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Publications (10 of 208) Show all publications
Brezinsek, S., Wirtz, M., Dorrow-Gesprach, D., Loewenhoff, T. & Rubel, M. (2017). 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications. Physica Scripta, T170, Article ID 010201.
Open this publication in new window or tab >>16th International Conference on Plasma-Facing Materials and Components for Fusion Applications
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2017 (English)In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 010201Article in journal (Refereed) Published
Place, publisher, year, edition, pages
Institute of Physics (IOP), 2017
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-220608 (URN)10.1088/1402-4896/aa9958 (DOI)000417694700001 ()
Note

QC 20180115

Available from: 2018-01-15 Created: 2018-01-15 Last updated: 2018-02-21Bibliographically approved
Masuzakil, S., Tokitanii, M., Otsuka, T., Oya, Y., Hatan, Y., Miyamoto, M., . . . Rubel, M. (2017). Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall. Paper presented at 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications, MAY 16-19, 2017, GERMANY. Physica Scripta, T170, Article ID 014031.
Open this publication in new window or tab >>Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall
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2017 (English)In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014031Article in journal (Refereed) Published
Abstract [en]

Results of the comprehensive surface analyses of divertor tiles and dusts retrieved from JET after the first ITER-like wall campaign (2011-2012) are presented. The samples cored from the divertor tiles were analyzed. Numerous nano-size bubble-like structures were observed in the deposition layer on the apron of the inner divertor tile, and a beryllium dust with the same structures were found in the matter collected from the inner divertor after the campaign. This suggests that the nano-size bubble-like structures can make the deposition layer to become brittle and may lead to cracking followed by dust generation. X-ray photoelectron spectroscopy analyses of chemical states of species in the deposition layers identified the formation of beryllium-tungsten intermetallic compounds on an inner vertical tile. Different tritium retention profiles along the divertor tiles were observed at the top surfaces and at deeper regions of the tiles by using the imaging plate technique.

Place, publisher, year, edition, pages
IOP PUBLISHING LTD, 2017
Keyword
JET, ITER-like wall, dust, XPS, divertor tiles, tritium retention, microstructure
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-217723 (URN)10.1088/1402-4896/aa8bcc (DOI)000414120500031 ()
Conference
16th International Conference on Plasma-Facing Materials and Components for Fusion Applications, MAY 16-19, 2017, GERMANY
Note

QC 20171123

Available from: 2017-11-23 Created: 2017-11-23 Last updated: 2017-11-23Bibliographically approved
Catarino, N., Barradas, N. P., Corregidor, V., Widdowson, A., Baron-Wiechec, A., Coad, J. P., . . . Alves, E. (2017). Assessment of erosion, deposition and fuel retention in the JET-ILW divertor from ion beam analysis data. NUCLEAR MATERIALS AND ENERGY, 12, 559-563
Open this publication in new window or tab >>Assessment of erosion, deposition and fuel retention in the JET-ILW divertor from ion beam analysis data
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2017 (English)In: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, p. 559-563Article in journal (Refereed) Published
Abstract [en]

Post-mortem analyses of individual components provide relevant information on plasma-surface interactions like tungsten erosion, beryllium deposition and plasma fuel retention with divertor tiles via implantation or co-deposition. Ion Beam techniques are ideal tools for such purposes and have been extensively used for post-mortem analyses of selected tiles from JET following each campaign. In this contribution results from tiles removed from the JET ITER-Like Wall (JET-ILW) divertor following the 2013-2014 campaign are presented. The results summarize erosion, deposition and fuel retention along the poloidal cross section of the divertor surface and provide data for comparison with the first JET-ILW campaign, showing a similar pattern of material migration with the exception of Tile 6 where the strike point time on the tile was similar to 4 times longer in 2013-2014 than in 2011-2012, which is likely to account for more material migration to this region. The W deposition on top of the Mo marker coating of Tile 4 shows that the enrichment takes place at the strike point location.

Place, publisher, year, edition, pages
Elsevier, 2017
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-220641 (URN)10.1016/j.nme.2016.10.027 (DOI)000417293300088 ()2-s2.0-85006922988 (Scopus ID)
Note

QC 20180111

Available from: 2018-01-11 Created: 2018-01-11 Last updated: 2018-03-12Bibliographically approved
Tsavalas, P., Lagoyannis, A., Mergia, K., Rubel, M., Triantou, K., Harissopulos, S., . . . Petersson, P. (2017). Be ITER-like wall at the JET tokamak under plasma. Paper presented at 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications, MAY 16-19, 2017, GERMANY. Physica Scripta, T170, Article ID 014049.
Open this publication in new window or tab >>Be ITER-like wall at the JET tokamak under plasma
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2017 (English)In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014049Article in journal (Refereed) Published
Abstract [en]

The JET tokamak is operated with beryllium and tungsten plasma-facing components to prepare for the exploitation of ITER. To determine beryllium erosion and migration in JET a set of markers were installed. Specimens from different beryllium marker tiles of the main wall of the ITER-like wall (ILW) JET tokamak from the first and the second D-D campaign were analyzed with nuclear reaction analysis, x-ray fluorescence spectroscopy, scanning electron microscopy and x-ray diffraction (XRD). Emphasis was on the determination of carbon plasma impurities deposited on beryllium surfaces. The C-12(d, p(0))C-13 reaction was used to quantify carbon deposition and to determine depth profiles. Carbon quantities on the surface of the Be tiles are low, varying from (0.35 +/- 0.07) x 10(17) to (11.8 +/- 0.6) x 10(17) at cm(-2) in the deposition depth from 0.4 to 6.7 mu m, respectively. In the 0.4-0.5 mm wide grooves of castellation sides the carbon content is found up to (14.3 +/- 2.5) x 10(17) at cm(-2) while it is higher (up to (38 +/- 4) x 10(17) at cm(-2)) in wider gaps (0.8 mm) separating tile segments. Oxygen (O), titanium (Ti), chromium (Cr), manganese (Mn), iron (Fe), nickel (Ni) and tungsten (W) were detected in all samples exposed to plasma and the reference one but at lower quantities at the latter. In the central part of the Inner Wall Guard Limiter from the first ILW campaign and in the Outer Poloidal Limiter from the second ILW campaign the Ni interlayer has been completely eroded. XRD shows the formation of BeNi in most specimens.

Place, publisher, year, edition, pages
IOP PUBLISHING LTD, 2017
Keyword
beryllium, JET tokamak, ITER like wall, plasma, nuclear reaction analysis, erosiond-eposition
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-217725 (URN)10.1088/1402-4896/aa8ff4 (DOI)000414120500049 ()
Conference
16th International Conference on Plasma-Facing Materials and Components for Fusion Applications, MAY 16-19, 2017, GERMANY
Note

QC 20171123

Available from: 2017-11-23 Created: 2017-11-23 Last updated: 2017-11-23Bibliographically approved
Louche, F., Wauters, T., Ragona, R., Moeller, S., Durodie, F., Litnovsky, A., . . . Van Schoor, M. (2017). Design of an ICRF system for plasma-wall interactions and RF plasma production studies on TOMAS. Paper presented at 29th Symposium on Fusion Technology (SOFT), SEP 05-09, 2016, Prague, CZECH REPUBLIC. Fusion engineering and design, 123, 317-320
Open this publication in new window or tab >>Design of an ICRF system for plasma-wall interactions and RF plasma production studies on TOMAS
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2017 (English)In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 123, p. 317-320Article in journal (Refereed) Published
Abstract [en]

Ion cyclotron wall conditioning (ICWC) is being developed for ITER and W7-X as a baseline conditioning technique in which the ion cyclotron heating and current drive system will be employed to produce and sustain the currentless conditioning plasma. The TOMAS project (TOroidal MAgnetized System, operated at the FZ-juelich, Germany) proposes to explore several key aspects of ICWC. For this purpose we have designed an ICRF system made of a single strap antenna within a metallic box, connected to a feeding port and a pre-matching system. We discuss the design work of the antenna system with the help of the commercial electromagnetic software CST Microwave Studio (R). The simulation results for a given geometry provide input impedance matrices for the two-port system. These matrices are afterwards inserted into various circuit models to assess the accessibility of the required frequency range. The sensitivity of the matching system to uncertainties on plasma loading and capacitance values is notably addressed. With a choice of three variable capacitors we show that the system can cope with such uncertainties. We also demonstrate that the system can cope as well with the high reflected power levels during the short breakdown phase of the RF discharge, but at the cost of a significantly reduced coupled power.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2017
Keyword
ICWC, Antennas, Capacitors, TOMAS, Matching, Simulations
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-221890 (URN)10.1016/j.fusengdes.2017.04.123 (DOI)000418992000063 ()
Conference
29th Symposium on Fusion Technology (SOFT), SEP 05-09, 2016, Prague, CZECH REPUBLIC
Note

QC 20180130

Available from: 2018-01-30 Created: 2018-01-30 Last updated: 2018-01-30Bibliographically approved
Vizvary, Z., Bourdel, B., Garcia Carrasco, A., Lam, N., Leipold, F., Pitts, R. A., . . . Widdowson, A. (2017). Engineering design and analysis of an ITER-like first mirror test assembly on JET. Paper presented at 29th Symposium on Fusion Technology (SOFT), SEP 05-09, 2016, Prague, CZECH REPUBLIC. Fusion engineering and design, 123, 1054-1057
Open this publication in new window or tab >>Engineering design and analysis of an ITER-like first mirror test assembly on JET
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2017 (English)In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 123, p. 1054-1057Article in journal (Refereed) Published
Abstract [en]

The ITER first mirrors are the components of optical diagnostic systems closest to the plasma. Deposition may build up on the surfaces of the mirror affecting their ability to fulfil their function. However, physics modelling of this layer growth is fraught with uncertainty. A new experiment is underway on JET, under contract to ITER, with primary objective to test if, under realistic plasma and wall material conditions and with ITER-like first mirror aperture geometry, deposits do grow on first mirrors. This paper describes the engineering design and analysis of this mirror test assembly. The assembly was installed in the 2014-15 shutdown and will be removed in the 2016-17 shutdown.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2017
Keyword
ITER-like first mirror, JET, Additive manufacturing, Remote handling, Disruption loads
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-221891 (URN)10.1016/j.fusengdes.2016.12.016 (DOI)000418992000221 ()
Conference
29th Symposium on Fusion Technology (SOFT), SEP 05-09, 2016, Prague, CZECH REPUBLIC
Note

QC 20180131

Available from: 2018-01-31 Created: 2018-01-31 Last updated: 2018-01-31Bibliographically approved
Fortuna-Zalesna, E., Grzonka, J., Moon, S., Rubel, M., Petersson, P. & Widdowson, A. (2017). Fine metal dust particles on the wall probes from JET-ILW. Paper presented at 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications, MAY 16-19, 2017, GERMANY. Physica Scripta, T170, Article ID 014038.
Open this publication in new window or tab >>Fine metal dust particles on the wall probes from JET-ILW
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2017 (English)In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014038Article in journal (Refereed) Published
Abstract [en]

Collection and ex situ studies of dust generated in controlled fusion devices during plasma operation are regularly carried out after experimental campaigns. Herewith results of the dust survey performed in JET after the second phase of operation with the metal ITER-like wall (2013-2014) are presented. For the first-time-ever particles deposited on silicon plates acting as dust collectors installed in the inner and outer divertor have been examined. The emphasis is on analysing metal particles (Be and W) with the aim to determine their composition, size and surface topography. The most important is the identification of beryllium dust in the form of droplets (both splashes and spherical particles), flakes of co-deposits and small fragments of Be tiles. Tungsten and nickel rich (from Inconel) particles are also identified. Nitrogen from plasma edge cooling has been detected in all types of particles. They are categorized and the origin of various constituents is discussed.

Place, publisher, year, edition, pages
IOP PUBLISHING LTD, 2017
Keyword
plasma, dust, JET tokamak, ITER-like wall, beryllium, tungsten
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-217721 (URN)10.1088/1402-4896/aa8ddf (DOI)000414120500038 ()
Conference
16th International Conference on Plasma-Facing Materials and Components for Fusion Applications, MAY 16-19, 2017, GERMANY
Note

QC 20171123

Available from: 2017-11-23 Created: 2017-11-23 Last updated: 2017-11-23Bibliographically approved
Rubel, M., Petersson, P., Zhou, Y., Coad, J. P., Lungu, C., Jepu, I., . . . Alves, E. (2017). Fuel inventory and deposition in castellated structures in JET-ILW. Nuclear Fusion, 57(6), Article ID 066027.
Open this publication in new window or tab >>Fuel inventory and deposition in castellated structures in JET-ILW
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2017 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 6, article id 066027Article in journal (Refereed) Published
Abstract [en]

Since 2011 the JET tokamak has been operated with a metal ITER-like wall (JET-ILW) including castellated beryllium limiters and lamellae-type bulk tungsten tiles in the divertor. This has allowed for a large scale test of castellated plasma-facing components (PFC). Procedures for sectioning the limiters into single blocks of castellation have been developed. This facilitated morphology studies of morphology of surfaces inside the grooves for limiters after experimental campaigns 2011-2012 and 2013-2014. The deposition in the 0.4-0.5 mm wide grooves of the castellation is 'shallow'. It reaches 1-2 mm into the 12 mm deep gap. Deuterium concentrations are small (mostly below 1 × 1018 cm-2). The estimated total amount of deuterium in all the castellated limiters does not exceed the inventory of the plasma-facing surfaces (PFS) of the limiters. There are only traces of Ni, Cr and Fe deposited in the castellation gaps. The same applies to the carbon content. Also low deposition of D, Be and C has been measured on the sides of the bulk tungsten lamellae pieces. Modelling clearly reflects: (a) a sharp decrease in the measured deposition profiles and(b) an increase in deposition with the gap width. Both experimental and modelling data give a strong indication and information to ITER that narrow gaps in the castellated PFC are essential. X-ray diffraction on PFS has clearly shown two distinct composition patterns: Be with an admixture of Be-W intermetallic compounds (e.g. Be22W) in the deposition zone, whilst only pure Be has been detected in the erosion zone. The lack of compound formation in the erosion zone indicates that no distinct changes in the thermo-mechanical properties of the Be PFC might be expected.

Place, publisher, year, edition, pages
Institute of Physics Publishing, 2017
Keyword
beryllium limiters, castellation, deposition, fuel inventory, ITER-like wall, JET, Beryllium, Carbon, Deuterium, Erosion, Facings, Fighter aircraft, Fusion reactor divertors, Intermetallics, Jets, Magnetoplasma, Tungsten, Tungsten compounds, X ray diffraction, Experimental campaign, Fuel inventories, Morphology of surfaces, Plasma facing surfaces, Plasma-facing components, Thermomechanical properties, Tokamak devices
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-216517 (URN)10.1088/1741-4326/aa6864 (DOI)000425870600001 ()2-s2.0-85019426928 (Scopus ID)
Note

QC 20171201

Available from: 2017-12-01 Created: 2017-12-01 Last updated: 2018-03-09Bibliographically approved
Widdowson, A., Coad, J. P., Alves, E., Baron-Wiechec, A., Barradas, N. P., Catarino, N., . . . Rubel, M. (2017). Impurity re-distribution in the corner regions of the JET divertor. Paper presented at 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications (PFMC), MAY, 2017, GERMANY. Physica Scripta, T170, Article ID 014060.
Open this publication in new window or tab >>Impurity re-distribution in the corner regions of the JET divertor
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2017 (English)In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014060Article in journal (Refereed) Published
Abstract [en]

The International Thermonuclear Experimental Reactor (ITER) will use a mixture of deuterium (D) and tritium (T) as the fuel to generate power. Since T is both radioactive and expensive the Joint European Torus (JET) has been at the forefront of research to discover how much T is used and where it may be retained within the main reaction chamber. Until the year 2010 the JET plasma facing components were constructed of carbon fibre composites. During the JET carbon (C) phases impurities accumulated at the corners of the divertor located towards the bottom of the chamber in regions shadowed from the plasma where they are very difficult to reach and remove. This build-up of C and the associated H-isotope (including T) retention were of particular concern for future fusion reactors therefore, in 2010 JET changed the wall protection to (mainly) Be and the divertor to tungsten (W)-the JET ITER-like wall (ILW)-the choice of materials for ITER. This paper reveals that with the JET ILW impurities are still accumulating in the shadowed regions, with Be being the majority element, though the overall quantities are very much reduced from those in the C phases. Material will be transported into the shadowed regions principally when the plasma strike points are on the corner tiles, but particles typically have about a 75% probability of reflection from line-of sight surfaces, and multiple reflection/scattering results in deposition over all surfaces.

Place, publisher, year, edition, pages
Institute of Physics (IOP), 2017
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-220612 (URN)10.1088/1402-4896/aa90d5 (DOI)000417694700005 ()
Conference
16th International Conference on Plasma-Facing Materials and Components for Fusion Applications (PFMC), MAY, 2017, GERMANY
Note

QC 20180112

Available from: 2018-01-12 Created: 2018-01-12 Last updated: 2018-02-26Bibliographically approved
Matejicek, J., Weinzettl, V., Mackova, A., Malinsky, P., Havranek, V., Naydenkova, D., . . . Tervakangas, S. (2017). Interaction of candidate plasma facing materials with tokamak plasma in COMPASS. Journal of Nuclear Materials, 493, 102-119
Open this publication in new window or tab >>Interaction of candidate plasma facing materials with tokamak plasma in COMPASS
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2017 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 493, p. 102-119Article in journal (Refereed) Published
Abstract [en]

The interaction of tokamak plasma with several materials considered for the plasma facing components of future fusion devices was studied in a small-size COMPASS tokamak. These included mainly tungsten as the prime candidate and chromium steel as an alternative whose suitability was to be assessed. For the experiments, thin coatings of tungsten, P92 steel and nickel on graphite substrates were prepared by arc-discharge sputtering. The samples were exposed to hydrogen and deuterium plasma discharges in the COMPASS tokamak in two modes: a) short exposure (several discharges) on a manipulator in the proximity of the separatrix, close to the central column, and b) long exposure (several months) at the central column, aligned with the other graphite tiles. During the discharges, standard plasma diagnostics were used and a local emission of spectral lines in the visible near ultraviolet regions, corresponding to the material erosion, was monitored. Before and after the plasma exposures, the sample surfaces were observed using scanning electron microscopy, the coatings thickness was measured using Rutherford backscattering spectroscopy, and the concentration profiles of hydrogen and deuterium were measured by elastic recoil detection analysis. The uniformity of the coatings and their thickness was verified before the exposure. After the exposure, no reduction of the thickness was observed, indicating the absence of 'global' erosion. Erosion was observed only in isolated spots, and attributed to unipolar arcing. Slightly larger erosion was found on the steel coatings compared to the tungsten ones. Incorporation of deuterium in a thin surface layer was observed, in dependence on the exposure mode. Additionally, boron enrichment of the long-exposure samples was observed, as a result of the tokamak chamber boronization.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE BV, 2017
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-214323 (URN)10.1016/j.jnucmat.2017.06.009 (DOI)000408044000013 ()2-s2.0-85020041707 (Scopus ID)
Note

QC 20170914

Available from: 2017-09-14 Created: 2017-09-14 Last updated: 2017-09-14Bibliographically approved
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ORCID iD: ORCID iD iconorcid.org/0000-0001-9901-6296

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