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  • 151.
    Rubel, Marek J.
    et al.
    KTH, Superseded Departments, Alfvén Laboratory.
    Coad, J. P.
    Bekris, N.
    Erents, S. K.
    Hole, D.
    Matthews, G. F.
    Penzhorn, R. D.
    Beryllium and carbon films in JET following D-T operation2003In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 313, p. 321-326Article in journal (Refereed)
    Abstract [en]

    After the D-T operation (DTE-1 campaign) at JET a large number of limiter and divertor tiles were dismounted from the torus for ex situ examination. The relative distributions of deuterium, tritium, beryllium and carbon are presented and discussed. Significant asymmetry observed in the distribution of erosion and deposition zones indicates preferential flow of the deuterium background plasma and impurities towards the inner divertor leg. The comparison of the beryllium content on the limiter tiles from the main chamber and the content of this element on the inner divertor tiles clearly proves the beryllium erosion from the main chamber wall and its transport to the divertor. However, no beryllium is detected in the shadowed regions of the divertor where the formation of thick and fuel-rich carbon films occurs. This is interpreted in terms of different mechanisms governing the erosion and transport of Be and C. The results allow a conclusion that the operation with a full beryllium wall would lead to a significantly decreased fuel inventory due to removal of the carbon source.

  • 152.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Coad, J. P.
    Hole, D.
    Accelerator-based ion beam analysis of fusion reactor materials2005In: Vacuum, ISSN 0042-207X, E-ISSN 1879-2715, Vol. 78, no 2-4, p. 255-261Article in journal (Refereed)
    Abstract [en]

    First wall components in controlled fusion devices undergo severe modification by various physical and chemical processes arising from plasma-material interactions: material erosion, its migration in the plasma and re-deposition. Ion beam analysis techniques play a prominent role in the studies of wall components exposed to hot plasmas. The intention of this work is to give a concise overview of methods used in the determination of surface composition of materials facing the plasma in the JET tokamak. The aim was to determine the amount and distribution (lateral and in-depth) of several elements or isotopes on large areas, i.e. mainly on high heat flux components such as limiter and divertor tiles. The major interest was in the quantification of deuterium co-deposited together with plasma impurity species: beryllium, boron, carbon (12 and 13) originating either from the wall erosion or deliberately introduced to the plasma as material transport markers. Special instrumentation, analytical approach and highlights of results obtained in studies of material transport by means of tracer techniques are presented.

  • 153.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Coad, J. P.
    Hole, D.
    Overview of long-term fuel inventory and co-deposition in castellated beryllium limiters at JET2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 386, p. 729-732Article in journal (Refereed)
    Abstract [en]

    Morphology of castellated Be tiles from the belt limiter exposed to the JET plasma for 56,000 s was examined on both sides of castellated grooves, on plasma-facing and side surfaces of the tiles. The essential results are (i) deuterium retention in the castellated grooves and in other locations is associated with co-deposition of carbon; (ii) the decay length of deposition in the castellation is around 1.5 mm; (iii) no deuterium is detected in bulk Be; (iv) bridging of gaps by molten beryllium occurred but gaps were not filled with Be; (v) on side surfaces of the tiles the formation of BeO layer was detected at a distance of 20 mm and more from the plasma-facing surface. The consequences for a long-term operation of a reactor-class device with several different plasma-facing materials are addressed.

  • 154.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Coad, J. P.
    Hole, D.
    Likonen, J.
    Vainonen-Ahlgren, E.
    Fuel retention in the gas box divertor of JET2005In: Fusion science and technology, ISSN 1536-1055, E-ISSN 1943-7641, Vol. 48, no 1, p. 569-572Article in journal (Refereed)
    Abstract [en]

    Components of the JET Mk-II Gas Box divertor have been analysed ex-situ after 18 months of operation with that divertor structure. The aim was to give an account on the distribution of the retained fuel along the poloidal cross-section of the divertor and, in particular, in the septum. Inside the gas box thick hydrogenated deposits were formed only on surfaces located in the near-plasma region from the inner divertor side whereas very little deposition was detected deep inside the gas box, i.e. on the support and divider plates.

  • 155.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Coad, J. P.
    Likonen, J.
    Matthews, G. F.
    Hole, D.
    Vainonen-Ahlgren, E.
    Pitts, R.
    Brezinsek, S.
    Coffey, I.
    Material migration studies at JET using tracer techniques2005Conference paper (Refereed)
  • 156.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Coad, J. P.
    Likonen, J.
    Philipps, V.
    Analysis of fuel retention in plasma-facing components from controlled fusion devices2009In: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 267, no 4, p. 711-717Article in journal (Refereed)
    Abstract [en]

    First wall components in controlled fusion devices undergo severe modification by various physical and chemical processes arising from plasma-wall interactions: material erosion, its transport in the plasma and re-deposition. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks. The importance of surface analysis in studies of fuel inventory and material migration is presented. Experimental procedures and analysis methods are briefly reviewed with emphasis on ion beam techniques which play a prominent role in studies of wall components exposed to hot plasmas. Practical aspects in the analytical approach are addressed and special instrumentation used in these studies is described.

  • 157.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Coad, J. P.
    Pitts, R. A.
    Overview of co-deposition and fuel inventory in castellated divertor structures at JET2007In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 367, p. 1432-1437Article in journal (Refereed)
    Abstract [en]

    The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 x 10(15) cm(-2). Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced.

  • 158.
    Rubel, Marek J.
    et al.
    KTH, Superseded Departments, Alfvén Laboratory.
    Coad, J. P.
    Wienhold, P.
    Matthews, G.
    Philipps, V.
    Stamp, M.
    Tanabe, T.
    Fuel inventory and co-deposition in grooves and gaps of divertor and limiter structures2004In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T111, p. 112-117Article in journal (Refereed)
    Abstract [en]

    Plasma facing components from JET and TEXTOR were studied. The emphasis was on the comparison of co-deposition, material mixing and fuel inventory on plasma facing and side surfaces of tiles, i.e. in gaps separating the tiles. Integrated fuel content in gaps of the Mk-I JET divertor floor was approximately two times greater than detected on the plasma facing surfaces. Taking into account similarities between the Mk-I structure and the castellation in the ITER divertor, the impact of the tile shaping on the tritium inventory is addressed. Deposition on the side of limiter tiles in TEXTOR was around 3-5% of that on the plasma facing surfaces. Experiments aiming at a deeper insight into the deposition on ITER-relevant components are also proposed.

  • 159.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    De Temmerman, G.
    Coad, J. P.
    Vince, J.
    Drake, James R.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Le Guern, F.
    Murari, A.
    Pitts, R. A.
    Walker, C.
    Jet-Efda Contributors,
    Mirror test for International Thermonuclear Experimental reactor at the JET tokamak: An overview of the program2006In: Review of Scientific Instruments, ISSN 0034-6748, E-ISSN 1089-7623, Vol. 77, no 6Article in journal (Refereed)
  • 160.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    De Temmerman, G.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Coad, J. P.
    Widdowson, A.
    Hole, D.
    Le Guern, F.
    Stamp, M.
    Vince, J.
    An overview of a comprehensive First Mirror Test for ITER at JET2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 390-91, p. 1066-1069Article in journal (Refereed)
    Abstract [en]

    The test was performed with 32 stainless steel and molybdenum mirrors placed in pan-pipe shaped cassettes and exposed in JET in the divertor and on the main chamber wall for 127000 s including 97000 s of X-point operation. Surface composition and total reflectivity were determined afterwards All mirrors. from the divertor were coated with deuterated carbon deposits causing the reflectivity loss by a factor of 6-10 in the visible range. Flaking and exfoliation of deposits were observed in some cases On the main. chamber wall the deposition occurred mainly on mirrors located deep in cassette channels whereas mirrors close to the channels entrances were free from deposits and retained fair reflectivity (similar to 90% of initial value) especially in the infra-red range. No significant differences in behaviour of steel and molybdenum were noted. The need for development of methods for mirror cleaning and/or protection in a reactor-class device is addressed.

  • 161.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    De Temmerman, Gregory
    University of Basel.
    Sergienko, Gennady
    Forschungszentrum Juelich.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Emmoth, Birger
    KTH, School of Information and Communication Technology (ICT), Microelectronics and Information Technology, IMIT.
    Philipps, Volker
    Forschungszentrum Juelich.
    Fuel removal from plasma-facing components by oxidation-assisted technique: An overview of surface morphology after oxidation2007In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 363-365, p. 877-881Article in journal (Refereed)
    Abstract [en]

    Oxygen-assistedfuelremoval is reported for laboratory-prepared a-C:D films and for layers obtained by boronisation in a tokamak and then exposed to a helium–oxygen glow discharge in TEXTOR. Oxidation of thick mixed-material co-deposits under laboratory conditions is also presented. The essential results are following: (i) laboratory-prepared amorphous deuterated carbon (a-C:D) layers are decomposed efficiently by the He–O2 glow: D and C contents are decreased by a factor of 45–220 and 25–60, respectively; (ii) the same treatment of the boronised films leads to the release of D but no removal of carbon is observed; (iii) the thermal oxidation (at 300 °C in air under laboratory conditions) of co-deposits on PFC and probes exposed to the SOL reduces the D content by a factor of 4–5 after 2 h, whereas nearly complete fuelremoval (98%) occurs after 10 h at 300 °C. The study shows that the fuelremoval efficiency is dependent on the overall composition of the mixed layer. It is high from pure a-C:D films but distinctly less efficient from real co-deposits.

  • 162.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Fortuna, E.
    Kreter, A.
    Wessel, E
    Philipps, V.
    Kurzydlowski, K. J.
    Overview of comprehensive characterisation of erosion zones on plasma facing components2006In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 81, no 1-7, p. 211-219Article in journal (Refereed)
    Abstract [en]

    Morphology of carbon plasma facing components retrieved from the TEXTOR tokamak after long operation periods and exposure to total particle doses exceeding 7 x 10(26) m(-2) was determined. Emphasis was on the composition and structure of the erosion zones. Tiles from two limiters-the main toroidal belt pump ALT-II and auxiliary inner bumper-were examined using high-resolution microscopy, surface profilometry, ion beam analysis techniques and energy dispersive X-ray spectroscopy. The essence of results regarding the net-erosion zones is following: (i) microstructure of surfaces is significantly smoother than on a non-exposed graphite, whereas carbon fibre composites show similar appearance prior to the exposure and after; (ii) deuterium retention is 2-5 x 10(21) m(-2); (iii) the presence of plasma impurity atoms (e.g. metals) is detected predominantly in small cavities acting as local shadowed areas on the surface. The results are discussed in terms of processes of material erosion/re-deposition and tokamak operation conditions influencing the morphology of wall components.

  • 163.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Ivanova, Darya
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Philipps, V.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Freisinger, M.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Fortuna, E.
    Warsaw University of Technology, Association Euratom - IPPLM, Poland.
    Huang, Z.
    KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Huber, A.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Linke, J.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Neubauer, O.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Penkalla, H.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Schweer, B.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Sergienko, G.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Wessel, E.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Dust particles in controlled fusion devices: generation mechanism and analysis2009In: 36th EPS Conference on Plasma Physics and Controlled Fusion, 2009, p. 129-132Conference paper (Refereed)
  • 164.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Ivanova-Stanik, IrenaInstitute for Plasma Physics, Warsaw, Poland.Miklaszewski, RyszardInstitute for Plasma Physics, Warsaw, Poland.Scholz, MarekInstitute for Plasma Physics, Warsaw, Poland.
    The Fifth International Workshop and Summer School on Plasma Physics2006Conference proceedings (editor) (Refereed)
  • 165.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Jacob, WolfgangMax-Planck-institute for Plasma physics, Garching, Germany.
    11th International Workshop on Plasma-Fcaing Materials and Components for Fusion Applications2007Conference proceedings (editor) (Refereed)
  • 166.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Kirschner, AndreasForschungszentrum Juelich, Germany.
    10th International Workshop on Carbon Materials for Fusion Applications2004Conference proceedings (editor) (Refereed)
  • 167.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Kreter, ArkadiForschungszentrum Juelich, Germany.
    12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications2009Conference proceedings (editor) (Refereed)
  • 168.
    Rubel, Marek J.
    et al.
    KTH, Superseded Departments, Alfvén Laboratory.
    Philipps, V.
    Pospieszczyk, A.
    Tanabe, T.
    Kotterl, S.
    Overview of fuel retention in composite and tungsten limiters2002In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 307, p. 111-115Article in journal (Refereed)
    Abstract [en]

    A number of materials, in a form of limiters, have been tested at the TEXTOR tokamak as candidates for plasma facing components in future devices. The results for fuel distribution and content in sole tungsten and composites (vacuum plasma sprayed tungsten layer on graphite and a B4C coating on copper) are reported and compared with earlier obtained results for sole tungsten and graphite. In sole tungsten the deuterium content is below 1 x 10(15) cm(-2), whereas in case of the composite targets the inventory found in the hottest parts of the limiters is one at least order of magnitude greater. The fuel content on composites is predominantly related to the co-deposition process with carbon and boron. The influence of material damage (melting, detachment of a coating) and material mixing processes on the fuel retention is also addressed.

  • 169.
    Rubel, Marek J.
    et al.
    KTH, Superseded Departments, Alfvén Laboratory.
    Philipps, V.
    Tanabe, T.
    Wienhold, P.
    Freisinger, M.
    Linke, J.
    von Seggern, J.
    Wessel, E.
    Thick co-deposits and dust in controlled fusion devices with carbon walls: Fuel inventory and growth rate of co-deposited layers2003In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T103, p. 20-24Article in journal (Refereed)
    Abstract [en]

    Recent results regarding the formation of co-deposits, fuel accumulation and overall material transport at the TEXTOR tokamak are described. Two categories of brittle flaking co-deposits were identified: (i) smooth stratified layers of a thickness of up to 50 mum and a fuel content of up to 16 at.%. (ii) granular and columnar structures reaching 1 mm in thickness and containing around 0.5 at.% of fuel species. They were formed on the blades of the toroidal belt pump limiter (similar to 15000 s of plasma operation) and on the neutraliser plates of this limiter (similar to 90000 s), respectively. A comparison is made to the fuel inventory measured in other controlled fusion devices with carbon walls.

  • 170.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Philipps, Volker
    Forschungszentrum Juelich, Germany.
    Marot, Laurent
    University of Basel, Switzerland.
    Petersson, Per
    Ångström Laboratory, Uppsala University, Association EURATOM – VR, Uppsala, Sweden.
    Pospieszczyk, Albrecht
    Forschungszentrum Juelich, Germany.
    Schweer, Bernd
    Forschungszentrum Juelich, Germany.
    Nitrogen and Neon Retention in Plasma-Facing Materials2011In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 415, no 1, p. S223-S226Article in journal (Refereed)
    Abstract [en]

    Tungsten plate was exposed in the TEXTOR tokamak during nitrogen-assisted discharges. In order to determine material mixing on tungsten, the plate was examined ex situ with ion beam analysis techniques including time-of-flight heavy ion elastic recoil detection analysis and also with X-ray photoelectron spectroscopy. Nitrogen content in the range from 1.3 × 1015 to 3.4 × 1015 cm−2 is measured in the outermost surface layer (20 nm) of the W plate. Photoelectron spectroscopy detects nitrogen both in the elemental and compound form, i.e. tungsten nitride (WN/W2N). Nitrogen is measured even in hot areas free from deuterium. Also neon co-implantation into the plasma-facing components has been identified following Ne-cooled pulses.

  • 171.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Sergienko, G.
    Kreter, A.
    Pospieszczyk, A.
    Psoda, M.
    Wessel, E.
    An overview of fuel retention and morphology in a castellated tungsten limiter2008In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 83, no 7-9, p. 1049-1053Article in journal (Refereed)
    Abstract [en]

    A castellated tungsten test limiter composed of detachable segments was exposed to plasma discharges in the TEXTOR to kamak operated with graphite main limiters. Dismantling of the limiter enabled the analysis Of Surfaces located inside the castellation, The emphasis was on the determination of: (i) deposition and fuel retention; (ii) material mixing and new Compound formation on plasma-facing Surfaces and in the grooves of castellation. The investigation performed by means of accelerator-based ion beam analysis methods, microscopy and X-ray diffraction has brought several essential results: (i) deuterium retention oil plasma-facing Surfaces and in the castellation of metal PFC is strongly related to the co-deposition with carbon; (ii) both carbon and deuterium are detected only in narrow belts, a few millimetre broad, clown the gap with the decay length of around 1.2-1.8 mm; (iii) the presence of copper droplets and tungsten oxide (WO(2)) has been identified in the gaps. Different pathways leading to the oxide formation are considered.

  • 172.
    Rubel, Marek J.
    et al.
    KTH, Superseded Departments, Alfvén Laboratory.
    Tanabe, T.
    Philipps, V.
    Emmoth, Birger
    KTH, Superseded Departments, Microelectronics and Information Technology, IMIT.
    Kirschner, A.
    von Seggern, J.
    Wienhold, P.
    Graphite-tungsten twin limiters in studies of material mixing processes on high heat flux components2000In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 283, p. 1089-1093Article in journal (Refereed)
    Abstract [en]

    Graphite-tungsten twin limiters have been used at the TEXTOR tokamak for testing of high-Z metals as plasma facing materials and, in particular, for studies of the local and global transport of tungsten. The emphasis was on the change in surface morphology of limiters: the formation and properties of multicomponent co-deposits and the trapping characteristics of fuel on carbon and high-Z substrates exposed to the plasma under various operation conditions, i.e., heating scenarios, configuration of limiters, etc. Vast quantities of tungsten have been found to be locally transported to the adjacent graphite surfaces. Ion beam analysis also indicated strong intermixing of carbon, tungsten and boron on the hottest parts of the limiters. The results are discussed in terms of various mechanisms involving the transport of tungsten-containing species, possibilities of oxide production and formation of mixed (W-C-B) compounds.

  • 173.
    Rubel, Marek J.
    et al.
    KTH, Superseded Departments, Alfvén Laboratory.
    Wienhold, P.
    Hildebrandt, D.
    Fuel accumulation in co-deposited layers on plasma facing components2001In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 290, p. 473-477Article in journal (Refereed)
    Abstract [en]

    The work is focused on the determination of the distribution and the total content of deuterium in co-deposits formed in the TEXTOR tokamak on a toroidal belt limiter which is the machine's major plasma facing component (PFC). Limiter tiles in use for 14 100 s of plasma operation were dismounted for examination with surface analysis and microscopy methods. Mapping of the deuterium distribution by means of nuclear reaction analysis (NRA) revealed the presence of deposition zones covering about 35% of the tiles'surface area. Besides C and D, other constituents of the layers were boron, silicon and inconel components. The co-deposit, with a stratified structure and a thickness of up to 50 mum, could be detached from the tiles. Deuterium depth profiling on both sides of the detached flakes and in the underlying graphite substrate enabled the D content in the deposition zones to be estimated at a level of 3.5 x 10(19) cm(-2). Adding the fuel content found in the erosion zone (3-7 x 10(17) cm(-2)) and on the back side of the tile (0.9-1.8 x 10(17) cm(-2)), the total amount of D atoms trapped in all the limiter tiles was assessed to be about 2 x 10(23) atoms. D content in the co-deposits accounted for approximately 10 at.% (C-D/C-C similar to 0.1) which was considered to be low in comparison to much greater values observed in other devices. The results of the ion beam analyses (IBA) agree well with the determination by thermal desorption spectrometry (TDS).

  • 174.
    Rubel, Marek J.
    et al.
    KTH, Superseded Departments, Alfvén Laboratory.
    Wienhold, P.
    Hildebrandt, D.
    Ion beam analysis methods in the studies of plasma facing materials in controlled fusion devices2003In: Vacuum, ISSN 0042-207X, E-ISSN 1879-2715, Vol. 70, no 03-feb, p. 423-428Article in journal (Refereed)
    Abstract [en]

    Application of ion beam analysis techniques in the studies of material transport and fuel inventory in the controlled fusion devices is exemplified. Enhanced proton scattering on the carbon isotopes C-12(p,p)C-12, C-13(p,p)C-13 and secondary ion mass spectrometry allowed for determination of carbon erosion and re-deposition on the wall components following the experiments with a tracer ((CH4)-C-13) injection into the plasma edge at the TEXTOR tokamak. For the assessment of the deuterium fuel accumulation in the plasma facing components depth profiling by means of nuclear reaction analysis, He-3(d,p)He-4, was performed. Advantages and limitations of those nuclear methods in solving experimental problems are addressed.

  • 175.
    Rubel, Marek
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Moon, Soonwoo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Garcia Carrasco, Alvaro
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Hallén, Anders
    KTH, School of Information and Communication Technology (ICT), Electronics.
    Krawczynska, A.
    Fortuna-Zalesna, E.
    Gilbert, M.
    Plocinski, T.
    Widdowson, A.
    Metallic mirrors for plasma diagnosis in current and future reactors: tests for ITER and DEMO2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014061Article in journal (Refereed)
    Abstract [en]

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

  • 176.
    Rubel, Marek
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Alves, Eduardo
    Brezinsek, Sebastijan
    Coad, Joseph Paul
    Heinola, Kalle
    Mayer, Matej
    Widdowson, Anna
    The role and application of ion beam analysis for studies of plasma-facing components in controlled fusion devices2016In: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 371, p. 4-11Article in journal (Refereed)
    Abstract [en]

    First wall materials in controlled fusion devices undergo serious modification by several physical and chemical processes arising from plasma-wall interactions. Detailed information is required for the assessment of material lifetime and accumulation of hydrogen isotopes in wall materials. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks, especially in JET with an ITER-Like Wall. IBA techniques play a particularly prominent role here because of their isotope selectivity in the low-Z range (1-10), high sensitivity and combination of several methods in a single run. The role of He-3-based NRA, RBS (standard and micro size beam) and HIERDA in fuel retention and material migration studies is presented. The use of tracer techniques with rare isotopes (e.g. N-15) or marker layers on wall diagnostic components is described. Special instrumentation, development of equipment to enhance research capabilities and issues in handling of contaminated materials are addressed. (C) 2015 Published by Elsevier B.V.

  • 177.
    Rubel, Marek
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Zhou, Yushan
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Coad, J. P.
    Lungu, C.
    Jepu, I.
    Porosnicu, C.
    Matveev, D.
    Kirschner, A.
    Brezinsek, S.
    Widdowson, A.
    Alves, E.
    Fuel inventory and deposition in castellated structures in JET-ILW2017In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 6, article id 066027Article in journal (Refereed)
    Abstract [en]

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

  • 178.
    Rubel, Marek
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Philipps, V.
    IEK-4, Forschungszentrum Jülich.
    Huber, A.
    IEK-4, Forschungszentrum Jülich.
    Ivanova, Darya
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Schweer, B.
    IEK-4, Forschungszentrum Jülich.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Zlobinski, M.
    IEK-4, Forschungszentrum Jülich.
    Overview of Fuel Removal Methods from Plasma-Facing Components2011In: 38th EPS Conference on Plasma Physics, 2011Conference paper (Refereed)
  • 179.
    Rubel, Marek
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Sergienko, G.
    Kreter, A.
    Pospieszczyk, A.
    Psoda, M.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Wessel, E.
    Fuel deposition and material mixing in a castellated tungsten limiter2007In: 34th EPS Conference on Plasma Physics 2007, EPS 2007 - Europhysics Conference Abstracts, 2007, no 1, p. 303-306Conference paper (Refereed)
  • 180.
    Rubel, Marek
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Weckmann, Armin
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Ström, Petter
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Garcia Carrasco, Alvaro
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brezinsek, S.
    Coenen, J.
    Kreter, A.
    Moeller, S.
    Wienhold, P.
    Wauters, T.
    Fortuna-Zalesna, E.
    Tracer techniques for the assessment of material migration and surface modification of plasma-facing components2015In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, p. 280-284Article in journal (Refereed)
    Abstract [en]

    Tracer techniques were used in the TEXTOR tokamak to determine high-Z metal migration and the deposition of species used for plasma edge cooling or wall conditioning under different types of operation conditions. Volatile molybdenum hexa-fluoride, nitrogen-15 and oxygen-18 were used as markers in tokamak or ion cyclotron wall conditioning discharges (ICWC). The objective was to obtain qualitative and quantitative of a global and local deposition pattern and material mixing effects. The deposition and retention was studied on plasma-facing components, collector probes and test limiters. Optical spectroscopy and ex-situ analysis techniques were used to determine the plasma response to tracer injection and the modification of surface composition. Molybdenum and light isotopes were detected on all types of limiters and short-term probes retrieved from the vessel showing that both helium and nitrogen are trapped following wall conditioning and edge cooling. Only small amounts below 1 x 10(19) m(-2) of O-18 were detected on surfaces treated by oxygen-assisted ICWC.

  • 181. Ruset, C.
    et al.
    Grigore, E.
    Luculescu, C.
    Tiseanu, I.
    Likonen, J.
    Mayer, M.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. EUROfusion Consortium, Culham Science Centre, JET, Abingdon, United Kingdom.
    Matthews, G. F.
    Investigation on the erosion/deposition processes in the ITER-like wall divertor at JET using glow discharge optical emission spectrometry technique2016In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T167, article id 014049Article in journal (Refereed)
    Abstract [en]

    As a complementary method to Rutherford back scattering (RBS), glow discharge optical emission spectrometry (GDOES) was used to investigate the depth profiles of W, Mo, Be, O and C concentrations into marker coatings (CFC/Mo/W/Mo/W) and the substrate of divertor tiles up to a depth of about 100 μm. A number of 10 samples cored from particular areas of the divertor tiles were analyzed. The results presented in this paper are valid only for those areas and they cannot be extrapolated to the entire tile. Significant deposition of Be was measured on Tile 3 (near to the top), Tile 6 (at about 40 mm from the innermost edge) and especially on Tile 0 (HFGC). Preliminary experiments seem to indicate a penetration of Be through the pores and imperfections of CFC material up to a depth of 100 μm in some cases. No erosion and a thin layer of Be (<1 μm) was detected on Tiles 4, 7 and 8. On Tile 1 no erosion was found at about 1/3 from bottom.

  • 182. Sergienko, G.
    et al.
    Bazylev, B.
    Hirai, T.
    Huber, A.
    Kreter, A.
    Mertens, Ph
    Nedospasov, A.
    Philipps, V.
    Pospieszczyk, A.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Samm, U.
    Schweer, B.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Tokar, M.
    Wessel, E.
    Textor team,
    Experience with bulk tungsten test-limiters under high heat loads: melting and melt layer propagation2007In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T128, p. 81-86Article in journal (Refereed)
    Abstract [en]

    The paper provides an overview of processes and underlying physics governing tungsten melt erosion in the fusion plasma environment. Experiments with three different bulk tungsten test-limiters were performed in TEXTOR: (i) thermally insulated solid plate fixed on a graphite roof-like limiter heated up by the plasma to the melting point, (ii) macro-brush of the ITER-relevant castellated structure and (iii) lamellae structure developed for the JET divertor. The main objectives were to determine the metal surface damage, the formation of the melt layer and its motion in the magnetic field. PHEMOBRID-3D and MEMOS-1.5D numerical codes were used to simulate the experiment with the roof-like test-limiter. Both experiments and simulation showed that the melting of tungsten can lead to a large material redistribution due to thermo-electron emission currents without ejection of molten material to the plasma.

  • 183. Sergienko, G.
    et al.
    Bazylev, B.
    Huber, A.
    Kreter, A.
    Litnovsky, A.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Philipps, V.
    Pospieswzyk, A.
    Mertens, Ph
    Samm, U.
    Schweer, B.
    Schmitz, O.
    Tokar, M.
    Erosion of a tungsten limiter under high heat flux in TEXTOR2007In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 363, p. 96-100Article in journal (Refereed)
    Abstract [en]

    Erosion characteristics of a tungsten plate heated up in TEXTOR by the plasma load have been investigated at temperatures extending to the melting point. No enhancement of atomic release exceeding physical sputtering and normal thermal sublimation for temperatures below 3700 K was observed. The liquid tungsten moved fast along the plate in the direction perpendicular to the magnetic field lines. The motion is caused by the Lorentz force due to the thermoelectron current emitted from the hot tungsten surface. The motion of liquid tungsten caused a material loss of 2.85 g during two discharges. The material redistribution due to the melt layer motion is compared with a MEMOS-1.5D simulation.

  • 184. Sergienko, G.
    et al.
    Huber, A.
    Kreter, A.
    Philipps, V.
    Pospieszczyk, A.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Schweer, B.
    Schmitz, O.
    High temperature erosion of tungsten exposed to the TEXTOR edge plasma2005In: 32nd EPS Conference on Plasma Physics 2005, EPS 2005, Held with the 8th International Workshop on Fast Ignition of Fusion Targets: Europhysics Conference Abstracts, 2005, p. 373-376Conference paper (Refereed)
  • 185. Sergienko, G.
    et al.
    Huber, A.
    Kreter, A.
    Philipps, V.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Schweer, B.
    Schmitz, O.
    Tokar, M.
    Tungsten melting under high power load in the TEXTOR edge plasma2005In: 32nd EPS Conference on Plasma Physics 2005, EPS 2005, Held with the 8th International Workshop on Fast Ignition of Fusion Targets: Europhysics Conference Abstracts, 2005, p. 377-380Conference paper (Refereed)
  • 186. Strachan, J. D.
    et al.
    Likonen, J.
    Coad, P.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Widdowson, A.
    Airila, M.
    Andrew, P.
    Brezinsek, S.
    Corrigan, G.
    Esser, H. G.
    Jachmich, S.
    Kallenbach, A.
    Kirschner, A.
    Kreter, A.
    Matthews, G. F.
    Philipps, V.
    Pitts, R. A.
    Spence, J.
    Stamp, M.
    Wiesen, S.
    Modelling of carbon migration during JET C-13 injection experiments2008In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 48, no 10Article in journal (Refereed)
    Abstract [en]

    JET has performed two dedicated carbon migration experiments on the final run day of separate campaigns ( 2001 and 2004) using (CH4)-C-13 methane injected into repeated discharges. The EDGE2D/NIMBUS code modelled the carbon migration in both experiments. This paper describes this modelling and identifies a number of important migration pathways: ( 1) deposition and erosion near the injection location, ( 2) migration through the main chamber SOL, (3) migration through the private flux region (PFR) aided by E x B drifts and ( 4) neutral migration originating near the strike points. In H-Mode, type I ELMs are calculated to influence the migration by enhancing erosion during the ELM peak and increasing the long-range migration immediately following the ELM. The erosion/re-deposition cycle along the outer target leads to a multistep migration of C-13 towards the separatrix which is called 'walking'. This walking created carbon neutrals at the outer strike point and led to 13C deposition in the PFR. Although several migration pathways have been identified, quantitative analyses are hindered by experimental uncertainty in divertor leakage, and the lack of measurements at locations such as gaps and shadowed regions.

  • 187.
    Ström, Petter
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Possnert, Goran
    Uppsala Univ, Dept Phys & Astron, Tandem Lab, Box 529, SE-75120 Uppsala, Sweden..
    Erratum: "A combined segmented anode gas ionization chamber and time-of-flight detector for heavy ion elastic recoil detection analysis" [Rev. Sci. Instrum. 87, 103303 (2016)]2018In: Review of Scientific Instruments, ISSN 0034-6748, E-ISSN 1089-7623, Vol. 89, no 4, article id 049901Article in journal (Refereed)
  • 188.
    Ström, Petter
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Possnert, Göran
    A combined segmented anode gas ionization chamber and time-of-flight detector for heavy ion elastic recoil detection analysis2016In: Review of Scientific Instruments, ISSN 0034-6748, E-ISSN 1089-7623, Vol. 87, no 10, article id 103303Article in journal (Refereed)
    Abstract [en]

    A dedicated detector system for heavy ion elastic recoil detection analysis at the Tandem Laboratory of Uppsala University is presented. Benefits of combining a time-of-flight measurement with a segmented anode gas ionization chamber are demonstrated. The capability of ion species identification is improved with the present system, compared to that obtained when using a single solid state silicon detector for the full ion energy signal. The system enables separation of light elements, up to Neon, based on atomic number while signals from heavy elements such as molybdenum and tungsten are separated based on mass, to a sample depth on the order of 1 μm. The performance of the system is discussed and a selection of material analysis applications is given. Plasma-facing materials from fusion experiments, in particular metal mirrors, are used as a main example for the discussion. Marker experiments using nitrogen-15 or oxygen-18 are specific cases for which the described improved species separation and sensitivity are required. Resilience to radiation damage and significantly improved energy resolution for heavy elements at low energies are additional benefits of the gas ionization chamber over a solid state detector based system.

  • 189.
    Ström, Petter
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Primetzhofer, D.
    Brezinsek, S.
    Kreter, A.
    Unterberg, B.
    Sergienko, G.
    Sugiyama, K.
    Ion beam analysis of tungsten layers in EUROFER model systems and carbon plasma facing components2016In: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 371, p. 355-359Article in journal (Refereed)
    Abstract [en]

    The tungsten enriched surface layers in two fusion-relevant EUROFER steel model samples, consisting of an iron-tungsten mixture exposed to sputtering by deuterium ions, were studied by Rutherford backscattering spectrometry and medium energy ion scattering. Exposure conditions were the same for the two samples and the total amount of tungsten atoms per unit area in the enriched layers were similar (2e15 and 2.4e15 atoms/cm2 respectively), despite slightly different initial atomic compositions. A depth profile featuring exponential decrease in tungsten content towards higher depths with 10-20 atomic percent of tungsten at the surface and a decay constant between 0.05 and 0.08 Å-1 was indicated in one sample, whereas only the total areal density of tungsten atoms was measured in the other. In addition, two different beams, iodine and chlorine, were employed for elastic recoil detection analysis of the deposited layer on a polished graphite plate from a test limiter in the TEXTOR tokamak following experiments with tungsten hexafluoride injection. The chlorine beam was preferred for tungsten analysis, mainly because it (as opposed to the iodine beam) does not give rise to problems with overlap of forward scattered beam particles and recoiled tungsten in the spectrum.

  • 190.
    Ström, Petter
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Weckmann, Armin
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brezinsek, S.
    Kreter, A.
    Möller, S.
    Rozniatowski, K.
    Characterisation of surface layers formed on plasma-facing components in controlled fusion devices: Role of heavy ion elastic recoil detection2015In: Vacuum, ISSN 0042-207X, E-ISSN 1879-2715, Vol. 122, p. 260-267Article in journal (Refereed)
    Abstract [en]

    Wall components retrieved from the TEXTOR tokamak after tracer experiments with nitrogen-15 and molybdenum hexafluoride (MoF6) injection were studied to determine deposition patterns and, by this, to conclude on material migration. Toroidal limiter tiles made of carbon fibre composites and fine grain graphite were examined using time-of-flight heavy ion elastic recoil detection analysis. Molybdenum deposition patterns indicated migration based on erosion and prompt re-deposition. Nitrogen-15 was trapped together with the deposited molybdenum. Some information on the depth distribution of species in the top 400 nm layer of the limiters was obtained; however surface roughness of the samples strongly limited resolution. In the case of molybdenum, the largest concentration was found in the 100 nm outermost layer, whereas fluorine and nitrogen-15 displayed more irregular profiles. Other species, besides deuterium fuel and carbon-12, were also identified: boron-10 and boron-11 originating from boronisations, carbon-13 from earlier tracer experiments, nitrogen-14 from plasma edge cooling and metals eroded from the Inconel wall.

  • 191. Sugiyama, K.
    et al.
    Tanabe, T.
    Miyasaka, K.
    Masaki, K.
    Tobita, K.
    Miya, N.
    Philipps, V.
    Rubel, Marek J.
    KTH, Superseded Departments, Alfvén Laboratory.
    Skinner, C. H.
    Gentile, C. A.
    Saze, T.
    Nishizawa, K.
    Tritium profile in plasma-facing components following D-D operation2004In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 329-33, p. 874-879Article in journal (Refereed)
    Abstract [en]

    We have investigated the tritium depth profile near the surface of the limiter/divertor tiles used in the deuterium fueled machines, such as TEXTOR, TFTR and JT-60U by means of the imaging plate technique and a tritium survey monitor. Tritium depth profiles near the surface of the sample tiles were estimated by comparing the experimental results to a calculation using a 3-D Monte-Carlo code. In every sample tile, there was little tritium in the range from the surface to 1 mum depth. In contrast, tritium density tended to increase beyond 1 mum depth. These results indicate that the tritium retained near the surface was easily removed by isotope exchange with a deuterium plasma or various other tritium removal operations. On the other hand, such operations did not remove tritium retained beyond 1 mum depth, and this could be a potential issue in a next D-T machine.

  • 192.
    Sundelin, Per
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Emmoth, B.
    Philipps, V.
    Sergienko, G.
    A test of nitrogen-assisted plasma discharges for fuel removal from plasma-facing components in tokamaks2008In: PROCEEDINGS OF THE 17TH INTERNATIONAL VACUUM CONGRESS/13TH INTERNATIONAL CONFERENCE ON SURFACE SCIENCE/INTERNATIONAL CONFERENCE ON NANOSCIENCE AND TECHNOLOGY / [ed] Johansson LSO, Andersen JN, Gothelid M, Helmersson U, Montelius L, Rubel M, Setina J, Wernersson LE, Bristol: IOP PUBLISHING LTD , 2008, Vol. 100, p. 062027-Conference paper (Refereed)
    Abstract [en]

    Safety regulations limit the amount of tritium accumulated in wall components of a fusion reactor to 350g. Because of this, reduction of long-term fuel inventory is one of the most urgent tasks to be resolved to ensure the safe and economic operation of a reactor-class fusion device. Several methods have been suggested and tested. The aim of this paper is to evaluate the cleaning efficiency of plasma-facing components by ICRH-assisted plasma discharges with in nitrogen-hydrogen in the TEXTOR tokamak. Three types of probes were investigated: laboratory prepared a-C: D layers on silicon; boron layers on silicon obtained by pre-boronisation in TEXTOR and not coated Inconel substrates. The main results are following: (i) laboratory prepared a-C: D layers are not affected: deuterium and carbon contents did not decrease (ii) the morphology of layers pre-boronised in TEXTOR is not affected (iii) no significant effects were noticed on Inconel probes. A comparison of cleaning methods with nitrogen and oxygen is also presented.

  • 193.
    Sundelin, Per
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Schulz, C.
    Philipps, V.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Sergienko, G.
    Marot, L.
    Nitrogen-assisted removal of deuterated carbon layers2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 390-91, p. 647-650Article in journal (Refereed)
    Abstract [en]

    Deuterated carbon films prepared in laboratory and boronised films prepared in the TEXTOR tokamak were exposed to hydrogen-nitrogen plasmas in order to determine erosion characteristics and fuel removal efficiency. Exposures were performed in: (i) TEXTOR tokamak during ion cyclotron heated wall conditioning discharges (ICWC) and (ii) TOMAS magnetic plasma facility in radio frequency-assisted glow discharges. The essential results are: (i) films exposed in TEXTOR are not affected: deuterium and carbon content does not decrease and the morphology is unchanged, and (ii) deuterium and carbon contents in films exposed in TOMAS is reduced by 30-60% after 2 h of cleaning and topographical changes are noted. The study shows that while exposure to H-2-N-2 laboratory plasma removes a-C:D films, no effect is seen at the position of the sample exposure during tokamak ICWC plasmas. It also indicates that the removal efficiency is only weakly related to nitrogen, since the highest removal efficiency is seen with pure hydrogen plasma. A comparison to oxygen-assisted fuel removal is given.

  • 194. Tanabe, T.
    et al.
    Miyasaka, K.
    Rubel, Marek J.
    KTH, Superseded Departments, Alfvén Laboratory.
    Philipps, V.
    Tritium and deuterium retention in graphite limiters in TEXTOR2002In: Fusion science and technology, ISSN 1536-1055, E-ISSN 1943-7641, Vol. 41, no 3, p. 924-928Article in journal (Refereed)
    Abstract [en]

    In order to investigate tritium behavior in tokamak, we have measured surface distributions of deuterium and tritium on graphite limiter tiles used in TEXTOR under D-D operation by means of an ion beam analysis and tritium imaging plate technique, respectively. It was found that both distributions were quite different, i.e. deuterium retention was higher at the deposited area, whereas tritium retention was higher at the erosion dominated area. This is because tritium produced by the D-D reaction, initially having 1 MeV, did not fully lose its energy in the TEXTOR plasma and implanted into the plasma facing materials nearly homogeneously, whereas deuterium was codeposited with carbon and boron, the main impurities in the TEXTOR plasma. This is also confirmed by the finding that high level of tritium was detected beneath the deposited layer. Tritium distribution, however, was modified by the temperature increase due to plasma heat load. Thus the comparison of tritium profiles with the deuterium profile gives a large amount of important and new information on PMI, and may be used as a new diagnostic technique for PMI.

  • 195. Tanabe, T.
    et al.
    Ohgo, T.
    Wada, M.
    Rubel, Marek J.
    KTH, Superseded Departments, Alfvén Laboratory.
    Philipps, V.
    von Seggern, J.
    Ohya, K.
    Huber, A.
    Pospieszczyk, A.
    Schweer, B.
    Textor Team,
    Material mixing on W/C twin limiter in TEXTOR-942000In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 49, p. 355-362Article in journal (Refereed)
    Abstract [en]

    In order to investigate the effect of mutual contamination between tungsten (W) and carbon

  • 196. Tanabe, T.
    et al.
    Wada, M.
    Ohgo, T.
    Philipps, V.
    Rubel, Marek J.
    KTH, Superseded Departments, Alfvén Laboratory.
    Huber, A.
    von Seggern, J.
    Ohya, K.
    Pospieszczyk, A.
    Schweer, B.
    Textor Team,
    Application of tungsten for plasma limiters in TEXTOR2000In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 283, p. 1128-1133Article in journal (Refereed)
    Abstract [en]

    Three different types of W limiters were exposed in the TEXTOR plasma and the response of the plasma and materials performance of the limiters were investigated. 1. A W bulk limiter operated with preheating above 800 K withstood a plasma heat load of about similar to 20 MW/m(2) for a few seconds with some slight surface melting during the highest heat load shot. However, it was severely damaged when operated at around 500 K. 2. A C/W twin test limiter, half made of bulk W and the other half of graphite (EK-98) gave very useful information on how low- and high-Z materials behave under conditions of simultaneous utilization as PFM such as cross-contamination and the influence of a large mass difference on hydrogen reflection and deposition. 3. Two sets of main poloidal W limiters made of vacuum vapor sprayed (VPS)-W deposited on graphite (IG-430U) with a Re interlayer could absorb about 60% of the total convection heat and the ohmic plasma with a density as high as 5 x 10(13) cm(-3) was sustained. Most of the VPS-W coated limiters tolerated a heal load of similar to 20 MW/m(2). This series of W limiters experiments in TEXTOR has shown that W is applicable as a PFM, if its central accumulation is avoided by NBI and/or ICRH heating. Nevertheless, some concerns still remain, including difficulty of plasma startup, W behavior in higher temperature plasmas, and materials' selection.

  • 197. Thomser, C.
    et al.
    Bailescu, V.
    Brezinsek, S.
    Coenen, J. W.
    Greuner, H.
    Hirai, T.
    Linke, J.
    Lungu, C. P.
    Maier, H.
    Matthews, G.
    Mertens, Ph
    Neu, R.
    Philipps, V.
    Riccardo, V.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Ruset, C.
    Schmidt, A.
    Uytdenhouwen, I.
    Plasma facing materials for the jet iter-like wall2012In: Fusion science and technology, ISSN 1536-1055, E-ISSN 1943-7641, Vol. 62, no 1, p. 1-8Article in journal (Refereed)
    Abstract [en]

    The chosen materials for plasma facing components for the deuterium/tritium phase of ITER are beryllium and tungsten. These materials have already been widely investigated in various devices like ion beam or electron beam tests. However, the operation of this material combination in a large tokamak including plasma wall interaction, material degradation, erosion and material mixing has not been proven yet. The ITER-like Wall, which has been recently installed in JET, consists of a combination of bulk tungsten and tungsten coated CFC divertor tiles as well as bulk beryllium and beryllium coated INCONEL in the main chamber. The experiments in JET will provide the first fully representative test of the ITER material choice under relevant conditions. This paper concentrates on material research and developments for the materials of the JET ITER-like Wall with respect to mechanical and thermal properties. The impact of these materials and components on the JET operating limits with the ITER-like Wall and implications for the ongoing scientific program will be summarised.

  • 198. Tokitani, M.
    et al.
    Miyamoto, M.
    Masuzaki, S.
    Fujii, Y.
    Sakamoto, R.
    Oya, Y.
    Hatano, Y.
    Otsuka, T.
    Oyaidzu, M.
    Kurotaki, H.
    Suzuki, T.
    Hamaguchi, D.
    Isobe, K.
    Asakura, N.
    Widdowson, A.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Micro-/nano-characterization of the surface structures on the divertor tiles from JET ITER-like wall2017In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 116, p. 1-4Article in journal (Refereed)
    Abstract [en]

    Micro-/nano-characterization of the surface structures on the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with the ITER-like wall (JET ILW) were studied. The analyzed tiles were a single poloidal section of the tile numbers of 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. A sample from the apron of Tile 1 was deposition-dominated. Stratified mixed-material layers composed of Be, W, Ni, O and C were deposited on the original W-coating. Their total thickness was ∼1.5 μm. By means of transmission electron microscopy, nano-size bubble-like structures with a size of more than 100 nm were identified in that layer. They could be related to deuterium retention in the layer dominated by Be. The surface microstructure of the sample from Tile 4 also showed deposition: a stratified mixed-material layer with the total thickness of 200–300 nm. The electron diffraction pattern obtained with transmission electron microscope indicated Be was included in the layer. No bubble-like structures have been identified. The surface of Tile 3, originally coated by Mo, was identified as the erosion zone. This is consistent with the fact that the strike point was often located on that tile during the plasma operation. The study revealed the micro- and nano-scale modification of the inner tile surface of the JET ILW. In particular, a complex mixed-material deposition layer could affect hydrogen isotope retention and dust formation.

  • 199. Tsavalas, P.
    et al.
    Lagoyannis, A.
    Mergia, K.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Triantou, K.
    Harissopulos, S.
    Kokkoris, M.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Be ITER-like wall at the JET tokamak under plasma2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014049Article in journal (Refereed)
    Abstract [en]

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

  • 200. Tsitrone, E.
    et al.
    Pegourie, B.
    Marandet, Y.
    Artaud, J. F.
    Brosset, C.
    Bucalossi, J.
    Corre, Y.
    Dittmar, T.
    Gauthier, E.
    Languille, P.
    Linez, F.
    Loarer, T.
    Martin, C.
    Roubin, P.
    Kallenbach, A.
    Krieger, K.
    Mayer, M.
    Neu, R.
    Rohde, V.
    Roth, J.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Brezinsek, S.
    Kirschner, A.
    Kreter, A.
    Litnovsky, A.
    Philipps, V.
    Wienhold, P.
    Likonen, J.
    Coad, P.
    Lipschultz, B.
    Doerner, R.
    Multi machine scaling of fuel retention in 4 carbon dominated tokamaks2011In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 415, no 1, p. S735-S739Article in journal (Refereed)
    Abstract [en]

    In order to benchmark predictions for the in vessel tritium inventory in ITER, a survey of fuel retention measured in 4 carbon dominated tokamaks (TEXTOR, ASDEX Upgrade in the 2002-2003 carbon configuration, Tore Supra and JET) was performed, showing retention rates from similar to 1 g D/h in TEXTOR (L mode, limiter machine) up to similar to 6-12 g D/h in AUG (H mode, divertor machine). A simple scaling used for ITER predictions is applied for comparison with experimental values: (1) estimate of wall fluxes, (2) estimate of the gross carbon erosion, (3) estimate of the net erosion/redeposition assuming a redeposition fraction and (4) estimate of the retention rate using D/C ratio scalings. The validity of each step is discussed, showing that this approach yields the right order of magnitude, but tends to underestimate the experimental values unless a high wall flux, a low local redeposition fraction and/or a high D/C ratio are used.

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