors for ITER : A material choice and the impact of erosion and deposition on their performance
- 2007
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 363, s. 1395-1402
-
Tidskriftsartikel (refereegranskat)abstract
- Metal mirrors will be implemented in about half of the ITER diagnostics. Mirrors in ITER will have to withstand radiation loads, erosion by charge-exchange neutrals, deposition of impurities, particle implantation and neutron irradiation. It is believed that the optical properties of diagnostic mirrors will be primarily influenced by erosion and deposition. A solution is needed for optimal performance of mirrors in ITER throughout the entire lifetime of the machine. A multi-machine research on diagnostic mirrors is currently underway in fusion facilities at several institutions and laboratories worldwide. Among others, dedicated investigations of ITER-candidate mirror materials are ongoing in Tore-Supra, TEXTOR, DIII-D, TCV, T-10 and JET. Laboratory studies are underway at IPP Kharkov (Ukraine), Kurchatov Institute (Russia) and the University of Basel (Switzerland). An overview of current research on diagnostic mirrors along with an outlook on future investigations is the subject of this paper.
|
|
| 7. |
- Litnovsky, A., et al.
(författare)
-
Direct comparative test of single crystal and polycrystalline diagnostic mirrors exposed in TEXTOR in erosion conditions
- 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. 1726-1729
-
Konferensbidrag (refereegranskat)abstract
- First direct comparative test of single crystal and polycrystalline diagnostic mirror materials under erosion conditions has been made in TEXTOR. Before exposure in TEXTOR, glow discharge cleaning has efficiently restored the reflectivity of initially oxidized mirrors. After the exposure, no significant changes in total reflectivity were observed. Drastic increase of diffuse reflectivity was measured for polycrystalline molybdenum mirror, but not for the single crystal. Thus, specular reflectivity of single crystal is significantly higher than of polycrystalline one. The most affected wavelength range is 250-1000 nm, no significant changes of reflectivity was noticed in the range 1000-2000 nm. No or negligible effect of erosion on polarization characteristics of mirrors was measured.
|
|
| 8. |
- Maier, H., et al.
(författare)
-
Tungsten and beryllium armour development for the JET ITER-like wall project
- 2007
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 47:3, s. 222-227
-
Tidskriftsartikel (refereegranskat)abstract
- For the ITER-like wall project at JET the present main chamber CFC tiles will be exchanged with Be tiles and in parallel a fully tungsten-clad divertor will be prepared. Therefore three R&D programmes were initiated: Be coatings on Inconel as well as Be erosion markers were developed for the first wall of the main chamber. High heat flux screening and cyclic loading tests carried out on the Be coatings on Inconel showed excellent performance, above the required power and energy density. For the divertor a conceptual design for a bulk W horizontal target plate was investigated, with the emphasis on minimizing electromagnetic forces. The design consisted of stacks of W lamellae of 6 mm width that were insulated in the toroidal direction. High heat flux tests of a test module were performed with an electron beam at an absorbed power density Up to 9 MW m(-2) for more than 150 pulses and finally with increasing power loads leading to surface temperatures in excess of 3000 degrees C. No macroscopic failure occurred during the test while SEM showed the development of micro-cracks on the loaded surface. For all other divertor parts R&D was performed to provide the technology to coat the 2-directional CFC material used at JET with thin tungsten coatings. The W-coated CFC tiles were subjected to heat loads with power densities ranging up to 23.5 MW m(-2) and exposed to cyclic heat loading for 200 pulses at 10.5 MW m(-2). All coatings developed cracks perpendicular to the CFC fibres due to the stronger contraction of the coating upon cool-down after the heat pulses.
|
|
| 9. |
|
|
| 10. |
- 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.
|
|
