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  • 801.
    Weckmann, Armin
    et al.
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Petersson, Per
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Fortuna-Zalesna, Elzbieta
    Zielinski, Witold
    Romelczyk, Barbara
    Grigore, Eduard
    Ruset, Cristian
    Kreter, Arkadi
    Ageing of structural materials in tokamaks: TEXTOR liner study2017In: Physica scripta. T, ISSN 0281-1847, Vol. T170, article id 014053Article in journal (Refereed)
    Abstract [en]

    After the final shut-down of the tokamak TEXTOR, all of its machine parts became accessible for comprehensive studies. This unique chance enabled the study of the Inconel 625 liner by a wide range of methods. The aim was to evaluate eventual alteration of surface and bulk characteristics from recessed wall elements that may influence the machine performance. The surface was covered with stratified layers consisting mainly of boron, carbon, oxygen, and in some cases also silicon. Wall conditioning and limiter materials hence predominantly define deposition on the liner. Deposited layers on recessed wall elements reach micrometre thickness within decades, peel off and may contribute to the dust inventory in tokamaks. Fuel retention was about 4 at% of the overall layers, with no evidence for diffusion into the Inconel substrate. Inconel 625 retained its mechanical strength despite 26 years of cyclic heating, stresses and particle bombardment.

  • 802.
    Weckmann, Armin
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Ström, Petter
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Kurki-Suonio, T.
    Aalto Univ, Dept Appl Phys, Aalto 00076, Finland..
    Sarkimaki, K.
    Aalto Univ, Dept Appl Phys, Aalto 00076, Finland..
    Kirschner, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Kreter, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Brezinsek, S.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Romazanov, J.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Wienhold, P.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Pospieszczyk, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Hakola, A.
    VTT Tech Res Ctr Finland Ltd, Espoo 02044, Finland..
    Airila, M.
    VTT Tech Res Ctr Finland Ltd, Espoo 02044, Finland..
    Review on global migration, fuel retention and modelling after TEXTOR decommission2018In: NUCLEAR MATERIALS AND ENERGY, Vol. 17, p. 83-112Article, review/survey (Refereed)
    Abstract [en]

    Before decommissioning of the TEXTOR tokamak in 2013, the machine was conditioned with a comprehensive migration experiment where MoF6 and N-15(2) were injected on the very last operation day. Thereafter, all plasmafacing components (PFCs) were available for extensive studies of both local and global migration of impurities - Mo, W, Inconel alloy constituents, 15 N, F - and fuel retention studies. Measurements were performed on 140 limiter tiles out of 864 throughout the whole machine to map global transport. One fifth of the introduced molybdenum could be found. Wherever possible, the findings are compared to results obtained previously in other machines. This review incorporates both published and unpublished results from this TEXTOR study and combines findings with analytical methods as well as modelling results from two codes, ERO and ASCOT. The main findings are: Both local and global molybdenum transport can be explained by toroidal plasma flow and (sic) x (sic) drift. The suggested transport scheme for molybdenum holds also for other analysed species, namely tungsten from previous experiments and medium-Z metals (Cr-Cu) introduced on various occasions. Analytical interpretation of several deposition profile features is possible with basic geometrical and plasma physics considerations. These are deposition profiles on the collector probe, the lower part of the inner bumper limiter, the poloidal cross-section of the inner bumper limiter, and the poloidal limiter. Any deposition pattern found in this TEXTOR study, including fuel retention, has neither poloidal nor toroidal symmetry, which is often assumed when determining deposition profiles on global scale. Fuel retention is highly inhomogeneous due to local variation of plasma parameters - by auxiliary heating systems and impurity injection - and PFC temperature. Local modelling with ERO yields good qualitative agreement but too high local deposition efficiency. Global modelling with ASCOT shows that the radial electric field and source form have a high impact on global deposition patterns, while toroidal flow has little influence. Some of the experimental findings could be reproduced. Still, qualitative differences between simulated and experimental global deposition patterns remain. The review closes with lessons learnt during this extensive TEXTOR study which might be helpful for future scientific exploitation of other tokamaks to be decommissioned.

  • 803.
    Weckmann, Armin
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Wienhold, P.
    Brezinsek, S.
    Coenen, J. W.
    Kischner, A.
    Kreter, A.
    Pospieszczyk, A.
    Local Migration Studies of High-Z Metals in the TEXTOR Tokamak2016In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T167, article id 014058Article in journal (Refereed)
    Abstract [en]

    Volatile compounds of tungsten (WF6) and molybdenum (MoF6) were used as tracers of high-Z metal migration in the TEXTOR tokamak in several gas injection experiments when puffing was done through a test limiter. The experiments with W were performed prior major shut-downs, while the MoF6 was followed by the final shutdown in connection with TEXTOR decommissioning. In all cases a set of various surface probes and limiter tiles were retrieved and analysed with electron and ion beam techniques. The focus was on the local deposition in the vicinity of the gas inlet and in the inlet system. Depth profiles in the deposits and metal distribution maps clearly shown that only near the gas inlet significant amounts of Mo are deposited along the scrape-off layer flow and E×B drift directions, which could be reproduced by ERO-code modelling. Correlation between the plasma operation scenario and the deposition patterns is presented.

  • 804. Widdowson, A.
    et al.
    Alves, E.
    Ayres, C. F.
    Baron-Wiechec, A.
    Brezinsek, S.
    Catarino, N.
    Coad, J. P.
    Heinola, K.
    Likonen, J.
    Matthews, G. F.
    Mayer, M.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Material migration patterns and overview of first surface analysis of the JET ITER-like wall2014In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T159, p. 014010-Article in journal (Refereed)
    Abstract [en]

    Following the first JET ITER-like wall operations a detailed in situ photographic survey of the main chamber and divertor was completed. In addition, a selection of tiles and passive diagnostics were removed from the vessel and made available for post mortem analysis. From the photographic survey and results from initial analysis, the first conclusions regarding erosion, deposition, fuel retention and material transport during divertor and limiter phases have been drawn. The rate of deposition on inner and outer base divertor tiles and remote divertor corners was more than an order of magnitude less than during the preceding carbon wall operations, as was the concomitant deuterium retention. There was however beryllium deposition at the top of the inner divertor. The net beryllium erosion rate from the mid-plane inner limiters was found to be higher than for the previous carbon wall campaign although further analysis is required to determine the overall material balance due to erosion and re-deposition.

  • 805. Widdowson, A.
    et al.
    Alves, E.
    Baron-Wiechec, A.
    Barradas, N. P.
    Catarino, N.
    Coad, J. P.
    Corregidor, V.
    Garcia-Carrasco, A.
    Heinola, K.
    Koivuranta, S.
    Krat, S.
    Lahtinen, A.
    Likonen, J.
    Mayer, M.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Van Boxel, S.
    Overview of the JET ITER-like wall divertor2017In: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, p. 499-505Article in journal (Refereed)
    Abstract [en]

    The work presented draws on new analysis of components removed following the second JET ITER-like wall campaign 2013-14 concentrating on the upper inner divertor, inner and outer divertor corners, lifetime issues relating to tungsten coatings on JET carbon fibre composite divertor tiles and dust/particulate generation. The results show that the upper inner divertor remains the region of highest deposition in the JET-ILW. Variations in plasma configurations between the first and second campaign have altered material migration to the corners of the inner and outer divertor. Net deposition is shown to be beneficial in the sense that it reduces W coating erosion, covers small areas of exposed carbon surfaces and even encapsulates particles.

  • 806. Widdowson, A.
    et al.
    Ayres, C.F.
    Booth, S.
    Coad, J. P.
    Hakola, A.
    Heinola, K.
    Ivanova, Darya
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Koivuranta, S.
    Likonen, J.
    Mayer, M.
    Stamp, M.
    Comparison of JET main chamber erosion with dust collected in the divertor2013In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 436, no Suppl., p. 827-832Article in journal (Refereed)
    Abstract [en]

    A complete global balance for carbon in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust and flakes collecting in the divertor region. This paper describes a number of measurements on aspects of this global picture. Profiler measurements and cross section microscopy on tiles that were removed in the 2009 JET intervention are used to evaluate the net erosion in the main chamber and net deposition in the divertor. In addition the mass of dust and flakes collected from the JET divertor during the same intervention is also reported and included as part of the balance. Spectroscopic measurements of carbon erosion from the main chamber are presented and compared with the erosion measurements for the main chamber.

  • 807. Widdowson, A.
    et al.
    Baldwin, M. J.
    Coad, J. P.
    Doerner, R. P.
    Hanna, J.
    Hole, D. E.
    Matthews, G. F.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Seraydarian, R.
    Xu, H.
    Testing of beryllium marker coatings in PISCES-B for the JET ITER-like wall2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 390-91, p. 988-991Article in journal (Refereed)
    Abstract [en]

    Beryllium has been chosen as the first wall material for ITER. In order to understand the issues of material migration and tritium retention associated with the use of beryllium, a largely beryllium first wall will be installed in JET. As part of the JET ITER-like wall, beryllium tiles with marker coatings are proposed as a diagnostic tool for studying the erosion and deposition of beryllium around the vessel. The nominal structure for these coatings is a similar to 10 mu m beryllium surface layer separated from the beryllium tile by a 2-3 mu m metallic inter-layer. Two types of coatings are tested here; one with a nickel inter-layer anti one with a copper/beryllium mixed inter-layer. The coating samples were deposited by DC magnetron Sputtering at General Atomics and were exposed to deuterium plasma in PISCES-B. The results of this testing show that the beryllium/nickel marker coating would be suitable for installation in JET.

  • 808. Widdowson, A.
    et al.
    Baron-Wiechec, A.
    Batistoni, P.
    Belonohy, E.
    Coad, J. P.
    Dinca, P.
    Flammini, D.
    Fox, F.
    Heinola, K.
    Jepu, I.
    Likonen, J.
    Lilley, S.
    Lungu, C. P.
    Matthews, G. F.
    Naish, J.
    Pompilian, O.
    Porosnicu, C.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Villari, R.
    Experience of handling beryllium, tritium and activated components from JET ITER like wall2016In: Physica Scripta, Institute of Physics Publishing (IOPP), 2016, no T167, article id 014057Conference paper (Refereed)
    Abstract [en]

    JET components are removed periodically for surface analysis to assess material migration and fuel retention. This paper describes issues related to handling JET components and procedures for preparing samples for analysis; in particular a newly developed procedure for cutting beryllium tiles is presented. Consideration is also given to the hazards likely due to increased tritium inventory and material activation from 14 MeV neutrons following the planned TT and DT operations (DTE2) in 2017. Conclusions are drawn as to the feasibility of handling components from JET post DTE2.

  • 809. Widdowson, A.
    et al.
    Brezinsek, S.
    Coad, J. P.
    Hole, D. E.
    Likonen, J.
    Philipps, V.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Stamp, M. F.
    An overview of erosion-deposition studies for the JET Mk II high delta divertor2009In: Physica scripta. T, ISSN 0281-1847, Vol. T138, p. 014005-Article in journal (Refereed)
    Abstract [en]

    Post-mortem analyses of tiles removed from the JET MkII HD divertor in 2007 are presented. The results indicate an increase in deposition at the outer plasma-shadowed region of the divertor, not seen prior to 2004 and indicate a shift away from the asymmetric picture of net deposition at the inner divertor compared with no overall deposition or erosion at the outer divertor. Surface analysis of the inner and outer vertical divertor tiles is largely the same as observed previously; however, a notable increase in Be composition on the inner and outer floor tiles is observed. An attempt has been made to correlate these data with campaign-averaged plasma configurations and spectroscopy results. While some changes in deposition/erosion characteristics can be explained, further detailed analysis of diagnostics and especially of time-resolved data, such as from rotating collector and quartz microbalance diagnostics, is required.

  • 810. Widdowson, A.
    et al.
    Coad, J. P.
    Alves, E.
    Baron-Wiechec, A.
    Barradas, N. P.
    Brezinsek, S.
    Catarino, N.
    Corregidor, V.
    Heinola, K.
    Koivuranta, S.
    Krat, S.
    Lahtinen, A.
    Likonen, J.
    Matthews, G. F.
    Mayer, M.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. EUROfus Consortium JET, England.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. EUROfus Consortium JET, England.
    Overview of fuel inventory in JET with the ITER-like wall2017In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 8, article id 086045Article in journal (Refereed)
    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.

  • 811. Widdowson, A.
    et al.
    Coad, J. P.
    Alves, E.
    Baron-Wiechec, A.
    Barradas, N. P.
    Catarino, N.
    Corregidor, V.
    Heinola, K.
    Krat, S.
    Likonen, J.
    Matthews, G. F.
    Mayer, M.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Rubel, Marek
    KTH.
    Impurity re-distribution in the corner regions of the JET divertor2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014060Article in journal (Refereed)
    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.

  • 812. Widdowson, A.
    et al.
    Coad, J. P.
    Bekris, N.
    Counsell, G.
    Forrest, M. J.
    Gibson, K. J.
    Hole, D.
    Likonen, J.
    Parsons, W.
    Renvall, T.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Efficacy of photon cleaning of JET divertor tiles2007In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 363, p. 341-345Article in journal (Refereed)
    Abstract [en]

    Photon cleaning by means of a flash-lamp was used for in-situ detritiation of the inner wall tiles of the JET divertor in May 2004. Additional trials were also performed ex-situ in October 2004 on divertor base tiles. Early work confirmed that for pulse energies between 150 J and 300 J some deposited material was removed. To increase the amount of material removed during photon cleaning, further experiments with higher pulse energies (500 J) were performed and are reported here. Analysis of cross sections confirmed a removal rate of 0.04 mu m/pulse, removing similar to 80 mu m from 200 mu m thick deposits over a treatment area of 15 x 10(-4) m(2). During the photon cleaning tests at least 12% of the tritium inventory for the tile was removed. It was also shown that deuterium was desorbed from a depth similar to 7 mu m beyond the depth of material removed. Crown

  • 813.
    Widdowson, A.
    et al.
    EURATOM/CCFE Fusion Association, Culham Science Centre, UK.
    Coad, J. P.
    EURATOM/CCFE Fusion Association, Culham Science Centre, UK.
    de Temmerman, G.
    FOM Insititute for Plasma Physics Rijnhuizen, The Netherlands.
    Farcage, D.
    CEA Saclay, DEN/DANS/DPC/SCP/LILM, Farnce.
    Hole, D.
    Dept. of Engineering and Design, University of Sussex, UK.
    Ivanova, Darya
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Leontyev, A.
    CEA Saclay, DEN/DANS/DPC/SCP/LILM, Farnce.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Semerok, A.
    CEA Saclay, DEN/DANS/DPC/SCP/LILM, Farnce.
    Schmidt, A.
    IEF-Plasmaphysik, Forschungszentrum fülich, Association EURATOM-FZJ, Germany.
    Thro, P.-Y.
    CEA Saclay, DEN/DANS/DPC/SCP/LILM, Farnce.
    Removal of beryllium-containing films deposited in JET from mirror surfaces by laser cleaning2011In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, ISSN 0022-3115, Vol. 415, no 1, p. S1199-S1202Article in journal (Refereed)
    Abstract [en]

    A set of stainless steel (SS) and molybdenum mirror samples located in the divertor and at the outer mid-plane of the vessel were exposed in JET from 2005 to 2007. A selection of these mirror samples with well adhered deposits (i.e. not flaking) of up to a few hundred nanometers in thickness and with Be/C ratios ranging from 0 to similar to 1 have been cleaned using a laser system developed at CEA, Saclay. Following laser cleaning the recovered reflectivity was generally better in the infrared than the visible spectrum, with recovery of up to 90% of the initial reflectivity being obtained at 1600 nm for both Mo and SS mirrors falling as low as 20-30% of initial reflectivity at a wavelength of 400 nm for some SS mirrors, rising to similar to 80% for Mo mirrors. Some deposit remained on the mirrors after the cleaning trials.

  • 814. Wienhold, P.
    et al.
    Litnovsky, A.
    Philipps, V.
    Schweer, B.
    Sergienko, G.
    Oelhafen, P.
    Ley, M.
    De Temmerman, G.
    Schneider, W.
    Hildebrandt, D.
    Laux, M.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Emmoth, Birger
    KTH, School of Information and Communication Technology (ICT), Microelectronics and Information Technology, IMIT.
    Exposure of metal mirrors in the scrape-off layer of TEXTOR2005In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 337-39, no 03-jan, p. 1116-1120Article in journal (Refereed)
    Abstract [en]

    Large molybdenum mirrors have been exposed in the SOL of TEXTOR in order to simulate conditions relevant for ITER optical components. Distortions of the reflectivity - increase as well as decrease - are found in the erosion and deposition dominated areas, respectively. The changes are most pronounced in the near UV and level off in the IR and can partly be attributed to observed surface changes. A novel periscope system was installed and mirrors exposed in a pilot experiment to simulate the transmission of light to distant sensors in ITER.

  • 815. Wiesen, S
    et al.
    Brezinsek, M
    Wischmeier, M
    de la Luna, E
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Groth, M
    Järvinen, A
    Losada, U
    Martin, A
    Impact of the JET ITER-Like Wall on H-Mode Plasma Fueling2016In: 26th IAEA Fusion Energy Conference, 17-22 October 2016, 2016Conference paper (Refereed)
  • 816.
    Wiesen, S.
    et al.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Brezinsek, S.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Bonnin, X.
    ITER Org, Route Vinon Sur Verdon,CS90 046, F-13067 St Paul Les Durance, France..
    Delabie, E.
    Oak Ridge Natl Lab, POB 2009, Oak Ridge, TN 37831 USA..
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Groth, M.
    Aalto Univ, Espoo, Finland..
    Guillemaut, C.
    Univ Lisbon, Inst Plasmas & Fusao Nucl, IST, P-1049001 Lisbon, Portugal..
    Harrison, J.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Harting, D.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Henderson, S.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Huber, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Kruezi, U.
    ITER Org, Route Vinon Sur Verdon,CS90 046, F-13067 St Paul Les Durance, France..
    Pitts, R. A.
    ITER Org, Route Vinon Sur Verdon,CS90 046, F-13067 St Paul Les Durance, France..
    Wischmeier, M.
    Max Planck Inst Plasma Phys, D-85748 Garching, Germany..
    On the role of finite grid extent in SOLPS-ITER edge plasma simulations for JET H-mode discharges with metallic wall2018In: Nuclear Materials and Energy, E-ISSN 2352-1791, Vol. 17, p. 174-181Article in journal (Refereed)
    Abstract [en]

    The impact of the finite grid size in SOLPS-ITER edge plasma simulations is assessed for JET H-mode discharges with a metal wall. For a semi-horizontal divertor configuration it is shown that the separatrix density is at least 30% higher when a narrow scrape-off layer (SOL) grid width is chosen in SOLPS-ITER compared to the case for which the SOL grid width is maximised. The density increase is caused by kinetic neutrals being not confined inside the divertor region because of the reduced extent of the plasma grid. In this case, an enhanced level of reflections of energetic neutrals at the low-field side (LFS) metal divertor wall is observed. This leads to a shift of the ionisation source further upstream which must be accounted for as a numerical artefact. An overestimate in the cooling at the divertor entrance is observed in this case, identified by a reduced heat flux decay parameters lambda(div)(q). Otherwise and further upstream the mid-plane heat decay length lambda(q) parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that lambda(q) upstream is essentially set by the assumptions for the ratio radial to parallel heat conductivity. It is also shown that even for attached conditions the decay length relations lambda(ne)>lambda(Te)>lambda(q) hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of lambda(ne) into account, rather than lambda(q), as the former actually defines the required minimum upstream SOL grid extent.

  • 817. Wiesen, S.
    et al.
    Brezinsek, S.
    Wischmeier, M.
    De La Luna, E.
    Groth, M.
    Jaervinen, A. E.
    De La Cal, E.
    Losada, U.
    De Aguilera, A. M.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. Association VR.
    Gao, Y.
    Guillemaut, C.
    Harting, D.
    Meigs, A.
    Schmid, K.
    Sergienko, G.
    Impact of the JET ITER-like wall on H-mode plasma fueling2017In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 6, article id 066024Article in journal (Refereed)
    Abstract [en]

    JET ITER-like wall (ILW) experiments show that the edge density evolution is strongly linked with the poloidal distribution of the ionization source. The fueling profile in the JET-ILW is more delocalized as compared to JET-C (JET with carbon-based plasma-facing components PFCs). Compared to JET-C the H-mode pedestal fueling cycle is dynamically influenced by a combination of plasma-wall interaction features, in particular: (1) edge-localized modes (ELMs) induced energetic particles are kinetically reflected on W divertor PFCs leading to distributed refueling away from the divertor depending on the divertor plasma configuration, (2) delayed molecular re-emission and outgassing of particles being trapped in W PFCs (bulk-W at the high field side and W-coated CFCs at the low field side) with different fuel content and (3) outgassing from Be co-deposits located on top of the high-field side baffle region shortly after the ELM. In view of the results of a set of well diagnosed series of JET-ILW type-I ELMy H-mode discharges with good statistics, the aforementioned effects are discussed in view of H-mode pedestal fueling capacity. The ongoing modelling activities with the focus on coupled core-edge plasma simulations and plasma-wall interaction are described and discussed also in view of possible code improvements required.

  • 818. Wilson, H. R.
    et al.
    Bowman, C.
    Cowley, S. C.
    Cziegler, I.
    Dickinson, D.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Gibson, K.
    Ham, C.
    Horvath, L.
    Kirk, A.
    Lipschultz, B.
    Lunniss, A. E. L.
    Maggi, C. F.
    Roach, C. M.
    Saarelma, S.
    Snyder, P. B.
    Thornton, A.
    Wynn, A.
    Inter-ELM pedestal evolution in low triangularity JET-ILW discharges2017In: 44th EPS Conference on Plasma Physics, EPS 2017, European Physical Society (EPS) , 2017Conference paper (Refereed)
    Abstract [en]

    Studies of the pedestal stability in low triangularity, d=0.2, JET ITER-Like Wall discharges are presented, following the evolution between ELMs. The pressure gradient tracks the ideal MHD ballooning threshold, only lagging behind it when the threshold rises rapidly as the plasma penetrates the second stability region. This is consistent with a role for the kinetic ballooning mode in the pedestal structure. When the plasma has second stability access, e.g. for low gas puff discharges, the peeling-ballooning mode is marginally stable at ELM onset. In cases where there is no second stability access the discharges are some way short of the peeling-ballooning threshold, so this alone cannot be the trigger for the ELM. A low amplitude sinusoidal oscillation in the Be-II emission is observed that correlates well with the ELMs, and has an associated high frequency magnetic field fluctuation, ~100-250kHz, with modulated amplitude. This might be associated with a new filamentary equilibrium state. 

  • 819. Wisse, M.
    et al.
    Marot, L.
    Widdowson, A.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Ivanova, Darya
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Doerner, R. P.
    Baldwin, M. J.
    Likonen, J.
    Alves, E.
    Hakola, A.
    Koivuranta, S.
    Steiner, R.
    Meyer, E.
    Laser-assisted cleaning of beryllium-containing mirror samples from JET and PISCES-B2014In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 89, no 2, p. 122-130Article in journal (Refereed)
    Abstract [en]

    A set of seven polycrystalline mirror samples retrieved from the JET tokamak has been cleaned in vacuum using a pulsed laser system. The surfaces of samples exposed to plasma during 2008-2009 campaigns as part of the second phase of a comprehensive first mirror test contained a mixture of carbon, beryllium and tritium. For this reason, the samples were treated in a vacuum chamber constructed specially for this purpose. In some cases mirrors show an increase of the specular reflectivity after cleaning, though beryllium and carbon deposits were not fully removed. Additionally, three samples coated in PISCES-B with a 110-120 nm beryllium layer were subjected to laser cleaning tests as well.

  • 820.
    Yadikin, Dimitry
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Brunsell, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Paccagnella, R.
    Advanced feedback control methods in EXTRAP T2R2006In: Physics of Plasmas, ISSN 1070-664X, E-ISSN 1089-7674, Vol. 13, no 7, p. 072109-Article in journal (Refereed)
    Abstract [en]

    Previous experiments in the EXTRAP T2R reversed field pinch device have shown the possibility of suppression of multiple resistive wall modes (RWM). A feedback system has been installed in EXTRAP T2R having 100% coverage of the toroidal surface by the active coil array. Predictions based on theory and the previous experimental results show that the number of active coils should be sufficient for independent stabilization of all unstable RWMs in the EXTRAP T2R. Experiments using different feedback schemes are performed, comparing the intelligent shell, the fake rotating shell, and the mode control with complex feedback gains. Stabilization of all unstable RWMs throughout the discharge duration of t(d)approximate to 10 tau(w) is seen using the intelligent shell feedback scheme. Mode rotation and the control of selected Fourier harmonics is obtained simultaneously using the mode control scheme with complex gains. Different sensor signals are studied. A feedback system with toroidal magnetic field sensors could have an advantage of lower feedback gain needed for the RWM suppression compared to the system with radial magnetic field sensors. In this study, RWM suppression is demonstrated, using also the toroidal field component as a sensor signal in the feedback system.

  • 821.
    Yadikin, Dmitriy
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Brunsell, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Drake, James Robert
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Intelligent shell feedback control of resistive wall modes in EXTRAP T2R2005In: 32nd EPS Conference on Plasma Physics 2005, EPS 2005, Held with the 8th International Workshop on Fast Ignition of Fusion Targets: Europhysics Conference Abstracts, 2005, p. 1602-1605Conference paper (Refereed)
  • 822. Yadykin, D.
    et al.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Delabie, E.
    Chapman, I. T.
    Gerasimov, S.
    Kempenaars, M.
    Rimini, F. G.
    Studies of the non-axisymmetric plasma boundary displacement in JET in presence of externally applied magnetic field2015In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 57, no 10, article id 104003Article in journal (Refereed)
    Abstract [en]

    Non-axisymmetric plasma boundary displacement is caused by the application of the external magnetic field with low toroidal mode number. Such displacement affects edge stability, power load on the first wall and could affect efficiency of the ICRH coupling in ITER. Studies of the displacement are presented for JET tokamak focusing on the interaction between error field correction coils (EFCCs) and shape control system. First results are shown on the direct measurement of the plasma boundary displacement at different toroidal locations. Both qualitative and quantitative studies of the plasma boundary displacement caused by interaction between EFCCs and shape control system are performed for different toroidal phases of the external field. Axisymmetric plasma boundary displacement caused by the EFCC/shape control system interaction is seen for certain phase values of the external field. The value of axisymmetric plasma boundary displacement caused by interaction can be comparable to the non-axisymmetric plasma boundary displacement value produced by EFCCs.

  • 823. Yadykin, D.
    et al.
    Gryaznevich, M.
    Frasinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Gerasimov, S.
    Effect of the external helical fields on the plasma boundary shape in JET2014In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 54, no 1, p. 013016-Article in journal (Refereed)
    Abstract [en]

    Externally applied helical magnetic fields are now often used on tokamaks for various purposes. This paper presents results of studies of the effect of the external fields, produced by the error field correction coils (EFCCs) on JET, on the plasma boundary shape. Significant 3D distortions, predicted in the previous studies, have been confirmed using upgraded magnetic diagnostics and high-resolution Thomson scattering diagnostics. A simple method of estimating the edge distortion using magnetic diagnostics calibrated on the kinetic measurements is proposed and demonstrated.

  • 824. Yambe, Kiyoyuki
    et al.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Hirano, Yoichi
    Yagi, Yasuyuki
    Koguchi, Haruisa
    Sakakita, Hajime
    Frequency Dependence of Fast Magnetic Fluctuations in TPE-RX plasma2008In: Journal of Plasma and Fusion Research, ISSN 0918-7928, Vol. 3, p. 060-Article in journal (Refereed)
  • 825. Yambe, Myoyuki
    et al.
    Kiyama, Satoru
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Hirano, Yoichi
    Yagi, Yasuyuki
    Koguchi, Haruhisa
    Sakakita, Hajime
    Measurement of fast magnetic fluctuations in edge region of TPE-RX reversed-field pinch plasma2007In: Japanese Journal of Applied Physics, ISSN 0021-4922, E-ISSN 1347-4065, Vol. 46, no 10A, p. 6831-6833Article in journal (Refereed)
    Abstract [en]

    Fast magnetic fluctuation (0) levels are measured using a newly developed complex edge probe (CEP) in the edge region of TPE-RX reversed-field pinch plasma. The CEP is installed inside a vacuum vessel and is sensitive to fast delta B. The delta B levels measured using the CEP are compared with those measured using an extensive magnetic measurement system (MMS), which is located outside the vacuum vessel and has little sensitivity to fast delta B. The delta B levels before and after the appearance of the locked mode (LM) are compared in both the CEP and MMS signals. It was found that the rapid increase in the delta B signal level obtained using the MMS just after the appearance of LM is mainly caused by the slowing of plasma rotation.

  • 826. Yavorskij, V.
    et al.
    Cecconello, Marco
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics. Culham Science Centre, United Kingdom .
    Goloborod'ko, V.
    Brix, M.
    Challis, C.
    Gerasimov, S.
    Kiptily, V.
    Korotkov, A.
    Parail, V.
    Reznik, S.
    Santala, M.
    Schoepf, K.
    Sharapov, S. E.
    Surrey, E.
    De Vries, P.
    TF ripple effects on the NBI deuteron confinement in JET2007In: 34th EPS Conference on Plasma Physics 2007, EPS 2007 - Europhysics Conference Abstracts, 2007, no 2, p. 876-879Conference paper (Refereed)
    Abstract [en]

    Ripple induced reduction of the fluxes of deuterium neutrals in the 5-40 keV energy range from the plasma mid-plane was observed in recent JET experiments. The maximum observed reduction of D0 fluxes due to ripple is approximately 50 % and occurs at energies above 30 keV. In positive shear plasmas without ICRH ripple reduction of D0 fluxes vanishes at energies below 10 keV. However, in the case of plasmas with low or reversed shear core, increased D0 fluxes were observed for energies below 10-20 keV in the presence of additional ripple and ICRH heating. Interpretive modeling of the deuterium neutral emission that accounts for the superbanana ripple diffusion of NBI ions is in reasonable agreement with measurements at least for the scenarios without ICRH. Note that ripples may essentially effect the fast ion confinement in ITER where TF ripple magnitude at the outer separatrix is expected to be δ ∼ 0.5%.

  • 827. Zaitsev, F. S.
    et al.
    Gondhalekar, A.
    Johnson, Thomas J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Sharapov, S. E.
    Testa, D. S.
    Kurbet, I. I.
    Suprathermal deuterium ions produced by nuclear elastic scattering of ICRH driven He-3 ions in JET plasmas2007In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 49, no 11, p. 1747-1766Article in journal (Refereed)
    Abstract [en]

    Measurements of the suprathermal tail of the energy distribution function of deuterium ions, in plasmas containing MeV energy ICRH driven minority He-3 ions and majority deuterium ions, revealed that the suprathermal tail ion density exceeded by nearly an order of magnitude that expected due to nuclear elastic scattering (NES) of He-3 projectile ions on deuterium target ions. The experiments were performed on the Joint European Torus (JET), measurements of the line-of-sight integrated energy distribution functions of He-3 and suprathermal deuterium ions were made using a high energy neutral particle analyzer. The NES or 'knock-on' deuterium ion energy distribution function was simulated using the FPP-3D Fokker-Plank code (Zaitsev et al 2002 Nucl. Fusion 42 1340) which solves the 3D trajectory averaged kinetic equations in JET tokamak geometry while taking into account NES of He-3 ions on the deuterium ions. The required input energy distribution function of ICRH driven He-3 ions was simulated using the SELFO code (Hedin et al 2002 Nucl. Fusion 42 527). The comparison between measurement and simulation in the He-3 ICRH experiments is contrasted with an analogous previous comparison between measurements and simulation of JET plasmas in which 3.5MeV DT fusion alpha-particles were the projectile ions, where measurement and simulation roughly agreed. Possible explanations for the observed excess knock-on deuterium tail in the experiments with He-3 minority

  • 828. Zaitsev, F. S.
    et al.
    Gondhalekar, A.
    Johnson, Thomas Joe
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Kiptily, V. G.
    Sharapov, S. E.
    Simulations to elucidate suprathermal deuterium ion tail observed in He3 minority ICRF heated JET plasmas2008In: EPS Conf. Plasma Phys., EPS - Europhys. Conf. Abstr., 2008, no 1, p. 501-504Conference paper (Refereed)
  • 829. Zaitsev, F. S.
    et al.
    Gondhalekar, A.
    Johnson, Thomas Joe
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Sharapov, S. E.
    Testa, D. S.
    Kurbet, I. I.
    Simulation of deuteron tails produced by close collisions with ICRH Accelerated He3 ions in JET2006In: EPS Conf. Plasma Phys., EPS, 2006, p. 412-415Conference paper (Refereed)
  • 830. Zarzoso, D.
    et al.
    Beurskens, M. N. A.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Eich, T.
    Joffrin, E.
    Loarte, A.
    Maddison, G.
    Rimini, F. G.
    Saibene, G.
    Solano, E. R.
    Thomsen, H.
    ELM size analysis in JET advanced tokamak and hybrid scenarios2010In: 37th EPS Conference on Plasma Physics 2010, EPS 2010: Volume 2, 2010, p. 922-925Conference paper (Refereed)
  • 831. Zarzoso, D.
    et al.
    Beurskens, M. N. A.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Joffrin, E.
    Rimini, F. G.
    Solano, E. R.
    ELM size analysis in JET hybrid plasmas2011In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 51, no 11, p. 112001-Article in journal (Refereed)
    Abstract [en]

    Experimental results are presented in this paper, characterizing the behaviour of type I ELMs for a JET database of standard ELMy H-mode and hybrid plasmas. Whereas the collisionality scaling published in [ 1] has been reproduced for the new baseline discharges, no clear correlation can be established from the analysis of hybrid scenarios. The ELM losses normalized to the pedestal stored energy for high triangularity hybrid plasmas seem to be significantly larger than the energies for baseline plasmas at similar values of collisionality. For low triangularity hybrid plasmas, the ELM losses are of the same order as those obtained in baseline scenarios. The important scatter of the results seems to be due to the sensitivity of hybrid plasmas to gas fuelling. Analysis of the ITER-like wall compatibility of hybrid discharges is also reported.

  • 832. Zarzoso, D.
    et al.
    Beurskens, M.N.A.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Joffrin, E.
    ELM size scaling in JET advanced tokamak scenarios2010Conference paper (Refereed)
  • 833. Zhao, M. L.
    et al.
    Chen, Y. P.
    Guo, H. Y.
    Ye, M. Y.
    Tendler, Michael
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Li, G. Q.
    Luo, Z. P.
    Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene2014In: Physics of Plasmas, ISSN 1070-664X, E-ISSN 1089-7674, Vol. 21, no 5, p. 052503-Article in journal (Refereed)
    Abstract [en]

    The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10MW/m(2), which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10MW/m(2) even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure.

  • 834.
    Zhou, Yushan
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Impact of Surface Structures onDeposition and Erosion in a Tokamak2019Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    Fusion is a potentially unlimited and environmentally friendly energy source for human society in the future. However, along the way towards the application of fusion energy there are still unresolved complications. Among them, deposition and erosion are two critical issues. Deposition of fuel and impurities brings potential long-term fuel retention which may generate safety issues and limit the economic efficiency of fusion devices. Moreover, the erosion of the vacuum vessel wall in a fusion device generates impurities which contaminate core plasma and can restrict the life time of plasma facing component. The work in this thesis focuses on deposition and erosion on tiles in the JET-ILW project, which consist of tungsten (or tungsten coating carbon fibre composited) in the divertor and beryllium in limiters.

    For the deposition issue, micro ion beam analysis (µ-IBA) was used for observing deuterium and beryllium distributions over tile surfaces. The surface topography was obtained from SEM, optical microscope and confocal laser scan microscope. Distribution maps from IBA were compared with surface topography. To explain experimental results, modelling of ion trajectories was applied on real and artificial surfaces. Micro IBA results show that deuterium and beryllium accumulated in depressed areas, e.g. pits, cracks or craters. Modelling implies that ion gyration, surface roughness and inclination of the magnetic field could to some extent explain this non-uniform distribution of deuterium and beryllium. The same kind of issue, although on different scale length, occurs also for penetration of impurities into artificial castellation grooves, also studied experimentally in the thesis.

    For the erosion issue, the thesis includes analysis of a limiter marker tile which is designed for observing material erosion in JET. A new method to acquire erosion data from such marker tiles is proposed, by combining micro IBA and SEM image.  This method could separate the influence on IBA from roughness, a problem in applying IBA on rough surface. Similar Technique is applied to improve the interpretation of IBA measurements of deep penetration of deuterium into layered surface structures.

  • 835.
    Zhou, Yushan
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Bykov, Igor
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Paneta, C.
    Possnert, G.
    Micro ion beam analysis for the erosion of beryllium marker tiles in a tokamak limiter2019In: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 450, p. 200-204Article in journal (Refereed)
    Abstract [en]

    Beryllium limiter marker tiles were exposed to plasma in the Joint European Torus to diagnose the erosion of main chamber wall materials. A limiter marker tile consists of a beryllium coating layer (7–9 μm) on the top of bulk beryllium, with a nickel interlayer (2–3 μm) between them. The thickness variation of the beryllium coating layer, after exposure to plasma, could indicate the erosion measured by ion beam analysis with backscattering spectrometry. However, interpretations from broad beam backscattering spectra were limited by the non-uniform surface structures. Therefore, micro-ion beam analysis (μ-IBA) with 3 MeV proton beam for Elastic backscattering spectrometry (EBS) and PIXE was used to scan samples. The spot size was in the range of 3–10 μm. Scanned areas were analysed with scanning electron microscopy (SEM) as well. Combining results from μ-IBA and SEM, we obtained local spectra from carefully chosen areas on which the surface structures were relatively uniform. Local spectra suggested that the scanned area (≈600 μm × 1200 μm) contained regions with serious erosion with only 2–3 μm coating beryllium left, regions with intact marker tile, and droplets with 90% beryllium. The nonuniform erosion, droplets mainly formed by beryllium, and the possible mixture of beryllium and nickel were the major reasons that confused interpretation from broad beam EBS.

  • 836.
    Zhou, Yushan
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Bykov, Igor
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Possnert, G.
    Likonen, J.
    Pettersson, J.
    Koivuranta, S.
    Widdowson, A. M.
    Microanalysis of deposited layers in the inner divertor of JET with ITER-like wall2017In: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, p. 412-417Article in journal (Refereed)
    Abstract [en]

    In JET with ITER-like wall, beryllium eroded in the main chamber is transported to the divertor and deposited mainly at the horizontal surfaces of tiles 1 and 0 (high field gap closure, HFGC). These surfaces are tungsten coated carbon fibre composite (CFC). Surface sampleswere collected following the plasma operations in 2011-2012 and 2013-2014 respectively. The surfaces, as well as polished cross sections of the deposited layers at the surfaces have been studied with micro ion beam analysis methods (mu-IBA). Deposition of Beand other impurities, and retention of D is microscopically inhomogeneous. Impurities and trapped deuterium accumulate preferentially in cracks, pits and depressed regions, and at the sides of large pits in the substrate (e.g. arc tracks where the W coating has been removed). With careful overlaying of mu-NRA elemental maps with optical microscopy images, it is possible to separate surface roughness effects from depth profiles at microscopically flat surface regions.

  • 837.
    Zhou, Yushan
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Modelling of effect from rough surface on deuterium and beryllium deposition on divertor targetManuscript (preprint) (Other academic)
  • 838.
    Zhou, Yushan
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Possnert, G
    Likonen, J
    contributors, JET
    The effect of gyration on the deposition of beryllium and deuterium at rough surface on thedivertor tiles with ITER-like-wall in JET2019In: Nuclear Materials and Energy, E-ISSN 2352-1791Article in journal (Refereed)
  • 839. Zuin, M.
    et al.
    Dal Bello, S.
    Marrelli, L.
    Puiatti, M. E.
    Agostinetti, P.
    Agostini, M.
    Antoni, V.
    Auriemma, F.
    Barbisan, M.
    Barbui, T.
    Baruzzo, M.
    Belli, F.
    Bettini, P.
    Bigi, M.
    Bilel, R.
    Boldrin, M.
    Bolzonella, T.
    Bonfiglio, D.
    Brombin, M.
    Buffa, A.
    Bustreo, C.
    Canton, A.
    Cappello, S.
    Carraro, L.
    Cavazzana, R.
    Cester, D.
    Chacon, L.
    Chitarin, G.
    Cooper, W. A.
    Cordaro, L.
    Palma, M. Dalla
    Deambrosis, S.
    Delogu, R.
    De Lorenzi, A.
    De Masi, G.
    Dong, J. Q.
    Escande, D. F.
    Fassina, A.
    Felici, F.
    Ferro, A.
    Finotti, C.
    Franz, P.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Gaio, E.
    Ghezzi, F.
    Giudicotti, L.
    Gnesotto, F.
    Gobbin, M.
    Gonzalez, W. A.
    Grando, L.
    Guo, S. C.
    Hanson, J. D.
    Hirshman, S. P.
    Innocente, P.
    Jackson, J. L.
    Kiyama, S.
    Komm, M.
    Kudlacek, O.
    Laguardia, L.
    Li, C.
    Liu, B.
    Liu, S. F.
    Liu, Y. Q.
    Lopez-Bruna, D.
    Lorenzini, R.
    Luce, T. C.
    Luchetta, A.
    Maistrello, A.
    Manduchi, G.
    Mansfield, D. K.
    Marchiori, G.
    Marconato, N.
    Marcuzzi, D.
    Martin, P.
    Martines, E.
    Martini, S.
    Mazzitelli, G.
    McCormack, O.
    Miorin, E.
    Momo, B.
    Moresco, M.
    Narushima, Y.
    Okabayashi, M.
    Paccagnella, R.
    Patel, N.
    Pavei, M.
    Peruzzo, S.
    Pilan, N.
    Pigatto, L.
    Piovan, R.
    Piovesan, P.
    Piron, C.
    Piron, L.
    Predebon, I.
    Pucella, G.
    Rea, C.
    Recchia, M.
    Rizzolo, A.
    Rostagni, G.
    Ruset, C.
    Sajo-Bohus, L.
    Sakakita, H.
    Sanchez, R.
    Sarff, J. S.
    Sattin, F.
    Scarin, P.
    Schmitz, O.
    Schneider, W.
    Siragusa, M.
    Sonato, P.
    Spada, E.
    Spagnolo, S.
    Spolaore, M.
    Spong, D. A.
    Spizzo, G.
    Stevanato, L.
    Suzuki, Y.
    Taliercio, C.
    Terranova, D.
    Tudisco, O.
    Urso, G.
    Valente, M.
    Valisa, M.
    Vallar, M.
    Veranda, M.
    Vianello, N.
    Villone, F.
    Vincenzi, P.
    Visona, N.
    White, R. B.
    Xanthopoulos, P.
    Xu, X. Y.
    Yanovskiy, V.
    Zamengo, A.
    Zanca, P.
    Zaniol, B.
    Zanotto, L.
    Zhang, Y.
    Zilli, E.
    Overview of the RFX-mod fusion science activity2017In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 10, article id 102012Article in journal (Refereed)
    Abstract [en]

    This paper reports the main recent results of the RFX-mod fusion science activity. The RFX-mod device is characterized by a unique flexibility in terms of accessible magnetic configurations. Axisymmetric and helically shaped reversed-field pinch equilibria have been studied, along with tokamak plasmas in a wide range of q(a) regimes (spanning from 4 down to 1.2 values). The full range of magnetic configurations in between the two, the so-called ultra-low q ones, has been explored, with the aim of studying specific physical issues common to all equilibria, such as, for example, the density limit phenomenon. The powerful RFX-mod feedback control system has been exploited for MHD control, which allowed us to extend the range of experimental parameters, as well as to induce specific magnetic perturbations for the study of 3D effects. In particular, transport, edge and isotope effects in 3D equilibria have been investigated, along with runaway mitigations through induced magnetic perturbations. The first transitions to an improved confinement scenario in circular and D-shaped tokamak plasmas have been obtained thanks to an active modification of the edge electric field through a polarized electrode. The experiments are supported by intense modeling with 3D MHD, gyrokinetic, guiding center and transport codes. Proposed modifications to the RFX-mod device, which will enable further contributions to the solution of key issues in the roadmap to ITER and DEMO, are also briefly presented.

  • 840.
    Åkermark, Torbjörn
    et al.
    KTH, School of Information and Communication Technology (ICT), Material Physics.
    Emmoth, Birger
    KTH, School of Information and Communication Technology (ICT), Microelectronics and Applied Physics, MAP.
    Bergsåker, Henric
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Laser annealing in combination with mass spectroscopy, a technique to study deuterium on tokamak carbon samples, a tool for detritiation2006In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 359, no 3, p. 220-226Article in journal (Refereed)
    Abstract [en]

    In this study, a method is presented based on mass spectroscopy to measure the a real density of deuterium on a graphite surface exposed to tokamak discharges. The studied sample was cut from a bumper limiter exposed in the TEXTOR tokamak and annealed by a 1 J Excimer laser (KrF). The energy used was 400 mJ cm(-2), which is below the threshold for ablation. 1 J cm(-2). The release of HD and D, was measured by a mass spectroscopy set-up and no other species released from the sample were detected in this experiment. The amount of D released from the sample after 20 laser pulses was measured to 7 x 10(16) D atoms per cm(-2) (for this particular sample) and most of the hydrogen at the surface was released in the first pulse, as checked by nuclear reaction analysis (NRA) techniques, which gave changes of the amount of deuterium before and after laser annealing. The sensitivity in this experiment was 5 x 10(14) atoms per cm(-2) for HD and 5 x 10(13) atoms per cm(-2) for D-2.

  • 841.
    Štefániková, E.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Lomas, P.
    Nunes, I.
    Baruzzo, M.
    Rimini, F.
    Saarelma, S.
    Wiesen, S.
    Peterka, M.
    Confinement and pedestal structure in high performance scenarios in JET-ILW and comparison with JET-C2015In: 42nd European Physical Society Conference on Plasma Physics, EPS 2015, European Physical Society (EPS) , 2015Conference paper (Refereed)
14151617 801 - 841 of 841
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