Publications (8)1.2 Total impact
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Article: Fuel Inventory and Co-Deposition in Grooves and Gaps of Divertor and Limiter Structures
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ABSTRACT: Plasma facing components from JET and TEXTOR were studied. The emphasis was on the comparison of co-deposition, material mixing and fuel inventory on plasma facing and side surfaces of tiles, i.e. in gaps separating the tiles. Integrated fuel content in gaps of the Mk-I JET divertor floor was approximately two times greater than detected on the plasma facing surfaces. Taking into account similarities between the Mk-I structure and the castellation in the ITER divertor, the impact of the tile shaping on the tritium inventory is addressed. Deposition on the side of limiter tiles in TEXTOR was around 3–5% of that on the plasma facing surfaces. Experiments aiming at a deeper insight into the deposition on ITER-relevant components are also proposed.Physica Scripta 06/2006; 2004(T111):112. · 1.20 Impact Factor -
Article: Overview of tracer techniques in studies of material erosion, re-deposition and fuel inventory in tokamaks
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ABSTRACT: C-13 labeled methane and rhenium–boron coated plates were used at the JET tokamak as tracers for studies of the material transport, its erosion and re-deposition. Experimental procedures are described. The results are discussed in terms of processes underlying the material transport and the change of morphology of targets exposed to the plasma: physical sputtering, chemical erosion, prompt re-deposition. The influence of wall materials on fuel inventory is also addressed. C-14 measurements in the TEXTOR tokamak are presented and possibilities of using 14 C in carbon migration studies are considered. -
Article: Overview of co-deposition and fuel inventory in castellated divertor structures at JET
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ABSTRACT: The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 x 10(15) cm(-2). Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced. (C) 2007 Elsevier B.V. All rights reserved. -
Article: Overview of Co-deposition and Fuel Inventory in Castellated Structures at JET
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Article: Mirror test for International Thermonuclear Experimental reactor at the JET tokamak: An overview of the program
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ABSTRACT: Metallic mirrors will be essential components of all optical spectroscopy and imaging systems for plasma diagnosis that will be used at the next-step magnetic fusion experiment, International Thermonuclear Experimental Reactor (ITER). Any change of the mirror performance, in particular, reflectivity, will influence the quality and reliability of detected signals. At the instigation of the ITER Design Team, a dedicated technical and experimental activity aiming at the assessment of mirror surface degradation as a result of exposure to the plasma has been initiated on the JET tokamak. This article provides a comprehensive overview of the mirror test program, including design details of the mirror samples and their supports, their locations within JET, and the issue of optical characterization of the mirrors both before and after exposure. The postexposure characterization is particularly challenging in JET as a consequence of an environment in which both tritium and beryllium are present. -
Article: Material erosion and redeposition during the JET MkIIGB-SRP divertor campaign
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Article: Overview of long-term fuel inventory and co-deposition in castellated beryllium limiters at JET
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ABSTRACT: Morphology of castellated Be tiles from the belt limiter exposed to the JET plasma for 56,000 s was examined on both sides of castellated grooves, on plasma-facing and side surfaces of the tiles. The essential results are (i) deuterium retention in the castellated grooves and in other locations is associated with co-deposition of carbon; (ii) the decay length of deposition in the castellation is around 1.5 mm; (iii) no deuterium is detected in bulk Be; (iv) bridging of gaps by molten beryllium occurred but gaps were not filled with Be; (v) on side surfaces of the tiles the formation of BeO layer was detected at a distance of 20 mm and more from the plasma-facing surface. The consequences for a long-term operation of a reactor-class device with several different plasma-facing materials are addressed.Journal of Nuclear Materials. -
Article: Overview of co-deposition and fuel inventory in castellated divertor structures at JET
[show abstract] [hide abstract]
ABSTRACT: The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 × 1015 cm−2. Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced.Journal of Nuclear Materials.
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- Physica Scripta (1)
Institutions
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2006
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KTH Royal Institute of Technology
Stockholm, Stockholm, Sweden
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