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  • 1. Alegre, Daniel
    et al.
    Bergsåker, Henric
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bykov, Igor
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Gasior, Pawel
    Kubkowska, Monika
    Kowalska-Strzeciwilk, Ewa
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Tabares, Francisco L.
    Study of correlation of deuterium content in a-C:D dust induced by laser irradiation from the co-deposited surface with the grain size and velocity2014Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T161, s. 014010-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    In the study described here, the laser ablation method was applied to clean thick (40-60 m) a-C: D co-deposits on the ALT-II limiter blade from the TEXTOR tokamak, and at the same time to characterize the ejected particles formed during ablation and measure the amount of fuel carried by them. Ablation was accomplished by similar to 3.5 ns, 0.5 J Nd: YAG laser pulses in either vacuum or an O-2 atmosphere at different pressures. Fast camera tracking of the process provided an estimate of the population and velocity of up to 100ms(-1) for larger dust particles. In the same experiment, the dust particles were caught using ultra-light Si aerogel collectors placed in front of the ablation target. SEM analysis of aerogel surfaces verified the speed estimate, providing the trapped particles' size distribution and particle yield during ablation. The D/C atomic concentration ratio was measured with the 3HE ion beam nuclear reaction analysis method in deposited layers before ablation and with a micro-ion beam in individual particles on aerogel collectors. This indicated that most of the D was thermally released during ablation, leaving no more than 5% of its original amount in the particles. The effect of ablation conditions on the acceleration of ejected particles, their population, composition and D content is the main subject of this paper.

  • 2. Baron-Wiechec, A.
    et al.
    Widdowson, A.
    Alves, E.
    Ayres, C. F.
    Barradas, N. P.
    Brezinsek, S.
    Coad, J. P.
    Catarino, N.
    Heinola, K.
    Likonen, J.
    Matthews, G. F.
    Mayer, M.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    van Renterghem, W.
    Uytdenhouwen, I.
    Global erosion and deposition patterns in JET with the ITER-like wall2015Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, s. 157-161Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    A set of Be and W tiles removed after the first ITER-like wall campaigns (JET-ILW) from 2011 to 2012 has been analysed. The results indicate that the primary erosion site is in the main chamber (Be) as in previous carbon campaigns (JET-C). In particular the limiters tiles near the mid-plane are eroded probably during the limiter phases of discharges. W is found at low concentrations on all plasma-facing surfaces of the vessel indicating deposition via plasma transport initially from the W divertor and from main chamber W-coated tiles; there are also traces of Mo (used as an interlayer for these coatings). Deposited films in the inner divertor have a layered structure, and every layer is dominated by Be with some W and O content.

  • 3.
    Bergsåker, B. Henric M.
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. EUROfusion Consortium, Culham Science Centre, JET, Abingdon, United Kingdom.
    Bykov, Igor
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. EUROfusion Consortium, Culham Science Centre, JET, Abingdon, United Kingdom.
    Zhou, Yushan
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. EUROfusion Consortium, Culham Science Centre, JET, Abingdon, United Kingdom.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. EUROfusion Consortium, Culham Science Centre, JET, Abingdon, United Kingdom.
    Possnert, G.
    Likonen, J.
    Pettersson, J.
    Koivuranta, S.
    Widdowson, A. M.
    Deep deuterium retention and Be/W mixing at tungsten coated surfaces in the JET divertor2016Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T167, artikel-id 014061Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Surface samples from a full poloidal set of divertor tiles exposed in JET through operations 2010-2012 with ITER-like wall have been investigated using SEM, SIMS, ICP-AES analysis and micro beam nuclear reaction analysis (μ-NRA). Deposition of Be and retention of D is microscopically inhomogeneous. With careful overlaying of μ-NRA elemental maps with SEM images, it is possible to separate surface roughness effects from depth profiles at microscopically flat surface regions, without pits. With (3He, p) μ-NRA at 3-5 MeV beam energy the accessible depth for D analysis in W is about 9 μm, sufficient to access the W/Mo and Mo/W interfaces in the coatings and beyond, while for Be in W it is about 6 μm. In these conditions, at all plasma wetted surfaces, D was found throughout the whole accessible depth at concentrations in the range 0.2-0.7 at% in W. Deuterium was found to be preferentially trapped at the W/Mo and Mo/W interfaces. Comparison is made with SIMS profiling, which also shows significant D trapping at the W/Mo interface. Mixing of Be and W occurs mainly in deposited layers.

  • 4.
    Bergsåker, Henric
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bykov, Igor
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Possnert, G.
    Likonen, J.
    Koivuranta, S.
    Coad, J. P.
    Widdowson, A. M.
    Microstructure and inhomogeneous fuel trapping at divertor surfaces in the JET tokamak2014Ingår i: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 332, s. 266-270Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The plasma deposited layers at divertor surfaces in the JET tokamak with carbon wall have been studied post mortem, using micro ion beam analysis (mu-IBA) methods, optical microscopy and scanning electron microscopy (SEM). The layers were formed during plasma operations over different periods through 1998-2009. They frequently have a columnar structure. For mu-IBA a 3 MeV He-3 beam was used, focused to about 5-15 mu m size. Nuclear reaction analysis was used to measure D, Be and C. Elemental mapping was carried out both at the original surface and on polished layer cross sections. Trapped deuterium is predominantly found in remote areas on the horizontal bottom divertor tiles and in regions with locally enhanced deuterium concentration on the vertical tiles. Pockets with enhanced deuterium concentration are found in the carbon fibre composite (CFC) substrate. Areas with dimensions of about 100 mu m with enhanced deuterium concentration are also found inside the deposited layers. The inhomogeneous fuel trapping is tentatively explained with co-deposition in partly protected pits in the substrate and by incorporation of dust particles in the growing layers.

  • 5.
    Bergsåker, Henric
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bykov, Igor
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Possnert, G.
    Likonen, J.
    Koivuranta, S.
    Coad, J. P.
    Widdowson, A. M.
    Microanalysis of deposited layers in the divertor of JET following operations with carbon wall2013Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 438, nr Suppl., s. S668-S672Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Elemental mapping of cross sections of deposited layers on inboard tiles in the JET divertor after exposure to plasma operations with carbon wall are presented. The study was made using microbeam ion beam analysis methods in combination with optical microscopy and SEM. The surfaces had been exposed to plasma through different periods of operation (1998-2007, 2007-2009 and 1998-2009). The texture and composition of the layers are non-uniform. The physical structures include columnar, lamellar and disordered globular appearances. The distribution of trapped deuterium was frequently found to be lamellar, with well-defined sub layers with higher deuterium concentration. However, 3D regions with dimensions of about 100 μm with enhanced deuterium content were also found, both at the layer surfaces and in the layer cross sections. The distributions of beryllium and Inconel components were lamellar but did not otherwise show large non-uniformity on the same scale length as the deuterium.

  • 6.
    Bergsåker, Henric
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Possnert, G.
    Bykov, Igor
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Heinola, K.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Miettunen, J.
    Widdowson, A.
    Riccardo, V.
    Nunes, I.
    Stamp, M.
    Brezinsek, S.
    Groth, M.
    Kurki-Suonio, T.
    Likonen, J.
    Coad, J. P.
    Borodin, D.
    Kirschner, A.
    Schmid, K.
    Krieger, K.
    First results from the Be-10 marker experiment in JET with ITER-like wall2014Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 54, nr 8, s. 082004-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    When the ITER-like wall was installed in JET, one of the 218 Be inner wall guard limiter tiles had been enriched with Be-10 as a bulk isotopic marker. During the shutdown in 2012-2013, a set of tiles were sampled nondestructively to collect material for accelerator mass spectroscopy measurements of Be-10 concentration. The letter shows how the marker experiment was set up, presents first results and compares them to preliminary predictions of marker redistribution, made with the ASCOT numerical code. Finally an outline is shown of what experimental data are likely to become available later and the possibilities for comparison with modelling using the WallDYN, ERO and ASCOT codes are discussed.

  • 7.
    Bergsåker, Henric
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Possnert, G.
    Bykov, Igor
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Heinola, K.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Miettunen, J.
    Widdowson, C.
    Riccardo, V.
    Nunes, I.
    Stamp, M.
    et al.,
    First results from10Be marker experiment in JET with ITER-like wallManuskript (preprint) (Övrigt vetenskapligt)
  • 8.
    Bergsåker, Henrik
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bykov, Igor
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Possnert, G.
    Likonen, J.
    Koivuranta, S.
    Coad, J. P.
    Van Renterghem, W.
    Uytdenhouwen, I.
    Widdowson, A. M.
    Microscopically nonuniform deposition and deuterium retention in the divertor in JET with ITER-like wall2015Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, s. 956-960Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The divertor surfaces in JET with ITER-like wall (ILW) have been studied using micro ion beam analysis (mu-IBA) methods and scanning electron microscopy (SEM). Deposited layers with beryllium as main constituent had been formed during plasma operations through 2011-2012. The deuterium trapping and impurity deposition were non-uniform, frequently enhanced within pits, cracks and valleys, regions reaching in size from 10 mu m to 200 mu m. The impurity deposition and fuel retention were correlated with the surface slope with respect to the direction of ion incidence. Typically more than 70% of the total measured areal density of trapped D was found in less than 30% of the surface area. This is of consequence for the interpretation of other surface analyses and in extrapolation from fuel retention in JET with ITER-like wall and rough divertor surfaces to ITER with smoother surfaces.

  • 9. Brezinsek, S.
    et al.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ratynskaia, Svetlana
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ström, Petter
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Tolias, Panagiotis
    KTH, Skolan för elektro- och systemteknik (EES), Rymd- och plasmafysik.
    Weckmann, Armin
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Zaplotnik, R.
    et al.,
    Plasma-wall interaction studies within the EUROfusion consortium: Progress on plasma-facing components development and qualification2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 11, artikel-id 116041Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma-material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.

  • 10. Brezinsek, S.
    et al.
    Widdowson, A.
    Mayer, M.
    Philipps, V.
    Baron-Wiechec, P.
    Coenen, J. W.
    Heinola, K.
    Huber, A.
    Likonen, J.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Stamp, M. F.
    Borodin, D.
    Coad, J. P.
    Carrasco, Alvaro Garcia
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Kirschner, A.
    Krat, S.
    Krieger, K.
    Lipschultz, B.
    Linsmeier, Ch.
    Matthews, G. F.
    Schmid, K.
    Beryllium migration in JET ITER-like wall plasmas2015Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, nr 6, artikel-id 063021Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (E-in = 35 eV) and more than 100%, caused by Be self-sputtering (E-in = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at E-in = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.

  • 11.
    Bykov, Igor
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bergsaker, Henric
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Possnert, G.
    Zhou, Y.
    Heinola, K.
    Pettersson, J.
    Conroy, S.
    Likonen, J.
    Petersson, Per
    Widdowson, A.
    Studies of Be migration in the JET tokamak using AMS with Be-10 marker2016Ingår i: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 371, s. 370-375Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The JET tokamak is operated with beryllium limiter tiles in the main chamber and tungsten coated carbon fiber composite tiles and solid W tiles in the divertor. One important issue is how wall materials are migrating during plasma operation. To study beryllium redistribution in the main chamber and in the divertor, a Be-10 enriched limiter tile was installed prior to plasma operations in 2011-2012. Methods to take surface samples have been developed, an abrasive method for bulk Be tiles in the main chamber, which permits reuse of the tiles, and leaching with hot HCl to remove all Be deposited at W coated surfaces in the divertor. Quantitative analysis of the total amount of Be in cm(2) sized samples was made with inductively coupled plasma atomic emission spectroscopy (ICP-AES). The Be-10/Be-9 ratio in the samples was measured with accelerator mass spectrometry (AMS). The experimental setup and methods are described in detail, including sample preparation, measures to eliminate contributions in AMS from the B-10 isobar, possible activation due to plasma generated neutrons and effects of diffusive isotope mixing. For the first time marker concentrations are measured in the divertor deposits. They are in the range 0.4-1.2% of the source concentration, with moderate poloidal variation. (C) 2015 Elsevier B.V. All rights reserved.

  • 12.
    Bykov, Igor
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bergsåker, Henric
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ogata, Douglas
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ratynskaia, Svetlana
    KTH, Skolan för elektro- och systemteknik (EES), Rymd- och plasmafysik.
    Collection of mobile dust in the T2R reversed field pinch2012Ingår i: Nukleonika, ISSN 0029-5922, E-ISSN 1508-5791, Vol. 57, nr 1, s. 55-60Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Intensive plasma-wall interactions in fusion devices result in the impurity production and the formation of films of redeposited material, debris and dust. In present day devices, with short pulses, the mobile dust does not pose any serious operational problems, but it is a matter of serious concern for ITER and for later power producing devices with a high duty cycle. We report results of a dust collection experiment carried out at the T2R reversed field pinch device and related heavy impurity flux measurements. Dust and impurities were collected on passive Si surface probes and on ultralow density silica aerogel collectors. The advantage of the latter method is the possibility of nondestructive capture of the micron- and submicron-sized dust particles. The toroidal and radial deposition fluxes of dust particles and impurities are estimated and discussed in the light of the dominant forces acting on the dust.

  • 13.
    Bykov, Igor
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bergsåker, Henric
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Likonen, Jari
    Possnert, G.
    Widdowson, C.
    Combined ion micro probe and SEM analysis of strongly non uniform deposits in fusion devices2015Ingår i: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 342, s. 19-28Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Conventional ion beam analysis (IBA) of deposited layers from fusion devices may have insufficient accuracy due to strongly uneven appearance of the layers. Surface roughness and spatial variation of the matrix composition make interpretation of broad beam spectra complex and non obvious. We discuss complications of applied IBA arising for fusion-relevant surfaces and demonstrate how quantification can be improved by employing micro IBA methods. The analysis is bound to pre-defined regions on the sample surface and can be extended by employing beams of several types, scanning electron microscopy (SEM) and stereo SEM techniques.

  • 14.
    Bykov, Igor
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bergsåker, Henric
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Likonen, Jari
    Possnert, Göran
    Quantitative plasma-fuel and impurity profiling in thick plasma-deposited layers by means of micro ion beam analysis and SIMS2014Ingår i: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 332, s. 280-285Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The operation of the Joint European Torus (JET) with full-carbon wall during the last decades has proven the importance of material re-deposition processes in remote areas of the tokamak. The thickness of the deposits in shadowed areas can reach 1 mm. The main constituent is carbon, with little inclusion of Inconel components. Atomic fractions Be/C and D/C can locally reach 1. Three methods were used to study thick deposits on JET divertor surfaces: (i) NRA analysis with a 15 mu m wide, 3 MeV He-3 ion microbeam on a polished cross section of the layer to determine the concentration distribution of D, Be and C and the distribution of Ni by particle induced X-ray emission; (ii) elastic proton scattering (EPS) from the top of the layers with a broad proton beam at 3.5 and 4.6 MeV. These methods were absolutely calibrated using thick elemental targets. (iii) Depth profiling of D, Be and Ni was done with secondary ion mass spectrometry (SIMS), sputtering the layers from the surface. The three methods are complementary. The thickest layers are accessible only by microbeam mapping of the cross sections, albeit with limited spatial resolution. The SIMS has the best depth resolution, but is difficult for absolute quantification and is limited in accessible depth. The probed depth with proton backscattering is limited to about 30 mu m. The combination of all three methods provided a coherent picture of the layer composition. It was possible to correlate the SIMS profiling results to quantitative data obtained by the microbeam method.

  • 15.
    Bykov, Igor
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bergsåker, Henric
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ratynskaia, Svetlana
    KTH, Skolan för elektro- och systemteknik (EES), Rymd- och plasmafysik.
    Litnovsky, A.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Possnert, G.
    Time resolved collection and characterization of dust particles moving in the TEXTOR scrape-off layer2013Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 438, nr Suppl., s. S681-S685Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Moving dust has been collected in the SOL of TEXTOR in a time-resolved way with silica aerogel collectors [1-3]. The collectors were exposed to the toroidal particle flux in NBI heated discharges during the startup and flat top phase. Intrinsic dust was collected in several discharges. Other discharges were accompanied with injection of known amounts of pre-characterized dust (W, C flakes and C microspheres) from a position toroidally 120° away from the collector. Particle flux, composition and dust size distribution have been determined with SEM and EDX. Calibration allowed particle velocity estimates to be made. Upper limits for the deuterium content of individual dust grains have been determined by NRA.

  • 16.
    Bykov, Igor
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bergsåker, Henrik
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Possnert, G.
    Heinola, K.
    Miettunen, J.
    Groth, M.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Widdowson, A.
    Likonen, J.
    Materials migration in JET with ITER-like wall traced with a Be-10 isotopic marker2015Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, s. 773-776Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The current configuration of JET with ITER-like Wall (ILW) is the best available proxy for the ITER first wall. Beryllium redistribution in JET-ILW can be used for estimates of its migration in ITER. To trace it, a localized isotopic Be marker has been implemented. A bulk Be-9 tile has been enriched with Be-10 up to atomic concentrations of 1.7 x 10(-9) and installed at the inner midplane of JET before the campaign. During the 2012 shutdown over 100 surface samples were taken non destructively from surfaces of two toroidally opposite limiter beams. The absolute areal densities of the marker were inferred from Be-15 atomic concentration in each sample, measured with Accelerator Mass Spectrometry with sensitivity <10(-14). The results of marker mapping are compared with predictions made with the ASCOT orbit following code.

  • 17.
    Bykov, Igor
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bergsåker, Henric
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Hallén, A.
    Possnert, G.
    Investigation of tritium analysis methods for ion microbeam application2012Ingår i: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 273, s. 250-253Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The trapping and retention of tritium in deposited layers on plasma-facing components is a critical issue for the international tokamak experimental reactor (ITER) and for future power producing tokamak fusion reactors. Cross sections of deposited layers at surfaces in the JET tokamak divertor are being investigated using ion microbeam analysis. To include tritium analysis with high spatial resolution, a number of plausible ion beam techniques have been investigated. Calibration samples with 150 nm tritiated titanium films were used. Absolute concentrations were determined with classical ERD using 2.5-3.5 MeV C-12(+). Cross sections for non-Rutherford ERD and for the T(C-12,p)C-14 and T(C-12,alpha)B-11 nuclear reactions were measured for different angles in the energy range 2.5-15 MeV. Background spectra were collected from pure carbon, beryllium and deuterium enriched samples and the sensitivity for microbeam NRA measurements of the tritium concentration in thick targets with predominantly Be-C-D matrix was estimated.

  • 18. Coad, J. P.
    et al.
    Alves, E.
    Barradas, N. P.
    Baron-Wiechec, A.
    Catarino, N.
    Heinola, K.
    Likonen, J.
    Mayer, M.
    Matthews, G. F.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Widdowson, A.
    Surface analysis of tiles and samples exposed to the first JET campaigns with the ITER-like wall2014Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T159, s. 014012-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    This paper reports on the first post-mortem analyses of tiles removed from JET after the first campaigns with the ITER-like wall (ILW) during 2011-12 [1]. Tiles from the divertor have been analysed by ion beam analysis techniques and by secondary ion mass spectrometry to determine the amount of beryllium deposition and deuterium retention in the tiles exposed to the scrape-off layer. Films 10-20 mu m thick were present at the top of tile 1, but only very thin films (<1 mu m) were found in the shadowed areas and on other divertor tiles. The total amount of Be found in the divertor following the ILW campaign was a factor of similar to 9 less than the material deposited in the 2007-09 carbon campaign, after allowing for the longer operations in 2007-09.

  • 19. Coenen, J. W.
    et al.
    Matthews, G. F.
    Krieger, K.
    Iglesias, D.
    Bunting, P.
    Corre, Y.
    Silburn, S.
    Balboa, I.
    Bazylevs, B.
    Conway, N.
    Coffey, I.
    Dejarnac, R.
    Gauthier, E.
    Gaspar, J.
    Jachmich, S.
    Jepu, I.
    Makepeace, C.
    Scannell, R.
    Stamp, M.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Pitts, R. A.
    Wiesen, S.
    Widdowson, A.
    Heinola, K.
    Baron-Wiechec, A.
    Transient induced tungsten melting at the Joint European Torus (JET)2017Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, artikel-id 014013Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Melting is one of the major risks associated with tungsten (W) plasma-facing components (PFCs) in tokamaks like JET or ITER. These components are designed such that leading edges and hence excessive plasma heat loads deposited at near normal incidence are avoided. Due to the high stored energies in ITER discharges, shallow surface melting can occur under insufficiently mitigated plasma disruption and so-called edge localised modes-power load transients. A dedicated program was carried out at the JET to study the physics and consequences of W transient melting. Following initial exposures in 2013 (ILW-1) of a W-lamella with leading edge, new experiments have been performed on a sloped surface (15 degrees slope) during the 2015/2016 (ILW-3) campaign. This new experiment allows significantly improved infrared thermography measurements and thus resolved important issue of power loading in the context of the previous leading edge exposures. The new lamella was monitored by local diagnostics: spectroscopy, thermography and high-resolution photography in between discharges. No impact on the main plasma was observed despite a strong increase of the local W source consistent with evaporation. In contrast to the earlier exposure, no droplet emission was observed from the sloped surface. Topological modifications resulting from the melting are clearly visible between discharges on the photographic images. Melt damage can be clearly linked to the infrared measurements: the emissivity drops in zones where melting occurs. In comparison with the previous leading edge experiment, no runaway melt motion is observed, consistent with the hypothesis that the escape of thermionic electrons emitted from the melt zone is largely suppressed in this geometry, where the magnetic field intersects the surface at lower angles than in the case of perpendicular impact on a leading edge. Utilising both exposures allows us to further test the model of the forces driving melt motion that successfully reproduced the findings from the original leading edge exposure. Since the ILW-1 experiments, the exposed misaligned lamella has now been retrieved from the JET machine and post mortem analysis has been performed. No obvious mass loss is observed. Profilometry of the ILW-1 lamella shows the structure of the melt damage which is in line with the modell predictions thus allowing further model validation. Nuclear reaction analysis shows a tenfold reduction in surface deuterium concentration in the molten surface in comparison to the non-molten part of the lamella.

  • 20. Esser, H. G.
    et al.
    Philipps, V.
    Freisinger, M.
    Widdowson, A.
    Heinola, K.
    Kirschner, A.
    Moeller, S.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Brezinsek, S.
    Huber, A.
    Matthews, G. F.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Sergienko, G.
    Material deposition on inner divertor quartz-micro balances during ITER-like wall operation in JET2015Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, s. 796-799Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The migration of beryllium, tungsten and carbon to remote areas of the inner JET-ILW divertor and the accompanying co-deposition of deuterium has been investigated using post-mortem analysis of the housings of quartz-micro balances (QMBs) and their quartz crystals. The analysis of the deposition provides that the rate of beryllium atoms is significantly reduced compared to the analogue deposition rate of carbon during the carbon wall conditions (JET-C) at the same locations of the QMBs. A reduction factor of 50 was found at the entrance gap to the cryo-pumps while it was 14 under tile 5, the semi-horizontal target plate. The deposits consist of C/Be atomic ratios of typically 0.1-0.5 showing an enrichment of carbon in remote areas compared to directly exposed areas with less carbon. The deuterium retention fraction D/Be is between 0.3 and 1 at these unheated locations in the divertor.

  • 21.
    Fazinic, Stjepko
    et al.
    Rudjer Boskovic Inst, Bijenicka 54, Zagreb 10000, Croatia..
    Tadic, Tonic
    Rudjer Boskovic Inst, Bijenicka 54, Zagreb 10000, Croatia..
    Vuksic, Marin
    Rudjer Boskovic Inst, Bijenicka 54, Zagreb 10000, Croatia..
    Rubel, Marek
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Fortuna-Zalesna, Elibieta
    Warsaw Univ Technol, Fac Mat Sci & Technol, Woloska 141, PL-02507 Warsaw, Poland..
    Widdowson, Anna
    Culham Sci Ctr, Culham Ctr Fus Energy, Abingdon OX14 3DB, Oxon, England..
    Ion Microbeam Analyses of Dust Particles and Codeposits from JET with the ITER-Like Wall2018Ingår i: Analytical Chemistry, ISSN 0003-2700, E-ISSN 1520-6882, Vol. 90, nr 9, s. 5744-5752Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Generation of metal dust in the JET tokamak with the ITER-like wall (ILW) is a topic of vital interest to next-step fusion devices because of safety issues with plasma operation. Simultaneous Nuclear Reaction Analysis (NRA) and Particle Induced X-ray Emission (PIXE) with a focused four MeV He-3 microbeam was used to determine the composition of dust particles related to the JET operation with the ILW. The focus was on "Be-rich particles" collected from the deposition zone on the inner divertor tile. The particles found are composed of a mix of codeposited species up to 120 m in size with a thickness of 30-40 mu m, The main constituents are D from the fusion fuel, Be and W from the main plasma-facing components, and Ni and Cr from the Inconel grills of the antennas for auxiliary plasma heating. Elemental concentrations were estimated by iterative NRA-PIXE analysis. Two types of dust particles were found: (i) larger Be-rich particles with Be concentrations above 90 at% with a deuterium presence of up to 3.4 at% and containing Ni (1-3 at%), Cr (0.4-0.8 at%), W (0.2-0.9 at%), Fe (0.3-0.6 at%), and Cu and Ti in lower concentrations and (ii) small particles rich in Al and/or Si that were in some cases accompanied by other elements, such as Fe, Cu, or Ti or W and Mo.

  • 22. Fortuna-Zalesna, E.
    et al.
    Grzonka, J.
    Moon, Sunwoo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Widdowson, A.
    Fine metal dust particles on the wall probes from JET-ILW2017Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, artikel-id 014038Artikel i tidskrift (Refereegranskat)
    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.

  • 23.
    Garcia Carrasco, Alvaro
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Möller, S.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Ivanova, Darya
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Kreter, A.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Wauters, T.
    Impact of ion cyclotron wall conditioning on fuel removal from plasma-facing components at TEXTOR2014Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T159, s. 014017-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

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

  • 24.
    Garcia Carrasco, Alvaro
    et al.
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Widdowson, A.
    Fortuna-Zalesna, E.
    Jachmich, S.
    Brix, M.
    Marot, L.
    Plasma impact on diagnostic mirrors in JET2017Ingår i: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, s. 506-512Artikel i tidskrift (Refereegranskat)
    Abstract [en]

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

  • 25.
    Garcia Carrasco, Alvaro
    et al.
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Schwarz-Selinger, T.
    Wauters, T.
    Douai, D.
    Bobkov, V.
    Cavazzana, R.
    Krieger, K.
    Lyssoivan, A.
    Moeller, S.
    Spolaore, M.
    Rohde, V.
    Rubel, M.
    Investigation of probe surfaces after ion cyclotron wall conditioning in ASDEX upgrade2017Ingår i: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, s. 733-735Artikel i tidskrift (Refereegranskat)
    Abstract [en]

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

  • 26.
    Garcia Carrasco, Alvaro
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Wauters, T.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Drenik, A.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Crombé, K.
    Douai, D.
    Fortuna, E.
    Kogut, D.
    Kreter, A.
    Lyssoivan, A.
    Möller, S.
    Pisarek, M.
    Vervier, M.
    Nitrogen removal from plasma-facing components by ion cyclotron wall conditioning in TEXTOR2015Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, s. 688-692Artikel i tidskrift (Övrigt vetenskapligt)
    Abstract [en]

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

  • 27.
    Garcia-Carrasco, Alvaro
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Hallén, Anders
    KTH, Skolan för informations- och kommunikationsteknik (ICT), Integrerade komponenter och kretsar.
    Grzonka, J.
    Gilbert, M. R.
    Fortuna-Zalesna, E.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Impact of helium implantation and ion-induced damage on reflectivity of molybdenum mirrors2016Ingår i: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584Artikel i tidskrift (Refereegranskat)
    Abstract [en]

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

  • 28. Hakola, A.
    et al.
    Airila, M. I.
    Björkas, C.
    Borodin, D.
    Brezinsek, S.
    Coad, J. P.
    Groth, M.
    Järvinen, A.
    Kirschner, A.
    Koivuranta, S.
    Krieger, K.
    Kurki-Suonio, T.
    Likonen, J.
    Lindholm, V.
    Makkonen, T.
    Mayer, M.
    Miettunen, J.
    Mueller, H. W.
    Neu, R.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Rohde, V.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Widdowson, A.
    Global migration of impurities in tokamaks2013Ingår i: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 55, nr 12Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The migration of impurities in tokamaks has been studied with the help of tracer-injection (C-13 and N-15) experiments in JET and ASDEX Upgrade since 2001. We have identified a common pattern for the migrating particles: scrape-off layer flows drive impurities from the low-field side towards the high-field side of the vessel. Migration is also sensitive to the density and magnetic configuration of the plasma, and strong local variations in the resulting deposition patterns require 3D treatment of the migration process. Moreover, re-erosion of the deposited particles has to be taken into account to properly describe the migration process during steady-state operation of the tokamak.

  • 29. Hakola, A.
    et al.
    Brezinsek, S.
    Douai, D.
    Balden, M.
    Bobkov, V.
    Carralero, D.
    Greuner, H.
    Elgeti, S.
    Kallenbach, A.
    Krieger, K.
    Meisl, G.
    Oberkofler, M.
    Rohde, V.
    Schneider, P.
    Schwarz-Selinger, T.
    Lahtinen, A.
    De Temmerman, G.
    Caniello, R.
    Ghezzi, F.
    Wauters, T.
    Garcia Carrasco, Alvaro
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Radovic, I. Bogdanovic
    Siketic, Z.
    Plasma-wall interaction studies in the full-W ASDEX upgrade during helium plasma discharges2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 6, artikel-id 066015Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Plasma-wall interactions have been studied in the full-W ASDEX Upgrade during its dedicated helium campaign. Relatively clean plasmas with a He content of > 80% could be obtained by applying ion cyclotron wall conditioning (ICWC) discharges upon changeover from D to He. However, co-deposited layers with significant amounts of He and D were measured on W samples exposed to ICWC plasmas at the low-field side (outer) midplane. This is a sign of local migration and accumulation of materials and residual fuel in regions shadowed from direct plasma exposure albeit globally D was removed from the vessel. When exposing W samples to ELMy H-mode helium plasmas in the outer strike-point region, no net erosion was observed but the surfaces had been covered with co-deposited layers mainly consisting of W, B, C, and D and being the thickest on rough and modified surfaces. This is different from the typical erosion-deposition patterns in D plasmas, where usually sharp net-erosion peaks surrounded by prominent net-deposition maxima for W are observed close to the strike point. Moreover, no clear signs of W nanostructure growth or destruction could be seen. The growth of deposited layers may impact the operation of future fusion reactors and is attributed to strong sources in the main chamber that under suitable conditions may switch the balance from net erosion into net deposition, even close to the strike points. In addition, the absence of noticeable chemical erosion in helium plasmas may have affected the thickness of the deposited layers. Retention of He, for its part, remained small and uniform throughout the strike-point region although our results indicate that samples with smooth surfaces can contain an order of magnitude less He than their rough counterparts.

  • 30. Hakola, A.
    et al.
    Koivuranta, S.
    Likonen, J.
    Groth, M.
    Kurki-Suonio, T.
    Lindholm, V.
    Miettunen, J.
    Krieger, K.
    Mayer, M.
    Müller, H. W.
    Neu, R.
    Rohde, V.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Global migration of C-13 impurities in high-density L-mode plasmas in ASDEX Upgrade2013Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 438, nr Suppl., s. S694-S697Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    We have studied the migration of 13C in ASDEX Upgrade after a global impurity injection experiment in 2011. The main chamber was observed to be the largest deposition region for carbon: almost 35% of the injected atoms end up there. Moreover, gaps between wall tiles account for surface densities which are comparable to those on the plasma-facing surfaces. SOLPS modeling of the experiment produced a set of background plasmas and poloidal flow profiles for simulating the transport of 13C with ASCOT; a match with measured deposition, however, required using an imposed flow profile. ASCOT reproduced the observed localized deposition at the outer midplane but work is needed to explain the measured deposition at the inner side of the torus and at the top of the vessel.

  • 31. Heinola, K.
    et al.
    Widdowson, A.
    Likonen, J.
    Alves, E.
    Baron-Wiechec, A.
    Barradas, N.
    Brezinsek, S.
    Catarino, N.
    Coad, P.
    Koivuranta, S.
    Krat, S.
    Matthews, G. F.
    Mayer, M.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. EUROfusion Consortium, Culham Science Centre, JET, Abingdon, United Kingdom.
    Contributors, J.
    Long-term fuel retention in JET ITER-like wall2016Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T167, artikel-id 014075Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Post-mortem studies with ion beam analysis, thermal desorption, and secondary ion mass spectrometry have been applied for investigating the long-term fuel retention in the JET ITERlike wall components. The retention takes place via implantation and co-deposition, and the highest retention values were found to correlate with the thickness of the deposited impurity layers. From the total amount of retained D fuel over half was detected in the divertor region. The majority of the retained D is on the top surface of the inner divertor, whereas the least retention was measured in the main chamber on the mid-plane of the inner wall limiter. The recessed areas of the inner wall showed significant contribution to the main chamber total retention. Thermal desorption spectroscopy analysis revealed the energetic T from DD reactions being implanted in the divertor. The total T inventory was assessed to be >0.3 mg.

  • 32. Heinola, K.
    et al.
    Widdowson, A.
    Likonen, J.
    Alves, E.
    Baron-Wiechec, A.
    Barradas, N.
    Brezinsek, S.
    Catarino, N.
    Coad, P.
    Koivuranta, S.
    Matthews, G. F.
    Mayer, M.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Fuel retention in JET ITER-Like Wall from post-mortem analysis2015Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, s. 961-965Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Selected Ion Beam Analysis techniques applicable for detecting deuterium and heavier impurities have been used in the post-mortem analyses of tiles removed after the first JET ITER-Like Wall (JET-ILW) campaign. Over half of the retained fuel was measured in the divertor region. The highest figures for fuel retention were obtained from regions with the thickest deposited layers, i.e. in the inner divertor on top of tile 1 and on the High Field Gap Closure tile, which resides deep in the plasma scrape-off layer. Least retention was found in the main chamber high erosion regions, i.e. in the mid-plane of Inner Wall Guard Limiter. The fuel retention values found typically varied with deposition layer thicknesses. The reported retention values support the observed decrease in fuel retention obtained with gas balance experiments of JET-ILW.

  • 33.
    Ivanova, Darya
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Likonen, Jari
    Widdowson, Anna
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    De Temmerman, Gregory
    Assessment of Cleaning Methods for First Mirrors Tested in JET for ITER2013Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 438, nr Suppl., s. S1241-S1244Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Two cleaning techniques were used for removal of co-deposits from the tested first mirrors exposed in JET: (a) ultrasonic bath; (b) a broad range of polishing conditions from manual buffing to machine polishing with the diamond grain size of up to 3 lm. Reflectivity measurements were performed after each step in the cleaning procedure. Surfaces were also examined with electron microscopy and ion beam analysis methods. Ultrasonic cleaning leads to partial recovery of reflectivity due to enhanced detachment of deposits. Typically 30-50% of the original reflectivity was recovered in the visible light and 50-90% in the infrared region. One mirror was cleaned completely. Polishing with diamond paste may lead to successful removal of deposits but the damage to the surface in case of the large diamond grains was observed. Recovery of up to 100% of the initial reflectivity was achieved for some mirrors.

  • 34.
    Ivanova, Darya
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Kreter, A.
    Möller, S.
    Philipps, V.
    Freisinger, M.
    Wauters, T.
    Impact of thermal treatment and ICWC on fuel inventory in Co-deposits2012Ingår i: 39th EPS Conference on Plasma Physics 2012, EPS 2012 and the 16th International Congress on Plasma Physics: volume 1, 2012, s. 317-320Konferensbidrag (Refereegranskat)
  • 35.
    Ivanova, Darya
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Philipps, V.
    Institute of Energy Research, Forschungszentrum Jülich.
    Freisinger, M.
    Institute of Energy Research, Forschungszentrum Jülich.
    Dittmar, T.
    CEA-Cadarache, IRFM.
    Huber, A.
    Institute of Energy Research, Forschungszentrum Jülich.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Schweer, B.
    Institute of Energy Research, Forschungszentrum Jülich.
    Sergienko, G.
    Institute of Energy Research, Forschungszentrum Jülich.
    Wessel, E.
    Institute of Energy Research, Forschungszentrum Jülich.
    Dust Particles in Controlled Fusion Devices: Generation Mechanism and Analysis2010Konferensbidrag (Refereegranskat)
    Abstract [en]

    Generation and in-vessel accumulation of carbon and metal dust are perceived to be serious safety andeconomy issues for a steady-state operation of a fusion reactor, e.g. ITER. This contribution provides acomprehensive account on: (a) properties of carbon and metal dust formed in the TEXTOR tokamak; (b) dustgeneration associated with removal of fuel and co-deposit from carbon PFC from TEXTOR and Tore Supra; (c)surface morphology of wall components after different cleaning treatments. The amount of loose dust found on thefloor of the TEXTOR liner does not exceed 2 grams with particle size range 0.1 m – 1 mm. The presence of fine(up to 1 m) crystalline graphite in the collected matter suggests that brittle destruction of carbon PFC could takeplace during off-normal events. Carbon is the main component, but there are also magnetic and non-magnetic metalagglomerates. The fuel content in dust and co-deposits varies from 10% on the main limiters to 0.03% on theneutralizer plates. Fuel removal by oxidative methods or by annealing in vacuum disintegrates co-deposits and, in thecase of thick layers, makes them brittle thus reducing the adherence to the target. Also photonic cleaning by laserpulses produces debris, especially under ablation conditions. The results obtained strongly indicate that in a carbonwall machine the disintegration of flaking co-deposits on PFC is the main source of dust.

  • 36.
    Ivanova, Darya
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Widdowson, A.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Likonen, J.
    Marot, L.
    Alves, E.
    Garcia-Carrasco, Alvaro
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Pintsuk, G.
    An overview of the comprehensive First Mirror Test in JET with ITER-like wall2014Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T159, s. 014011-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

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

  • 37. Lagoyannis, A.
    et al.
    Tsavalas, P.
    Mergia, K.
    Provatas, G.
    Triantou, K.
    Tsompopoulou, E.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. EUROfusion Consortium, United Kingdom.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. EUROfusion Consortium, United Kingdom.
    Widdowson, A.
    Harissopulos, S.
    Mertzimekis, T. J.
    Surface composition and structure of divertor tiles following the JET tokamak operation with the ITER-like wall2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 7, artikel-id 076027Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Samples extracted from several divertor tiles following the 2011-2012 operation of JET with the ITER-Like wall were analyzed using ion beam analysis methods, x-ray fluorescence spectroscopy, scanning electron microscopy with energy dispersive spectroscopy analysis and x-ray diffraction. The emphasis was on the determination of light species and on material mixing including compound formation on the bottom and the outer divertor tiles. Deposition of deuterium, beryllium, carbon, nitrogen, oxygen, iron, chromium, nickel and molybdenum has been detected on all studied tiles. The thickest deposition, of around 4 μm, was measured on the bottom of the outer divertor, whereas the other surfaces (inner bottom and vertical outer) the co-deposits were around 1 μm. x-ray diffraction measurements have revealed the formation of the compound W2C on all specimens.

  • 38. Likonen, J.
    et al.
    Alves, E.
    Baron-Wiechec, A.
    Brezinsek, S.
    Coad, J. P.
    Hakola, A.
    Heinola, K.
    Koivuranta, S.
    Matthews, G. F.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Stan-Sion, C.
    Widdowson, A.
    First results and surface analysis strategy for plasma-facing components after JET operation with the ITER-like wall2014Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T159, s. 014016-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    During the carbon wall operations of JET since 2001, an extensive post-mortem analysis programme has been carried out under the JET Task Force Fusion Technology and a similar analysis programme is underway for the JET-ILW tiles removed during the 2012 shutdown. The first post-mortem results from the JET ITER-like wall tiles have shown that the overall amount of deposition on the divertor tiles and on remote divertor areas has been reduced by more than an order of magnitude with respect to JET-C. In addition, the obtained data indicate a possible interaction between Be and W such as the formation of mixed Be-W layers. This could be due to the surface roughness of the tiles, or could be caused by diffusion or even alloying. Ion-beam analyses and secondary ion mass spectrometry techniques give only elemental information, so other techniques such as x-ray diffraction, x-ray photoelectron spectroscopy, secondary electron microscopy/energy dispersive x-ray spectroscopy and nuclear microprobing are required. Since the nature of deposition and erosion has changed during the JET-ILW operations, a change in the post-mortem analysis programme is needed. For example, no cross-sectional samples from the sloping parts of tiles 4 and 6 are required. A strategy for post-mortem analyses of the marker-coated tiles will be presented in this paper.

  • 39. Litnovsky, A.
    et al.
    Hellwig, M.
    Matveev, D.
    Komm, M.
    van den Berg, M.
    De Temmerman, G.
    Rudakov, D.
    Ding, F.
    Luo, G. -N
    Krieger, K.
    Sugiyama, K.
    Pitts, R. A.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Optimization of tungsten castellated structures for the ITER divertor2015Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, s. 174-179Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    In ITER, the plasma-facing components (PFCs) of the first wall and the divertor armor will be castellated to improve their thermo-mechanical stability and to limit forces due to induced currents. The fuel accumulation in the gaps may significantly contribute to the in-vessel fuel inventory. Castellation shaping may be the most straightforward way to minimize the fuel inventory and to alleviate the thermal loads onto castellations. A new castellation shape was proposed and comparative modeling of conventional (rectangular) and shaped castellation was performed for ITER conditions. Shaped castellation was predicted to be capable to operate under stationary heat load of 20 MW/m(2). An 11-fold decrease of beryllium (Be) content in the gaps of the shaped cells alone with a 7-fold decrease of carbon content was predicted. In order to validate the predictive capabilities of modeling tools used for ITER conditions, the dedicated modeling with the same codes was made for existing tokamaks and benchmarked with the results of multi-machine experiments. For the castellations exposed in TEXTOR and DIII-D, the carbon amount in the gaps of shaped cells was 1.9-2.3 times smaller than that of rectangular ones. Modeling for TEXTOR conditions yielded to 1.5-fold decrease of carbon content in the gaps of shaped castellation outlining fair agreement with the experiment. At the same time, a number of processes, like enhanced erosion of molten layer yet need to be implemented in the codes in order to increase the accuracy of predictions for ITER.

  • 40. Lituadon, Xavier
    et al.
    Bergsåker, Henric
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bykov, Igor
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Garcia-Carrasco, Alvaro
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Hellsten, Torbjörn
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Jonsson, Thomas
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Stefániková, Estera
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Tholerus, Emmi
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Vallejos Olivares, Pablo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Weckmann, Armin
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Zhou, Yushan
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ström, Petter
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    et al.,
    Overview of the JET results in support to ITER2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 10, artikel-id 102001Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H  =  1 at β N ~ 1.8 and n/n GW ~ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D–T campaign and 14 MeV neutron calibration strategy are reviewed.

  • 41. Louche, F.
    et al.
    Wauters, T.
    Ragona, R.
    Moeller, S.
    Durodie, F.
    Litnovsky, A.
    Lyssoivan, A.
    Messiaen, A.
    Ongena, J.
    Petersson, Per
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Brezinsek, S.
    Linsmeier, Ch.
    Van Schoor, M.
    Design of an ICRF system for plasma-wall interactions and RF plasma production studies on TOMAS2017Ingår i: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 123, s. 317-320Artikel i tidskrift (Refereegranskat)
    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.

  • 42. Malaquias, A.
    et al.
    Philipps, V.
    Huber, A.
    Hakola, A.
    Likonen, J.
    Kolehmainen, J.
    Tervakangas, S.
    Aints, M.
    Paris, P.
    Laan, M.
    Lissovski, A.
    Almaviva, S.
    Caneve, L.
    Colao, F.
    Maddaluno, G.
    Kubkowska, M.
    Gasior, P.
    Van Der Meiden, H. J.
    Lof, A. R.
    Zeijlmans Van Emmichoven, P. A.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Fortuna, E.
    Xiao, Q.
    Development of ITER relevant laser techniques for deposited layer characterisation and tritium inventory2013Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 438, nr Suppl., s. S936-S939Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Laser Induced Breakdown Spectroscopy (LIBS) is a potential candidate to monitor the layer composition and fuel retention during and after plasma shots on specific locations of the main chamber and divertor of ITER. This method is being investigated in a cooperative research programme on plasma devices such as TEXTOR, FTU, MAGNUM-PSI and in other various laboratorial experiments. In this paper LIBS results from targets of D-H-rich carbon films and mixed W-Al-C deposits on bulk tungsten substrates are reported (simulating ITER-like deposits with Al as proxy for Be). Two independent methods, one to determine the relative elemental composition and the other the absolute contents of the target based on the experimental LIBS signals are proposed. The results show that LIBS has the capability to provide the relative concentrations of the elements on the deposited layer when the experimental conditions on the targets surface are identical to the calibration samples.

  • 43. Matejicek, Jiri
    et al.
    Weinzettl, Vladimir
    Mackova, Anna
    Malinsky, Petr
    Havranek, Vladimir
    Naydenkova, Diana
    Klevarova, Veronika
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Gasior, Pawel
    Hakola, Antti
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Fortuna, Elzbieta
    Kolehmainen, Jukka
    Tervakangas, Sanna
    Interaction of candidate plasma facing materials with tokamak plasma in COMPASS2017Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 493, s. 102-119Artikel i tidskrift (Refereegranskat)
    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.

  • 44. Meyer, H.
    et al.
    Eich, T.
    Beurskens, M.
    Coda, S.
    Hakola, A.
    Martin, P.
    Adamek, J.
    Agostini, M.
    Aguiam, D.
    Ahn, J.
    Aho-Mantila, L.
    Akers, R.
    Albanese, R.
    Aledda, R.
    Alessi, E.
    Allan, S.
    Alves, D.
    Ambrosino, R.
    Amicucci, L.
    Anand, H.
    Anastassiou, G.
    Andrebe, Y.
    Angioni, C.
    Apruzzese, G.
    Ariola, M.
    Arnichand, H.
    Arter, W.
    Baciero, A.
    Barnes, M.
    Barrera, L.
    Behn, R.
    Bencze, A.
    Bernardo, J.
    Bernert, M.
    Bettini, P.
    Bilkova, P.
    Bin, W.
    Birkenmeier, G.
    Bizarro, J. P. S.
    Blanchard, P.
    Blanken, T.
    Bluteau, M.
    Bobkov, V.
    Bogar, O.
    Boehm, P.
    Bolzonella, T.
    Boncagni, L.
    Botrugno, A.
    Bottereau, C.
    Bouquey, F.
    Bourdelle, C.
    Bremond, S.
    Brezinsek, S.
    Brida, D.
    Brochard, F.
    Buchanan, J.
    Bufferand, H.
    Buratti, P.
    Cahyna, P.
    Calabro, G.
    Camenen, Y.
    Caniello, R.
    Cannas, B.
    Canton, A.
    Cardinali, A.
    Carnevale, D.
    Carr, M.
    Carralero, D.
    Carvalho, P.
    Casali, L.
    Castaldo, C.
    Castejon, F.
    Castro, R.
    Causa, F.
    Cavazzana, R.
    Cavedon, M.
    Cecconello, M.
    Ceccuzzi, S.
    Cesario, R.
    Challis, C. D.
    Chapman, I. T.
    Chapman, S.
    Chernyshova, M.
    Choi, D.
    Cianfarani, C.
    Ciraolo, G.
    Citrin, J.
    Clairet, F.
    Classen, I.
    Coelho, R.
    Coenen, J. W.
    Colas, L.
    Conway, G.
    Corre, Y.
    Costea, S.
    Crisanti, F.
    Cruz, N.
    Cseh, G.
    Czarnecka, A.
    D'Arcangelo, O.
    De Angeli, M.
    De Masi, G.
    De Temmerman, G.
    De Tommasi, G.
    Decker, J.
    Delogu, R. S.
    Dendy, R.
    Denner, P.
    Di Troia, C.
    Dimitrova, M.
    D'Inca, R.
    Doric, V.
    Douai, D.
    Drenik, A.
    Dudson, B.
    Dunai, D.
    Dunne, M.
    Duval, B. P.
    Easy, L.
    Elmore, S.
    Erdos, B.
    Esposito, B.
    Fable, E.
    Faitsch, M.
    Fanni, A.
    Fedorczak, N.
    Felici, F.
    Ferreira, J.
    Fevrier, O.
    Ficker, O.
    Fietz, S.
    Figini, L.
    Figueiredo, A.
    Fil, A.
    Fishpool, G.
    Fitzgerald, M.
    Fontana, M.
    Ford, O.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Fridstr, R.
    Frigione, D.
    Fuchert, G.
    Fuchs, C.
    Palumbo, M. Furno
    Futatani, S.
    Gabellieri, L.
    Galazka, K.
    Galdon-Quiroga, J.
    Galeani, S.
    Gallart, D.
    Gallo, A.
    Galperti, C.
    Gao, Y.
    Garavaglia, S.
    Garcia, J.
    Garcia-Carrasco, Alvaro
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Garcia-Lopez, J.
    Garcia-Munoz, M.
    Gardarein, J. -L
    Garzotti, L.
    Gaspar, J.
    Gauthier, E.
    Geelen, P.
    Geiger, B.
    Ghendrih, P.
    Ghezzi, F.
    Giacomelli, L.
    Giannone, L.
    Giovannozzi, E.
    Giroud, C.
    Gleason Gonzalez, C.
    Gobbin, M.
    Goodman, T. P.
    Gorini, G.
    Gospodarczyk, M.
    Granucci, G.
    Gruber, M.
    Gude, A.
    Guimarais, L.
    Guirlet, R.
    Gunn, J.
    Hacek, P.
    Hacquin, S.
    Hall, S.
    Ham, C.
    Happel, T.
    Harrison, J.
    Harting, D.
    Hauer, V.
    Havlickova, E.
    Hellsten, Torbjörn
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Helou, W.
    Henderson, S.
    Hennequin, P.
    Heyn, M.
    Hnat, B.
    Holzl, M.
    Hogeweij, D.
    Honore, C.
    Hopf, C.
    Horacek, J.
    Hornung, G.
    Horvath, L.
    Huang, Z.
    Huber, A.
    Igitkhanov, J.
    Igochine, V.
    Imrisek, M.
    Innocente, P.
    Ionita-Schrittwieser, C.
    Isliker, H.
    Ivanova-Stanik, I.
    Jacobsen, A. S.
    Jacquet, P.
    Jakubowski, M.
    Jardin, A.
    Jaulmes, F.
    Jenko, F.
    Jensen, T.
    Busk, O. Jeppe Miki
    Jessen, M.
    Joffrin, E.
    Jones, O.
    Jonsson, Thomas
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Kallenbach, A.
    Kallinikos, N.
    Kalvin, S.
    Kappatou, A.
    Karhunen, J.
    Karpushov, A.
    Kasilov, S.
    Kasprowicz, G.
    Kendl, A.
    Kernbichler, W.
    Kim, D.
    Kirk, A.
    Kjer, S.
    Klimek, I.
    Kocsis, G.
    Kogut, D.
    Komm, M.
    Korsholm, S. B.
    Koslowski, H. R.
    Koubiti, M.
    Kovacic, J.
    Kovarik, K.
    Krawczyk, N.
    Krbec, J.
    Krieger, K.
    Krivska, A.
    Kube, R.
    Kudlacek, O.
    Kurki-Suonio, T.
    Labit, B.
    Laggner, F. M.
    Laguardia, L.
    Lahtinen, A.
    Lalousis, P.
    Lang, P.
    Lauber, P.
    Lazanyi, N.
    Lazaros, A.
    Le, H. B.
    Lebschy, A.
    Leddy, J.
    Lefevre, L.
    Lehnen, M.
    Leipold, F.
    Lessig, A.
    Leyland, M.
    Li, L.
    Liang, Y.
    Lipschultz, B.
    Liu, Y. Q.
    Loarer, T.
    Loarte, A.
    Loewenhoff, T.
    Lomanowski, B.
    Loschiavo, V. P.
    Lunt, T.
    Lupelli, I.
    Lux, H.
    Lyssoivan, A.
    Madsen, J.
    Maget, P.
    Maggi, C.
    Maggiora, R.
    Magnussen, M. L.
    Mailloux, J.
    Maljaars, B.
    Malygin, A.
    Mantica, P.
    Mantsinen, M.
    Maraschek, M.
    Marchand, B.
    Marconato, N.
    Marini, C.
    Marinucci, M.
    Markovic, T.
    Marocco, D.
    Marrelli, L.
    Martin, Y.
    Solis, J. R. Martin
    Martitsch, A.
    Mastrostefano, S.
    Mattei, M.
    Matthews, G.
    Mavridis, M.
    Mayoral, M. -L
    Mazon, D.
    McCarthy, P.
    McAdams, R.
    McArdle, G.
    McClements, K.
    McDermott, R.
    McMillan, B.
    Meisl, G.
    Merle, A.
    Meyer, O.
    Milanesio, D.
    Militello, F.
    Miron, I. G.
    Mitosinkova, K.
    Mlynar, J.
    Mlynek, A.
    Molina, D.
    Molina, P.
    Monakhov, I.
    Morales, J.
    Moreau, D.
    Morel, P.
    Moret, J. -M
    Moro, A.
    Moulton, D.
    Mueller, H. W.
    Nabais, F.
    Nardon, E.
    Naulin, V.
    Nemes-Czopf, A.
    Nespoli, F.
    Neu, R.
    Nielsen, A. H.
    Nielsen, S. K.
    Nikolaeva, V.
    Nimb, S.
    Nocente, M.
    Nouailletas, R.
    Nowak, S.
    Oberkofler, M.
    Oberparleiter, M.
    Ochoukov, R.
    Odstrcil, T.
    Olsen, J.
    Omotani, J.
    O'Mullane, M. G.
    Orain, F.
    Osterman, N.
    Paccagnella, R.
    Pamela, S.
    Pangione, L.
    Panjan, M.
    Papp, G.
    Paprok, R.
    Parail, V.
    Parra, I.
    Pau, A.
    Pautasso, G.
    Pehkonen, S. -P
    Pereira, A.
    Cippo, E. Perelli
    Ridolfini, V. Pericoli
    Peterka, M.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Petrzilka, V.
    Piovesan, P.
    Piron, C.
    Pironti, A.
    Pisano, F.
    Pisokas, T.
    Pitts, R.
    Ploumistakis, I.
    Plyusnin, V.
    Pokol, G.
    Poljak, D.
    Poloskei, P.
    Popovic, Z.
    Por, G.
    Porte, L.
    Potzel, S.
    Predebon, I.
    Preynas, M.
    Primc, G.
    Pucella, G.
    Puiatti, M. E.
    Putterich, T.
    Rack, M.
    Ramogida, G.
    Rapson, C.
    Rasmussen, J. Juul
    Rasmussen, J.
    Ratta, G. A.
    Ratynskaia, Svetlana
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ravera, G.
    Refy, D.
    Reich, M.
    Reimerdes, H.
    Reimold, F.
    Reinke, M.
    Reiser, D.
    Resnik, M.
    Reux, C.
    Ripamonti, D.
    Rittich, D.
    Riva, G.
    Rodriguez-Ramos, M.
    Rohde, V.
    Rosato, J.
    Ryter, F.
    Saarelma, S.
    Sabot, R.
    Saint-Laurent, F.
    Salewski, M.
    Salmi, A.
    Samaddar, D.
    Sanchis-Sanchez, L.
    Santos, J.
    Sauter, O.
    Scannell, R.
    Scheffer, M.
    Schneider, M.
    Schneider, B.
    Schneider, P.
    Schneller, M.
    Schrittwieser, R.
    Schubert, M.
    Schweinzer, J.
    Seidl, J.
    Sertoli, M.
    Sesnic, S.
    Shabbir, A.
    Shalpegin, A.
    Shanahan, B.
    Sharapov, S.
    Sheikh, U.
    Sias, G.
    Sieglin, B.
    Silva, C.
    Silva, A.
    Fuglister, M. Silva
    Simpson, J.
    Snicker, A.
    Sommariva, C.
    Sozzi, C.
    Spagnolo, S.
    Spizzo, G.
    Spolaore, M.
    Stange, T.
    Pedersen, M. Stejner
    Stepanov, I.
    Stober, J.
    Strand, P.
    Susnjara, A.
    Suttrop, W.
    Szepesi, T.
    Tal, B.
    Tala, T.
    Tamain, P.
    Tardini, G.
    Tardocchi, M.
    Teplukhina, A.
    Terranova, D.
    Testa, D.
    Theiler, C.
    Thornton, A.
    Tolias, Panagiotis
    KTH, Skolan för elektro- och systemteknik (EES), Rymd- och plasmafysik.
    Tophoj, L.
    Treutterer, W.
    Trevisan, G. L.
    Tripsky, M.
    Tsironis, C.
    Tsui, C.
    Tudisco, O.
    Uccello, A.
    Urban, J.
    Valisa, M.
    Vallejos, Pablo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Valovic, M.
    Van den Brand, H.
    Vanovac, B.
    Varoutis, S.
    Vartanian, S.
    Vega, J.
    Verdoolaege, G.
    Verhaegh, K.
    Vermare, L.
    Vianello, N.
    Vicente, J.
    Viezzer, E.
    Vignitchouk, L.
    Vijvers, W. A. J.
    Villone, F.
    Viola, B.
    Vlahos, L.
    Voitsekhovitch, I.
    Vondracek, P.
    Vu, N. M. T.
    Wagner, D.
    Walkden, N.
    Wang, N.
    Wauters, T.
    Weiland, M.
    Weinzettl, V.
    Westerhof, E.
    Wiesenberger, M.
    Willensdorfer, M.
    Wischmeier, M.
    Wodniak, I.
    Wolfrum, E.
    Yadykin, D.
    Zagorski, R.
    Zammuto, I.
    Zanca, P.
    Zaplotnik, R.
    Zestanakis, P.
    Zhang, W.
    Zoletnik, S.
    Zuin, M.
    Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 10, artikel-id 102014Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n = 2 RMP maintaining good confinement H-H(98,H-y2) approximate to 0.95. Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes.

  • 45. Moeller, S.
    et al.
    Wauters, T.
    Kreter, A.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Carrasco, A. G.
    Fundamental processes of fuel removal by cyclotron frequency range plasmas and integral scenario for fusion application studied with carbon co-deposits2015Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, s. 1109-1112Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Plasma impact removal using radio frequency heated plasmas is a candidate method to control the co-deposit related tritium inventory in fusion devices. Plasma parameters evolve according to the balance of input power to losses (transport, radiation, collisions). Material is sputtered by the ion fluxes with impact energies defined by the plasma sheath. H-2, D-2 and O-18(2) plasmas are produced in the carbon limiter tokamak TEXTOR. Pre-characterised a-C:D layers are exposed to study local removal rates, The D-2 plasma exhibits the highest surface release rate of 5.7 +/- 0.9 * 10(19) D/m(2)s. Compared to this the rate of the O-2 plasma is 3-fold smaller due to its 11-fold lower ion flux density. Re-deposition of removed carbon is observed, indicating that pumping and ionisation are limiting the removal in TEXTOR. Presented models can explain the observations and allow tailoring removal discharges. An integral application scenario using ICWC and thermo-chemical removal is presented, allowing to remove 700 g T from a-C:DT co-deposits in 20 h with fusion compatible wall conditions using technical specifications similar to ITER.

  • 46. Moser, L.
    et al.
    Marot, L.
    Steiner, R.
    Newman, M.
    Widdowson, A.
    Ivanova, Darya
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Likonen, J.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Pintsuk, G.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Meyer, E.
    Plasma cleaning of beryllium coated mirrors2016Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T167, artikel-id 014069Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Cleaning systems of metallic first mirrors are needed in more than 20 optical diagnostic systems from ITER to avoid reflectivity losses. Currently, plasma sputtering is considered as one of the most promising techniques to remove deposits coming from the main wall (mainly beryllium and tungsten). This work presents the results of plasma cleaning of rhodium and molybdenum mirrors exposed in JET-ILW and contaminated with typical tokamak elements (including beryllium and tungsten). Using radio frequency (13.56 MHz) argon or helium plasma, the removal of mixed layers was demonstrated and mirror reflectivity improved towards initial values. The cleaning was evaluated by performing reflectivity measurements, scanning electron microscopy, x-ray photoelectron spectroscopy and ion beam analysis.

  • 47. Möller, S.
    et al.
    Alegre, D.
    Kreter, A.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Esser, H. G.
    Samm, U.
    Thermo-chemical fuel removal from porous materials by oxygen and nitrogen dioxide2014Ingår i: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T159, s. 014065-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Thermo-chemical removal (TCR), or baking in reactive gases, is a candidate method to control the co-deposit related tritium inventory in fusion devices. TCR can be understood as reaction-diffusion processes in a porous material. O-2-TCR was applied to 150-550 nm thick a-C:D layers with similar textures. A linear relation between the integral TCR rate and the layer thickness, as predicted by the understanding, was observed in the experiment, i.e. the time to remove the hydrogen inventory is independent of its initial amount. TCR with nitrogen dioxide (NO2) at temperatures of 200-350 degrees C was conducted with a set of a-C:D and W-C-H layers. At 350 degrees C NO2 removed similar to 15% porosity a-C:D within 3 min. The O retention in remaining a-C:D was approximate to 10(17) Ocm(-2). An activation energy of approximate to 0.78 eV for reactions of NO2 with D and C was determined. The results were applied for predictions of the TCR effectivity in ITER. The treatment of W-C-H led to O uptake (O/W approximate to 2-3), while W and C contents remained unchanged.

  • 48. Paneta, V.
    et al.
    Fluch, U.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ott, S.
    Primetzhofer, D.
    Characterization of compositional modifications in metal-organic frameworks using carbon and alpha particle microbeams2017Ingår i: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 404, s. 198-201Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Zirconium-oxide based metal-organic frameworks (MOFs) were grown on p-type Si wafers. A modified linker molecule containing iodine was introduced by post synthetic exchange (PSE). Samples have been studied using Rutherford Backscattering Spectrometry (RBS) and Particle Induced X-ray Emission (PIXE) techniques, employing the 5 MV 15SDH-2 Pelletron Tandem accelerator at the Angstrom laboratory. The degree of post synthetic uptake of the iodine-containing linker has been investigated with both a broad beam and a focused beam of carbon and alpha particles targeting different kind of MOF crystals which were of similar to 1-10 mu m in size, depending on the linker used. Iodine concentrations in MOF crystallites were also measured by Nuclear Magnetic Resonance Spectroscopy (NMR) and are compared to the RBS results. In parallel to the ion beam studies, samples were investigated by Scanning Electron Microscopy (SEM) to quantify possible crystallite clustering, develop optimum sample preparation routines and to characterize the potential ion beam induced sample damage and its dependence on different parameters. Based on these results the reliability and accuracy of ion beam data is assessed.

  • 49.
    Petersson, Per
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Hakola, Antti
    Likonen, Jari
    Mayer, M.
    Miettunen, J.
    Neu, R.
    Rohde, V.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Injection of nitrogen-15 tracer into ASDEX-Upgrade: New technique in material migration studies2013Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 438, nr SUPPL., s. S616-S619Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    For the first time nitrogen-15 gas was used as a tracer for determining of global nitrogen retention in ASDEX-Upgrade. The injection done from the midplane gas inlet on the last day of the experimental campaign was followed by retrieval and ex situ analyses of many limiter and divertor tiles. The study was done by nuclear reaction analysis using the N-15(H-1,gamma He-4)C-12 process and detecting both gamma radiation and He-4. The highest and peaked concentrations of N-15, 8 x 10(16) cm(2), were found on limiters close to the injection point, while fairly homogeneous flat profiles were measured on most of the divertor plates. The measured concentrations are compared to an ASCOT simulation of the injection.

  • 50. Petersson, Per
    et al.
    Possnert, Göran
    Tandem Labb, Uppsala Universitet.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Dittmar, T.
    CEA Cadarache, France .
    Pégourié, B.
    CEA Cadarache, France .
    Tsitrone, E.
    CEA Cadarache, France .
    Wessel, E.
    CEA Cadarache, France .
    An Overview of Nuclear Microbeam Analysis of Surface and Bulk Fuel Retention in Carbon Fibre Composites from Tore Supra2011Ingår i: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 415, nr 1, s. S761-S764Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Surface and bulk retention of deuterium in tiles of the pump limiter from Tore Supra was examined with nuclear reaction analysis using both standard and micro-beam techniques. The aim was to determine the variations in the content and distribution of fuel species in carbon-fibre composites. On plasma-facing surfaces from the deposition zone, the D content reaches 2.5 × 1019 cm−2 in about 8 μm thick top layer, but lateral differences reach even more than one order of magnitude. This is also measured in the erosion zone: 6.6 × 1017 cm−2 to 7.7 × 1018 cm−2 D atoms. Bulk content was examined on cross-sections opened by fracturing the tiles. Fuel is detected up to the depth of 1–1.5 mm beneath the plasma-facing surface in tiles from both the erosion and deposition zones. It occurs in bands, about 100 μm wide and several mm long, roughly parallel to the original plasma-facing surface.

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