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Removal of beryllium-containing films deposited in JET from mirror surfaces by laser cleaning
EURATOM/CCFE Fusion Association, Culham Science Centre, UK.
EURATOM/CCFE Fusion Association, Culham Science Centre, UK.
FOM Insititute for Plasma Physics Rijnhuizen, The Netherlands.
CEA Saclay, DEN/DANS/DPC/SCP/LILM, Farnce.
Show others and affiliations
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.

Place, publisher, year, edition, pages
2011. Vol. 415, no 1, S1199-S1202 p.
Keyword [en]
Cleaning trials, Laser cleaning, Laser systems, Mirror surfaces, Visible spectra
National Category
Fusion, Plasma and Space Physics
Identifiers
URN: urn:nbn:se:kth:diva-29587DOI: 10.1016/j.jnucmat.2010.11.076ISI: 000298128100273Scopus ID: 2-s2.0-80054843301OAI: oai:DiVA.org:kth-29587DiVA: diva2:396338
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
In thesis
1. Fuel retention and fuel removal from first wall components in tokamaks
Open this publication in new window or tab >>Fuel retention and fuel removal from first wall components in tokamaks
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:nbn:se:kth:diva-29588 (URN)978-91-7415-867-0 (ISBN)
Note
QC 20110209Available from: 2011-02-09 Created: 2011-02-09 Last updated: 2011-03-10Bibliographically approved
2. Plasma-Facing Components in Tokamaks: Material Modification and Fuel Retention
Open this publication in new window or tab >>Plasma-Facing Components in Tokamaks: Material Modification and Fuel Retention
2012 (English)Doctoral 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 the steady-state operation of a fusion reactor, e.g. ITER. These topics have been explored in this thesis in order to contribute to a better understanding and the 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 the Nuclear Research Center Cadarache (France) and JET in Culham Centre for Fusion Energy (United Kingdom). Following issues were addressed: (a) properties of material migration products, i.e. co-deposited layers and dust particles; (b) impact of fuel removal methods on dust generation and on modification of plasma-facing components; (c) efficiency of fuel and deposit removal techniques; (d) degradation mechanism of diagnostic components - mirrors - and methods of their regeneration.

Place, publisher, year, edition, pages
Stockholm: KTH Royal Institute of Technology, 2012. xiv, 58 p.
Series
Trita-EE, ISSN 1653-5146 ; 2012:058
Keyword
magnetic confinement fusion, plasma-facing components, plasma-facing materials, fuel inventory, erosion and deposition
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-105099 (URN)978-91-7501-567-5 (ISBN)
Public defence
2012-12-12, F3, Lindstedtsvagen 26, KTH, Stockholm, 10:00 (English)
Opponent
Supervisors
Note

QC 20121116

Available from: 2012-11-16 Created: 2012-11-16 Last updated: 2012-11-16Bibliographically approved

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Rubel, Marek J.

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