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Impact of material migration on plasma-facing components in tokamaks
KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.ORCID iD: 0000-0001-5603-8559
2016 (English)Doctoral thesis, comprehensive summary (Other academic)
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.

Place, publisher, year, edition, pages
Stockholm: KTH Royal Institute of Technology, 2016. , 54 p.
Series
TRITA-EE, ISSN 1653-5146
Keyword [en]
Fusion, material migration, wall conditioning, diagnostic mirrors, plasma-facing materials
National Category
Fusion, Plasma and Space Physics
Research subject
Physics
Identifiers
URN: urn:nbn:se:kth:diva-190903ISBN: 978-91-7729-046-9OAI: oai:DiVA.org:kth-190903DiVA: diva2:953669
Public defence
2016-09-15, Kollegiesalen, Brinellvägen 8, Stockholm, 10:00 (English)
Opponent
Supervisors
Note

QC 20160819

Available from: 2016-08-19 Created: 2016-08-18 Last updated: 2016-08-19Bibliographically approved
List of papers
1. Impact of ion cyclotron wall conditioning on fuel removal from plasma-facing components at TEXTOR
Open this publication in new window or tab >>Impact of ion cyclotron wall conditioning on fuel removal from plasma-facing components at TEXTOR
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2014 (English)In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T159, 014017- p.Article in journal (Refereed) Published
Abstract [en]

Ion cyclotron wall conditioning (ICWC) is based on low temperature and low density plasmas produced and sustained by ion cyclotron resonance (ICR) pulses in reactive or noble gases. The technique is being developed for ITER. It is tested in tokamaks in the presence of toroidal magnetic field (0.2-3.8 T) and heating power of the order of 10(5) W. ICWC with hydrogen, deuterium and oxygen-helium mixture was studied in the TEXTOR tokamak. The exposed samples were pre-characterized limiter tiles mounted on specially designed probes. The objectives were to assess the reduction of deuterium content, the uniformity of the reduction and the retention of seeded oxygen. For the last objective oxygen-18 was used as a marker. ICWC in hydrogen caused a drop of deuterium content in the tile by a factor of more than 2: from 4.5x10(18) to 1.9x10(18) D cm(-2). A decrease of the fuel content by approximately 25% was achieved by the ICWC in oxygen, while no reduction of the fuel content was measured after exposure to discharges in deuterium. These are the first data ever obtained showing quantitatively the local decrease of deuterium in wall components treated by ICWC in a tokamak. The oxygen retention in the tiles exposed to ICWC with oxygen-helium was analyzed for different orientations and radial positions with respect to plasma. An average retention of 1.38x10(16) O-18 cm(-2) was measured. A maximum of the retention, 4.4x10(16) O-18 cm(-2), was identified on a sample surface near the plasma edge. The correlation with the gas inlet and antennae location has been studied.

Keyword
ion cyclotron wall conditioning, plasma-facing components, fuel retention, fuel removal, TEXTOR
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-145840 (URN)10.1088/0031-8949/2014/T159/014017 (DOI)000334847800018 ()2-s2.0-84902191459 (ScopusID)
Conference
14th International Conference on Plasma-Facing Materials and Components for Fusion Applications, PFMC 2013; Julich; Germany; 13 May 2013 through 17 May 2013
Funder
Swedish Research Council, 621-2009-4138 621-2012-4148
Note

QC 20140602

Available from: 2014-06-02 Created: 2014-06-02 Last updated: 2016-08-19Bibliographically approved
2. Self-consistent application of ion cyclotron wall conditioning for co-deposited layer removal and recovery of tokamak operation on TEXTOR
Open this publication in new window or tab >>Self-consistent application of ion cyclotron wall conditioning for co-deposited layer removal and recovery of tokamak operation on TEXTOR
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2013 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 53, no 12, 123001- p.Article in journal (Refereed) 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)).

Place, publisher, year, edition, pages
Institute of Physics (IOP), 2013
Keyword
Co-deposited layer, Plasma operations, Poloidal limiters, Recovery procedure, Spectrometry measurements, Tokamak operation, Total surface area, Wall conditioning
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-136992 (URN)10.1088/0029-5515/53/12/123001 (DOI)000327787000003 ()2-s2.0-84888986100 (ScopusID)
Note

QC 20131210

Available from: 2013-12-10 Created: 2013-12-10 Last updated: 2016-08-19Bibliographically approved
3. Nitrogen removal from plasma-facing components by ion cyclotron wall conditioning in TEXTOR
Open this publication in new window or tab >>Nitrogen removal from plasma-facing components by ion cyclotron wall conditioning in TEXTOR
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2015 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, 688-692 p.Article in journal (Other academic) Published
Abstract [en]

The efficiency of ion cyclotron wall conditioning (ICWC) in the removal of nitrogen from plasma-facing components in TEXTOR was assessed. In two experiments the wall was loaded with nitrogen and subsequently cleaned by ICWC in deuterium and helium. The retention and removal of nitrogen was studied in-situ by means of mass spectrometry, and ex-situ by surface analysis of a set of graphite, tungsten and TZM plates installed on test limiter systems. N-15 rare isotope was used as a marker. The results from the gas balance showed that about 25% of the retained nitrogen was removed after ICWC cleaning, whereas surface analysis of the plates based on ToF-HIERDA showed an increase of the deposited species after the cleaning. This indicates that during ICWC operation on carbon devices, nitrogen is not only pumped out but also transported to other locations on the wall. Additionally, deuterium surface content was studied before and after ICWC cleaning.

National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-155184 (URN)10.1016/j.jnucmat.2014.11.113 (DOI)000358467200144 ()2-s2.0-84937731653 (ScopusID)
Funder
Swedish Research Council, 621-2012-4148
Note

Updated from "In press" to "Published". QC 20150901

Available from: 2014-11-03 Created: 2014-11-03 Last updated: 2016-08-19Bibliographically approved
4. Investigation of probe surfaces after ion cyclotron wall conditioning in ASDEX Upgrade
Open this publication in new window or tab >>Investigation of probe surfaces after ion cyclotron wall conditioning in ASDEX Upgrade
(English)Manuscript (preprint) (Other academic)
Abstract [en]

For the first time, material analysis techniques have been applied to study the effect of ion cyclotron wall conditioning (ICWC) on probe surfaces in a metal-wall machine. ICWC is a technique envisaged to contribute to the removal of fuel and impurities from the first wall of ITER. The objective of this work was to assess impurity migration under the ICWC operation. Tungsten probes were exposed in ASDEX Upgrade to discharges in helium. After wall conditioning, the probes were covered with a co-deposited layer containing D, B, C, N, O and relatively high amount of He on virgin W substrates. The concentration ratio He/C+B was 0.7. The formation of the co-deposited layer indicates that a fraction of the impurities desorbed from the wall under ICWC operation are transported by plasma and deposited away from their original location.

Keyword
ICWC, Erosion-deposition, fuel removal, ion beam analysis, ASDEX Upgrade
National Category
Physical Sciences
Research subject
Physics
Identifiers
urn:nbn:se:kth:diva-190898 (URN)
External cooperation:
Note

QC 20160819

Available from: 2016-08-18 Created: 2016-08-18 Last updated: 2016-09-02Bibliographically approved
5. An overview of the comprehensive First Mirror Test in JET with ITER-like wall
Open this publication in new window or tab >>An overview of the comprehensive First Mirror Test in JET with ITER-like wall
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2014 (English)In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T159, 014011- p.Article in journal (Refereed) Published
Abstract [en]

The First Mirror Test in Joint European Torus (JET) with the International Thermonuclear Experimental Reactor-like wall was performed with polycrystalline molybdenum mirrors. Two major types of experiments were done. Using a reciprocating probe system in the main chamber, a short-term exposure was made during a 0.3 h plasma operation in 71 discharges. The impact on reflectivity was negligible. In a long-term experiment lasting 19 h with 13 h of X-point plasma, 20 Mo mirrors were exposed, including four coated with a 1 mu m-thick Rh layer. Optical performance of all mirrors exposed in the divertor was degraded by up to 80% because of beryllium, carbon and tungsten co-deposits on surfaces. Total reflectivity of most Mo mirrors facing plasma in the main chamber was only slightly affected in the spectral range 400-1600 nm, while the Rh-coated mirror lost its high original reflectivity by 30%, thus decreasing to the level typical of molybdenum surfaces. Specular reflectivity was decreased most strongly in the 250-400 nm UV range. Surface measurements with x-ray photoelectron spectroscopy and depth profiling with secondary ion mass spectrometry and heavy-ion elastic recoil detection analysis (ERDA) revealed that the very surface region on both types of mirrors had been modified by neutrals, resulting eventually in the composition change: Be, C, D at the level below 1x10(16) cm(-2) mixed with traces of Ni, Fe in the layer 10-30 nm thick. On several exposed mirrors, the original matrix material (Mo) remained as the major constituent of the modified layer. The data obtained in two major phases of the JET operation with carbon and full metal walls are compared. The implications of these results for first mirrors and their maintenance in a reactor-class device are discussed.

Keyword
diagnostic mirrors, surface analysis, reflectivity, erosion and deposition, JET, ITER-like wall
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-145842 (URN)10.1088/0031-8949/2014/T159/014011 (DOI)000334847800012 ()2-s2.0-84902125224 (ScopusID)
Conference
14th International Conference on Plasma-Facing Materials and Components for Fusion Applications, PFMC 2013; Julich; Germany; 13 May 2013 through 17 May 2013
Funder
Swedish Research Council, 621-2009-4138 621-2012-4148
Note

QC 20140602

Available from: 2014-06-02 Created: 2014-06-02 Last updated: 2016-08-19Bibliographically approved
6. Plasma impact on diagnostic mirrors in JET
Open this publication in new window or tab >>Plasma impact on diagnostic mirrors in JET
(English)Manuscript (preprint) (Other academic)
Abstract [en]

All optical diagnostics in ITER will rely on metallic mirrors acting as plasma-facing components. This contribution provides a comprehensive account on plasma impact on diagnostic mirrors in JET with the ITER-Like Wall. Specimens from the First Mirror Test and the lithium-beam diagnostic have been studied by spectrophotometry, ion beam analysis and electron microscopy. Test mirrors made of molybdenum were retrieved from the main chamber and the divertor after exposure to the 2013-2014 experimental campaign. In the main chamber, only mirrors located at the entrance of the carrier lost reflectivity (Be deposition from the eroded limiters), while those located deeper in the carrier were only slightly affected. The performance of mirrors in the JET divertor was strongly degraded by deposition of beryllium, tungsten and other species. Mirrors from the lithium-beam diagnostic have been studied for the first time. Gold coatings were severely damaged by intense arcing. As a consequence, material mixing of the gold layer with the stainless steel substrate occurred. Total reflectivity dropped from over 90% to less than 60 %, i.e. to the level typical for stainless steel.

Keyword
JET, First Mirror Test, diagnostic mirrors, erosion-deposition
National Category
Physical Sciences
Research subject
Physics
Identifiers
urn:nbn:se:kth:diva-190899 (URN)
External cooperation:
Note

QC 20160819

Available from: 2016-08-18 Created: 2016-08-18 Last updated: 2016-09-02Bibliographically approved
7. Impact of helium implantation and ion-induced damage on reflectivity of molybdenum mirrors
Open this publication in new window or tab >>Impact of helium implantation and ion-induced damage on reflectivity of molybdenum mirrors
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2016 (English)In: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584Article in journal (Refereed) Published
Abstract [en]

Molybdenum mirrors were irradiated with Mo and He ions to simulate the effect of neutron irradiation on diagnostic first mirrors in next-generation fusion devices. Up to 30 dpa were produced under molybdenum irradiation leading to a slight decrease of reflectivity in the near infrared range. After 3×1017 cm-2 of helium irradiation, reflectivity decreased by up to 20%. Combined irradiation by helium and molybdenum led to similar effects on reflectivity as irradiation with helium alone. Ion beam analysis showed that only 7% of the implanted helium was retained in the first 40nm layer of the mirror. The structure of the near-surface layer after irradiation was studied with scanning transmission electron microscopy and the extent and size distribution of helium bubbles was documented. The consequences of ion-induced damage on the performance of diagnostic components are discussed.

Place, publisher, year, edition, pages
Elsevier, 2016
Keyword
Diagnostic mirrors, Fusion, Helium, Ion-induced damage, Molybdenum, Fusion reactions, High resolution transmission electron microscopy, Infrared devices, Ion beams, Ions, Irradiation, Mirrors, Radiation, Reflection, Scanning electron microscopy, Transmission electron microscopy, Helium implantation, Helium irradiation, Implanted helium, Ion beam analysis, Near-infrared range, Near-surface layers, Scanning transmission electron microscopy, Neutron irradiation
National Category
Accelerator Physics and Instrumentation Fusion, Plasma and Space Physics Atom and Molecular Physics and Optics
Identifiers
urn:nbn:se:kth:diva-188336 (URN)10.1016/j.nimb.2016.02.065 (DOI)000382413500018 ()2-s2.0-84960914771 (ScopusID)
Note

QC 20160609

Available from: 2016-06-09 Created: 2016-06-09 Last updated: 2016-09-23Bibliographically approved

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