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Publications (10 of 206) 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, 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: 2018-01-15Bibliographically 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, 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, 559-563 p.Article 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: 2018-01-11Bibliographically 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, 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
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, 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, 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)2-s2.0-85019426928 (Scopus ID)
Note

QC 20171201

Available from: 2017-12-01 Created: 2017-12-01 Last updated: 2017-12-01Bibliographically 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, 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: 2018-01-12Bibliographically 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, 102-119 p.Article 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
Rubel, M., Moon, S., Petersson, P., Garcia Carrasco, A., Hallén, A., Krawczynska, A., . . . Widdowson, A. (2017). Metallic mirrors for plasma diagnosis in current and future reactors: tests for ITER and DEMO. Paper presented at 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications (PFMC), MAY, 2017, GERMANY. Physica Scripta, T170, Article ID 014061.
Open this publication in new window or tab >>Metallic mirrors for plasma diagnosis in current and future reactors: tests for ITER and DEMO
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2017 (English)In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, 014061Article in journal (Refereed) Published
Abstract [en]

Optical spectroscopy and imaging diagnostics in next-step fusion devices will rely on metallic mirrors. The performance of mirrors is studied in present-day tokamaks and in laboratory systems. This work deals with comprehensive tests of mirrors: (a) exposed in JET with the ITER-like wall (JET-ILW); (b) irradiated by hydrogen, helium and heavy ions to simulate transmutation effects and damage which may be induced by neutrons under reactor conditions. The emphasis has been on surface modification: deposited layers on JET mirrors from the divertor and on near-surface damage in ion-irradiated targets. Analyses performed with ion beams, microscopy and spectro-photometry techniques have revealed: (i) the formation of multiple co-deposited layers; (ii) flaking-off of the layers already in the tokamak, despite the small thickness (130-200 nm) of the granular deposits; (iii) deposition of dust particles (0.2-5 mu m, 300-400 mm(-2)) composed mainly of tungsten and nickel; (iv) that the stepwise irradiation of up to 30 dpa by heavy ions (Mo, Zr or Nb) caused only small changes in the optical performance, in some cases even improving reflectivity due to the removal of the surface oxide layer; (v) significant reflectivity degradation related to bubble formation caused by the irradiation with He and H ions.

Place, publisher, year, edition, pages
Institute of Physics Publishing (IOPP), 2017
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-220611 (URN)10.1088/1402-4896/aa8e27 (DOI)000417694700006 ()
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: 2018-01-12Bibliographically approved
Widdowson, A., Coad, J. P., Alves, E., Baron-Wiechec, A., Barradas, N. P., Brezinsek, S., . . . Rubel, M. (2017). Overview of fuel inventory in JET with the ITER-like wall. Nuclear Fusion, 57(8), Article ID 086045.
Open this publication in new window or tab >>Overview of fuel inventory in JET with the ITER-like wall
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2017 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 8, 086045Article in journal (Refereed) Published
Abstract [en]

Post mortem analyses of JET ITER-Like-Wall tiles and passive diagnostics have been completed after each of the first two campaigns (ILW-1 and ILW-2). They show that the global fuel inventory is still dominated by co-deposition; hence plasma parameters and sputtering processes affecting material migration influence the distribution of retained fuel. In particular, differences between results from the two campaigns may be attributed to a greater proportion of pulses run with strike points in the divertor corners, and having about 300 discharges in hydrogen at the end of ILW-2. Recessed and remote areas can contribute to fuel retention due to the larger areas involved, e.g. recessed main chamber walls, gaps in castellated Be main chamber tiles and material migration to remote divertor areas. The fuel retention and material migration due to the bulk W Tile 5 during ILW-1 are presented. Overall these tiles account for only a small percentage of the global accountancy for ILW-1.

Place, publisher, year, edition, pages
Institute of Physics Publishing (IOPP), 2017
Keyword
JET ITER-like wall, fuel retention, material migration
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-211596 (URN)10.1088/1741-4326/aa7475 (DOI)000405654600002 ()2-s2.0-85019476814 (Scopus ID)
Note

QC 20170815

Available from: 2017-08-15 Created: 2017-08-15 Last updated: 2017-08-15Bibliographically approved
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ORCID iD: ORCID iD iconorcid.org/0000-0001-9901-6296

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