| 2. |
- Ivanova, Darya, et al.
(författare)
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Laser-based and thermal methods for fuel removal and cleaning of plasma-facing components
- 2011
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 415:1, s. S801-S804
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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.
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| 3. |
- Philipps, V., et al.
(författare)
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Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices
- 2013
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 53:9, s. 093002-
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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.
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