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Fuel retention and fuel removal from first wall components in tokamaks
KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
2011 (English)Licentiate thesis, comprehensive summary (Other academic)
Abstract [en]

Fuel inventory and generation of carbon and metal dust in a tokamak are perceived to be serious safety and economy issues for a steady-state operation of a fusion reactor, e.g. ITER. These topics have been explored in the on-going Ph.D. work in order to contribute to the better understanding and development of methods for controlling and curtailing fuel accumulation and dust formation in controlled fusion devices. The work was carried out with material facing fusion plasmas in three tokamaks: TEXTOR in Forschungszentrum Jülich (Germany), Tore Supra in Cadarache (France) and JET in Culham Centre for Fusion Energy (United Kingdom).

 This thesis provides an account on studies of fuel removal techniques from plasmafacing components (PFCs) and on consequences of dust formation. Following issues are addressed: (a)  properties of carbon and metal dust formed in the TEXTOR tokamak;  (b)  dust generation associated with removal of fuel and co-deposited layers from carbon PFCs from TEXTOR and Tore Supra;  (c)  surface morphology of wall components after different cleaning treatments;  (d)  surface properties of diagnostic mirrors tested at JET for ITER. The study dealt with carbon, tungsten and beryllium, i.e. with the three major elements being used for PFC in present-day devices and foreseen for a next-step machine.

 Some essential results are summarised by the following.

 (i)  The amount of loose dust found on the floor of the TEXTOR liner does not exceed 2 grams with particle size range 0.1 mm – 1 mm. The presence of fine (up to 1 mm) crystalline graphite in the collected matter suggests that brittle destruction of carbon PFC could take place during off-normal events. Carbon is the main component, but there are also magnetic and non-magnetic metal agglomerates. The results obtained strongly indicate that in a carbon wall machine the disintegration of flaking co-deposits on PFC is the main source of dust:  (ii)  The fuel content in dust and co-deposits varies from 10% on the main limiters to 0.03% on the neutralizer plates as determined by thermal desorption and ionbeam methods:  (iii)  Fuel removal by annealing in vacuum or by oxidative methods disintegrates codeposits. In the case of thick layers, the treatment makes them brittle thus reducing the adherence to the target and, as a consequence, this leads to the formation of dust:  (iv)   Application of thermal methods for fuel removal from carbon-rich layers is effective only at high temperatures (above 800 K), i.e. in the range exceeding the allowed baking temperature of the ITER divertor:  (v)   Photonic cleaning by laser pulses effectively removes fuel-rich deposited layers, but it also produces debris, especially under ablation conditions:  (vi)  Photonic cleaning of mirrors exposed in JET results in partial recovery of reflectivity, but surfaces are modified by laser pulses.

The presentation of results is accompanied by a discussion of their consequences for the future development and the application of fuel and dust removal methods in a next-step fusion device.

Place, publisher, year, edition, pages
Stockholm: KTH , 2011. , ix, 40 p.
Series
Trita-EE, ISSN 1653-5146 ; 2011:011
National Category
Fusion, Plasma and Space Physics
Identifiers
URN: urn:nbn:se:kth:diva-29588ISBN: 978-91-7415-867-0 (print)OAI: oai:DiVA.org:kth-29588DiVA: diva2:396370
Note
QC 20110209Available from: 2011-02-09 Created: 2011-02-09 Last updated: 2011-03-10Bibliographically approved
List of papers
1. Survey of dust formed in the TEXTOR tokamak: structure and fuel retention
Open this publication in new window or tab >>Survey of dust formed in the TEXTOR tokamak: structure and fuel retention
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2009 (English)In: Physica scripta. T, ISSN 0281-1847, Vol. T138, 014025- p.Article in journal (Refereed) Published
Abstract [en]

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.

Place, publisher, year, edition, pages
Institute of Physics Publishing (IOPP), 2009
Keyword
CONTROLLED FUSION DEVICES, CO-DEPOSITED LAYERS, PLASMA-FACING COMPONENTS, HIGH HEAT-FLUX, CARBON, PARTICLES, EROSION, ITER, CONSEQUENCES, EMISSION
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-25575 (URN)10.1088/0031-8949/2009/T138/014025 (DOI)000273199200026 ()2-s2.0-77953881798 (Scopus ID)
Conference
12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications Julich, GERMANY, MAY, 2009
Note

QC 20101119

Available from: 2012-01-26 Created: 2010-10-27 Last updated: 2017-12-12Bibliographically approved
2. Dust particles in controlled fusion devices: generation mechanism and analysis
Open this publication in new window or tab >>Dust particles in controlled fusion devices: generation mechanism and analysis
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2009 (English)In: 36th EPS Conference on Plasma Physics and Controlled Fusion, 2009, 129-132 p.Conference paper, Published paper (Refereed)
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-29585 (URN)2-s2.0-84872693573 (Scopus ID)978-162276336-8 (ISBN)
Conference
36th European Physical Society Conference on Plasma Physics 2009, EPS 2009; Sofia; Bulgaria; 29 June 2009 through 3 July 2009
Note

QC 20110209

Available from: 2011-02-09 Created: 2011-02-09 Last updated: 2016-12-14Bibliographically approved
3. Laser-based and thermal methods for fuel removal and cleaning of plasma-facing components
Open this publication in new window or tab >>Laser-based and thermal methods for fuel removal and cleaning of plasma-facing components
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2011 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 415, no 1, S801-S804 p.Article in journal (Refereed) Published
Abstract [en]

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.

Keyword
carbon, ´first wall materials, hydrogen, laser, tungsten
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-29586 (URN)10.1016/j.jnucmat.2011.01.119 (DOI)000298128100183 ()2-s2.0-80054839170 (Scopus ID)
Note
Updated from in Press to Published. QC 20120227Available from: 2011-02-09 Created: 2011-02-09 Last updated: 2017-12-11Bibliographically approved
4. Removal of beryllium-containing films deposited in JET from mirror surfaces by laser cleaning
Open this publication in new window or tab >>Removal of beryllium-containing films deposited in JET from mirror surfaces by laser cleaning
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2011 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, ISSN 0022-3115, Vol. 415, no 1, S1199-S1202 p.Article in journal (Refereed) Published
Abstract [en]

A set of stainless steel (SS) and molybdenum mirror samples located in the divertor and at the outer mid-plane of the vessel were exposed in JET from 2005 to 2007. A selection of these mirror samples with well adhered deposits (i.e. not flaking) of up to a few hundred nanometers in thickness and with Be/C ratios ranging from 0 to similar to 1 have been cleaned using a laser system developed at CEA, Saclay. Following laser cleaning the recovered reflectivity was generally better in the infrared than the visible spectrum, with recovery of up to 90% of the initial reflectivity being obtained at 1600 nm for both Mo and SS mirrors falling as low as 20-30% of initial reflectivity at a wavelength of 400 nm for some SS mirrors, rising to similar to 80% for Mo mirrors. Some deposit remained on the mirrors after the cleaning trials.

Keyword
Cleaning trials, Laser cleaning, Laser systems, Mirror surfaces, Visible spectra
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-29587 (URN)10.1016/j.jnucmat.2010.11.076 (DOI)000298128100273 ()2-s2.0-80054843301 (Scopus ID)
Conference
19th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI), May 24-28, 2010,Univ Calif, Gen Atom, San Diego, CA
Note

QC 20110209

Available from: 2011-02-09 Created: 2011-02-09 Last updated: 2017-12-11Bibliographically approved

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