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Self-consistent application of ion cyclotron wall conditioning for co-deposited layer removal and recovery of tokamak operation on TEXTOR
KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.ORCID-id: 0000-0001-9901-6296
Vise andre og tillknytning
2013 (engelsk)Inngår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 53, nr 12, s. 123001-Artikkel i tidsskrift (Fagfellevurdert) Published
Abstract [en]

This paper presents a demonstration experiment of ion cyclotron wall conditioning (ICWC) on TEXTOR covering all ITER wall conditioning aims and discusses the implications for ITER. O-2/He-ICWC applied to erode carbon co-deposits removed 6.6x10(21) C-atoms (39 pulses, 158 s cumulated discharge time). Large oxygen retention (71% of injected oxygen) prevented subsequent ohmic discharge initiation. Plasma operation was recovered by a 1h47 multi-pulse D-2-ICWC procedure including pumping time between pulses with duty cycle of 2 s/20 s, cleaning the vessel from oxygen impurities, followed by a 23 min He-ICWC procedure (2 s/20 s), applied to desaturate the deuterium-loaded walls. A stable ohmic discharge was established on the first attempt right after the recovery procedure. The discharges showed improved density control and only slightly increased oxygen characteristic radiation levels (1-1.5 times). After the recovery procedure 36% of the injected O-atoms remained retained in the vessel, derived from mass spectrometry measurements. This amount is in the estimated range for storage in remote areas obtained from surface analysis of locally exposed samples. The removed amount of oxygen by D-2 and He-ICWC obtained from mass spectrometry corresponds to the retention in plasma-wetted areas estimated by surface analysis. It is concluded that most of the removed oxygen stems from plasma-wetted areas while shadowed areas, e. g. behind poloidal limiters, may feature net retention of the discharge gas. On ITER, designed with a shaped first wall, the ICWC plasma-wetted area will approach the total surface area, reducing consequently the retention in remote areas. A tentative extrapolation of the carbon removal on TEXTOR to tritium removal from co-deposits on ITER in the 39 x 4 s O-2/He-ICWC discharges, including pumping time between the RF pulses, corresponds on ITER to a tritium removal in the order of the estimated retention per 400 s DT-burn (140-500 mgT (Shimada and Pitts 2011 J. Nucl. Mater. 415 S1013-6)).

sted, utgiver, år, opplag, sider
Institute of Physics (IOP), 2013. Vol. 53, nr 12, s. 123001-
Emneord [en]
Co-deposited layer, Plasma operations, Poloidal limiters, Recovery procedure, Spectrometry measurements, Tokamak operation, Total surface area, Wall conditioning
HSV kategori
Identifikatorer
URN: urn:nbn:se:kth:diva-136992DOI: 10.1088/0029-5515/53/12/123001ISI: 000327787000003Scopus ID: 2-s2.0-84888986100OAI: oai:DiVA.org:kth-136992DiVA, id: diva2:677713
Merknad

QC 20131210

Tilgjengelig fra: 2013-12-10 Laget: 2013-12-10 Sist oppdatert: 2017-12-06bibliografisk kontrollert
Inngår i avhandling
1. Plasma-Facing Components in Tokamaks: Studies of Wall Conditioning Processes and Plasma Impact on Diagnostic Mirrors
Åpne denne publikasjonen i ny fane eller vindu >>Plasma-Facing Components in Tokamaks: Studies of Wall Conditioning Processes and Plasma Impact on Diagnostic Mirrors
2014 (engelsk)Licentiatavhandling, med artikler (Annet vitenskapelig)
Abstract [en]

Understanding of material migration and its impact on the formation of co-deposited mixed material layers on plasma-facing components is essential for the development of fusion reactors. This thesis focuses on this topic. It is based on experiments performed at JET and TEXTOR tokamaks. The major objectives were to determine: (i) fuel and impurity removal from plasma-facing components by ICWC in different gas mixtures, (ii) fuel and impurity transport connected to ICWC operation, (iii) plasma impact on diagnostic mirrors. All these issues are in line with the ITER needs: mitigation of co-deposition and fuel inventory, and the performance of first mirrors in long-term operation. The novelty in research is demonstrated by several elements. In wall conditioning studies, tracer techniques based on injection of rare isotopes (N-15, O-18) were used to determine conclusively the impact of respective gases. Also, a new approach to ICWC was developed by combining global gas balance studies based on mass spectrometry and the use of multiple surface probes exposed to discharges and then studied ex-situ with accelerator-based techniques. Impact of plasma on diagnostic mirrors was determined after exposure to the entire first experimental campaign in JET-ILW.

sted, utgiver, år, opplag, sider
Stockholm: KTH Royal Institute of Technology, 2014. s. xiv, 44
Serie
TRITA-EE, ISSN 1653-5146 ; 2014:060
Emneord
Plasa-wall interactions, wall conditioning, tracers, diagnostic mirrors, ICWC
HSV kategori
Forskningsprogram
Fysik
Identifikatorer
urn:nbn:se:kth:diva-154621 (URN)978-91-7595-309-0 (ISBN)
Presentation
2014-11-07, Seminar room, Teknikringen 31, KTH- Royal Institute of Technology, Stockholm, 10:00 (engelsk)
Opponent
Veileder
Merknad

QC 20141103

Tilgjengelig fra: 2014-11-03 Laget: 2014-10-26 Sist oppdatert: 2014-11-03bibliografisk kontrollert
2. Impact of material migration on plasma-facing components in tokamaks
Åpne denne publikasjonen i ny fane eller vindu >>Impact of material migration on plasma-facing components in tokamaks
2016 (engelsk)Doktoravhandling, med artikler (Annet vitenskapelig)
Abstract [en]

Plasma-wall interaction plays an essential role in the performance and safety of a fusion reactor. This thesis focuses on the impact of material migration on plasma-facing components. It is based on experiments performed in tokamaks: JET, TEXTOR and ASDEX Upgrade. The objectives of the experiments were to assess fuel and impurity removal under ion cyclotron wall conditioning (ICWC) and plasma impact on diagnostic mirrors.

In wall conditioning studies, tracer techniques based on the injection of rare isotopes (15N, 18O) were used to determine conclusively the impact of the respective gases. For the first time, probe surfaces and wall components exposed to ICWC were examined by surface analysis methods. Discharges in hydrogen were the most efficient to erode carbon co-deposits, resulting in a reduction of the initial deuterium content by a factor of two. It was also found that impurities desorbed under ICWC are partly re-deposited on the wall.

Plasma impact on diagnostic mirrors was determined by surface analysis of test mirrors exposed at JET. Reflectivity of mirrors from the divertor region was severely decreased due to deposits of beryllium, deuterium, carbon and other impurities. This result points out the need to develop mirror maintenance procedures. Neutron damage on mirrors was simulated by ion irradiation in an ion implanter. It was shown that damage levels similar to those expected in the first wall of a fusion reactor do not produce a significant change in reflectivity.

sted, utgiver, år, opplag, sider
Stockholm: KTH Royal Institute of Technology, 2016. s. 54
Serie
TRITA-EE, ISSN 1653-5146
Emneord
Fusion, material migration, wall conditioning, diagnostic mirrors, plasma-facing materials
HSV kategori
Forskningsprogram
Fysik
Identifikatorer
urn:nbn:se:kth:diva-190903 (URN)978-91-7729-046-9 (ISBN)
Disputas
2016-09-15, Kollegiesalen, Brinellvägen 8, Stockholm, 10:00 (engelsk)
Opponent
Veileder
Merknad

QC 20160819

Tilgjengelig fra: 2016-08-19 Laget: 2016-08-18 Sist oppdatert: 2016-08-19bibliografisk kontrollert

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