Self-consistent application of ion cyclotron wall conditioning for co-deposited layer removal and recovery of tokamak operation on TEXTOR
2013 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 53, no 12, 123001- p.Article in journal (Refereed) Published
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)).
Place, publisher, year, edition, pages
Institute of Physics (IOP), 2013. Vol. 53, no 12, 123001- p.
Co-deposited layer, Plasma operations, Poloidal limiters, Recovery procedure, Spectrometry measurements, Tokamak operation, Total surface area, Wall conditioning
Fusion, Plasma and Space Physics
IdentifiersURN: urn:nbn:se:kth:diva-136992DOI: 10.1088/0029-5515/53/12/123001ISI: 000327787000003ScopusID: 2-s2.0-84888986100OAI: oai:DiVA.org:kth-136992DiVA: diva2:677713
QC 201312102013-12-102013-12-102016-08-19Bibliographically approved