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Global material migration in tokamaks: patterns, material balance and transport mechanisms in TEXTOR
KTH, School of Electrical Engineering (EES), Space and Plasma Physics.ORCID iD: 0000-0003-1062-8101
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2017 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326Article in journal, Editorial material (Refereed) Submitted
Abstract [en]

As the last experiment before the final shutdown of the TEXTOR tokamak, equipped with all-graphite plasma-facing components (PFCs), MoF6 had been injected into the vacuum vessel. During decommissioning all PFCs became available for surface studies, enabling detailed mapping of previously injected Mo, W and intrinsic 625 Inconel metals. As a result, detailed deposition patterns for these metals were obtained, revealing a number of findings: a) High-Z metals are mainly globally deposited, with concentrations decaying exponentially with distance from the injection source; b) the decay length is of the order of 0.1 m on the main PFC and 1.0 m on the receded PFC; c) ion flow velocities co-decide the position of maximum deposition. Modelling with ERO shows exponential decay. Simulated decay lengths are between 0.15 – 1.30 m, depending on the anomalous cross field diffusion coefficient. Extensive measurements of Mo have been undertaken in order to quantify the amounts deposited on the graphite PFCs, showing that the ratio of local to global deposited high-Z metals is 0.3-0.4. However, only up to 20% of the injected Mo could be detected on all the PFCs. A large fraction of injected Mo may have been pumped out before being deposited. The bumper limiter is found to be the major repository for all probed elements. For W and F, this is solely due to the limiter size – the areal concentration is not enhanced. For Mo, Inconel metals, 15N and D the areal concentration on the bumper limiter is higher than on the toroidal belt limiter ALTII acting as main PFC.

Place, publisher, year, edition, pages
2017.
National Category
Fusion, Plasma and Space Physics
Identifiers
URN: urn:nbn:se:kth:diva-210414OAI: oai:DiVA.org:kth-210414DiVA: diva2:1118346
Note

QC 20170630

Available from: 2017-06-30 Created: 2017-06-30 Last updated: 2017-06-30Bibliographically approved
In thesis
1. Material migration in tokamaks: Erosion-deposition patterns and transport processes
Open this publication in new window or tab >>Material migration in tokamaks: Erosion-deposition patterns and transport processes
2017 (English)Doctoral thesis, comprehensive summary (Other academic)
Abstract [en]

Controlled thermonuclear fusion may become an attractive future electrical power source. The most promising of all fusion machine concepts is called a tokamak. The fuel, a plasma made of deuterium and tritium, must be confined to enable the fusion process. It is also necessary to protect the wall of tokamaks from erosion by the hot plasma. To increase wall lifetime, the high-Z metal tungsten is foreseen as wall material in future fusion devices due to its very high melting point. This thesis focuses on the following consequences of plasma impact on a high-Z wall: (i) erosion, transport and deposition of high-Z wall materials; (ii) fuel retention in tokamak walls; (iii) long term effects of plasma impact on structural machine parts; (iv) dust production in tokamaks.

An extensive study of wall components has been conducted with ion beam analysis after the final shutdown of the TEXTOR tokamak. This unique possibility offered by the shutdown combined with a tracer experiment led to the largest study of high-Z metal migration and fuel retention ever conducted. The most important results are:

 

- transport is greatly affected by drifts and flows in the plasma edge;

- stepwise transport along wall surfaces takes place mainly in the toroidal direction;

- fuel retention is highest on slightly retracted wall elements;

- fuel retention is highly inhomogeneous.

 

A broad study on structural parts of a tokamak has been conducted on the TEXTOR liner. The plasma impact does neither degrade mechanical properties nor lead to fuel diffusion into the bulk after 26 years of duty time. Peeling deposition layers on the liner retain fuel in the order of 1g and represent a dust source. Only small amounts of dust are found in TEXTOR with overall low deuterium content. Security risks in future fusion devices due to dust explosions or fuel retention in dust are hence of lesser concern.

Place, publisher, year, edition, pages
Stockholm: KTH Royal Institute of Technology, 2017. 109 p.
Series
TRITA-EE, ISSN 1653-5146 ; 2017:060
Keyword
TEXTOR, fusion, plasma physics, transport, migration, tracers, tokamak, limiter, divertor, high-Z, ion beam analysis, Rutherford backscattering, Nuclear reaction analysis, Elastic recoil detection analysis
National Category
Other Electrical Engineering, Electronic Engineering, Information Engineering
Research subject
Electrical Engineering
Identifiers
urn:nbn:se:kth:diva-209758 (URN)978-91-7729-461-0 (ISBN)
Public defence
2017-09-19, F3, Lindstedtsvägen 26, Stockholm, 14:59 (English)
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Note

QC 20170630

Available from: 2017-06-30 Created: 2017-06-22 Last updated: 2017-06-30Bibliographically approved

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