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  • 1.
    Emmoth, Birger
    et al.
    KTH, School of Information and Communication Technology (ICT), Microelectronics and Information Technology, IMIT.
    Kreter, A.
    Hallén, Anders
    KTH, School of Information and Communication Technology (ICT), Microelectronics and Information Technology, IMIT.
    Jakubowski, M.
    Lehnen, M.
    Litnovsky, A.
    Petersson, P.
    Philipps, V.
    Possnert, G.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Schweer, B.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Unterberg, B.
    Wienhold, P.
    In-situ measurements of carbon and deuterium deposition using the fast reciprocating probe in TEXTOR2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 390-91, p. 179-182Article in journal (Refereed)
    Abstract [en]

    Silicon samples were exposed in the scrape-off layer of the TEXTOR plasma using a fast reciprocating probe, with the aim of studying carbon deposition and deuterium retention during Dynamic Ergodic Divertor (DED) operation. Separate samples were exposed for 300 ms at the flat-top phase of neutral beam heated discharges. The exposure conditions were varied on a shot-to-shot basis by external magnetic perturbations generated by the DED in the m/n = 3/1, DC regime, base configuration. Nuclear Reaction Analysis (NRA) was used to characterise collector sample surfaces after their exposure. Enhanced concentrations of both carbon and deuterium (C 3-10 x 10(16) at./cm(2), D 8-60 x 10(15) at./cm(2)) were found. The D/C ratio was less than unity which indicates that most of the carbon and deuterium were co-deposited. Carbon e-folding lengths of about 2 cm were found on both toroidal sides of the probe independent of DED perturbations.

  • 2. Gasior, P.
    et al.
    Irrek, F.
    Petersson, P.
    Penkalla, Hj.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Schweer, B.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Wessel, E.
    Linke, J.
    Philipps, V.
    Emmoth, Birger
    KTH, School of Information and Communication Technology (ICT), Microelectronics and Information Technology, IMIT.
    Wolowski, J.
    Hirai, T.
    Laser-induced removal of co-deposits from graphitic plasma-facing components: Characterization of irradiated surfaces and dust particles2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 390-91, p. 585-588Article in journal (Refereed)
    Abstract [en]

    Laser-induced fuel desorption and ablation of co-deposited layers on limiter plates from the TEXTOR tokamak have been studied. Gas phase composition was monitored in situ, whereas the ex situ studies have been focused on the examination of irradiated surfaces and broad analysis of dust generated by ablation of co-deposits. The size of the dust grains is in the range of few nanometers to hundreds of micrometers. These are fuel-rich dust particles, as determined by nuclear reaction analysis. The presence of deuterium in dust indicates that not all fuel species are transferred to the gas phase during irradiation. This also suggests that photonic removal of fuel and the ablation of co-deposit from plasma-facing components may lead to the redistribution of fuel-containing dust to surrounding areas.

  • 3. Hirai, T.
    et al.
    Linke, J.
    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.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Kuehnlein, W.
    Wessel, E.
    Coad, J. P.
    Lungu, C. P.
    Matthews, G. F.
    Pedrick, L.
    Piazza, G.
    Characterization and heat flux testing of beryllium coatings on Inconel for JET ITER-like wall project2007In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T128, p. 166-170Article in journal (Refereed)
    Abstract [en]

    In order to perform a fully integrated material test, JET has launched the ITER-like wall project with the aim of installing a full metal wall during the next major shutdown. The material foreseen for the main chamber wall is bulk Be at the limiters and Be coatings on inconel tiles elsewhere. R&D process comprises global characterization ( structure, purity etc) of the evaporated films and testing of their performance under heat loads. The major results are (i) the layers have survived energy loads of 20 MJ m(-2) which is significantly above the required level of 5 - 10 MJ m(-2), (ii) melting limit of beryllium coating would be at the energy level of 30 MJ m(-2), (iii) cyclic thermal load of 10 MJ m(-2) for up to 50 cycles have not induced any noticeable damage such as flaking or detachment.

  • 4.
    Ivanova, Darya
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), 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.
    Huang, Z.
    KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    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.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Wessel, E.
    Institute for Energy Research, Forschungszentrum Jülich, Association EURATOM-FZJ, Germany.
    Survey of dust formed in the TEXTOR tokamak: structure and fuel retention2009In: Physica scripta. T, ISSN 0281-1847, Vol. T138, p. 014025-Article in journal (Refereed)
    Abstract [en]

    A detailed survey of erosion and deposition on plasma-facing components was performed in the TEXTOR tokamak. Co-deposits and dust particles were collected from graphite limiters and from several locations on the Inconel liner. The total amount of dust (loose material), originating mainly from carbon-rich co-deposits detached from the limiters and the liner, was around 2 g, with sizes from 0.1 mu m to 1 mm. The morphology and fuel retention was determined using microscopy methods, ion beam analysis and thermal desorption spectrometry. The study revealed differences in structure and fuel content between deposits from the toroidal and main poloidal limiters. There were also splashes, up to 1 mm in diameter, of molten metal (mainly nickel) on the toroidal limiters. Issues of the dust conversion factor (erosion-to-dust) are addressed and a comparison with results of previous dust surveys at TEXTOR is also briefly presented.

  • 5. Kreter, A.
    et al.
    Brezinsek, S.
    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.
    Emmoth, Birger
    KTH, School of Information and Communication Technology (ICT), Microelectronics and Information Technology, IMIT.
    Freisinger, M.
    Pelicon, P.
    Philipps, V.
    Schmitz, O.
    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.
    Sergienko, G.
    Deuterium retention in different carbon materials exposed in TEXTOR2007In: 34th EPS Conference on Plasma Physics 2007, EPS 2007 - Europhysics Conference Abstracts, 2007, no 1, p. 315-318Conference paper (Refereed)
  • 6. Kreter, A.
    et al.
    Brezinsek, S.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Emmoth, Birger
    KTH, School of Information and Communication Technology (ICT), Microelectronics and Applied Physics, MAP.
    Freisinger, M.
    Pelicon, P.
    Philipps, V.
    Schmitz, O.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Sergienko, G.
    Deuterium retention in different carbon materials exposed in TEXTOR:  2008In: PROCEEDINGS OF THE 17TH INTERNATIONAL VACUUM CONGRESS/13TH INTERNATIONAL CONFERENCE ON SURFACE SCIENCE/INTERNATIONAL CONFERENCE ON NANOSCIENCE AND TECHNOLOGY, Bristol: IOP PUBLISHING LTD , 2008, Vol. 100, no PART 6, p. 062024-Conference paper (Refereed)
    Abstract [en]

    CFC NB31, CFC DMS780 and fine-grain graphite EK98 were simultaneously exposed in the SOL of TEXTOR in order to measure the retention of deuterium in the material bulk. After exposure, the samples were analysed by thermal desorption spectrometry and nuclear reaction analysis with a conventional beam as well as with a microbeam. The deuterium retention amounts (2-4).10(21) D/m(2) for the incident fluence of similar to 2.10(25) D/m(2). The retention values are similar for both CFC materials and lower by similar to 20-40% for EK98. The retention in all three materials scales roughly with a square root of incident fluence without saturation for the range of fluences obtained. The majority of deuterium is stored in a surface layer of <8 mu m. However, in NB31 deuterium is detectable as deep as 80 mu m. The in-bulk retention estimated for a TEXTOR experimental campaign of approximate to 7500 s of plasma has a contribution of approximate to 10% to the total retention, which is dominated by deuterium-carbon co-deposition.

  • 7. Krieger, K.
    et al.
    Brezinsek, S.
    Jachmich, S.
    Lisgo, S.
    Stamp, M.
    Esser, H. G.
    Kreter, A.
    Menmuir, Sheena
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Mertens, Ph.
    Philipps, V.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Be wall sources and migration in L-mode discharges after Be evaporation in the JET tokamak2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 390-91, p. 110-114Article in journal (Refereed)
    Abstract [en]

    First wall material erosion and migration after fresh Be evaporation in the JET tokamak were studied in a series of consecutive identical L-mode discharges. The evolution of Be and C wall and divertor sources towards steady state conditions after deposition of a finite amount of Be at the carbon first wall of JET by the beryllium evaporation procedure provides a data set for benchmarking impurity transport simulations. Furthermore the experiment serves as a reference case for comparison of Be erosion to that of the planned ITER-like wall experiment with mainly Be plasma facing components (PFCs) in the main chamber. The experimental results confirm the migration pattern obtained by campaign integrated accounting of impurity sources and sinks, which is characterised by the main chamber wall and outer divertor as erosion dominated zones and migration of eroded material predominantly to the inner divertor where the material is finally deposited at the target plates.

  • 8. Psoda, M.
    et al.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Sergienko, G.
    Sundelin, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Pospieszczyk, A.
    Material mixing on plasma-facing components: Compound formation2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 386-388, p. 740-743Article in journal (Refereed)
    Abstract [en]

    Two different tungsten limiters (castellated bulk metal block and W-coated graphite), subjected to high power loads in the TEXTOR tokamak, were examined in order to determine chemical composition of deposits inside the castellated grooves and on side surfaces of the coated limiter. Comprehensive analyses carried out by X-ray diffraction, ion beam analysis and other methods revealed: (i) the formation of tungsten oxide (WO2) inside the castellated grooves: (ii) the formation of tungsten carbides (WC main phase and traces of W2C) on side surfaces of the coated limiter. Elemental tungsten was found in deposits on side surfaces only in trace quantities thus indicating that tungsten eroded from the limiter top and transported to the scrape-off layer reacted with carbon. Based on thermodynamic data, the pathways leading to the formation of compounds are discussed.

  • 9.
    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.
    Coad, J. P.
    Association EURATOM-CCFE, Culham Science Centre.
    Temmerman, G. De
    Association EURATOM-FOM, Institute for Plasma Physics.
    Hole, D.
    School of Science and Technology, University of Sussex, Brighton.
    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.
    Likonen, J.
    Association EURATOM-TEKES, VTT.
    Rödig, M.
    Association EURATOM-FZJ, Forschungszentrum Jülich.
    Schmidt, A.
    Association EURATOM-FZJ, Forschungszentrum Jülich.
    Uytdenhouwen, I.
    Association EURATOM-SCK-CEN, Nuclear Research Centre.
    Hakola, A.
    Association EURATOM-TEKES, VTT.
    Semerok, A.
    CEA Saclay, DEN/DPC/SCP/LILM.
    Stamp, M.
    Association EURATOM-CCFE, Culham Science Centre.
    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.
    Vince, J.
    Association EURATOM-CCFE, Culham Science Centre.
    Comprehensive First Mirror Test for ITER at JET with Carbon Walls2010In: Proceedings of the23rd IAEA Fusion Energy Conference, 2010Conference paper (Refereed)
    Abstract [en]

    Metallic mirrors will be essential components of all optical spectroscopy and imaging systems forplasma diagnosis that will be used on the next-step magnetic fusion experiment, ITER. Any change of the mirrorperformance, in particular reflectivity, will influence the quality and reliability of detected signals. On therequest of the ITER Design Team, a First Mirror Test (FMT) has been carried out at JET during campaigns in2005-2007 and 2008-2009. To date, it has been the most comprehensive test performed with a large number oftest mirrors exposed in an environment containing both carbon and beryllium; the total plasma time (in 2005-2007 period) over 35 h including 27 h of X-point operation. 32 stainless steel and polycrystalline molybdenumflat-front and 45oangled mirrors were installed in separate channels of cassettes on the outer wall and in the MkII HD divertor: inner leg, outer leg and base plate under the load bearing tile. Post exposure studies comprisedreflectivity measurements and surface analyses with microscopy, secondary ion mass spectrometry, ion beamanalysis and energy dispersive X-ray spectroscopy.. The essential results are: (i) on the outer wall highreflectivity (~90%) is maintained for mirrors close to the channel entrance but it is degraded by 30-40 % deeperin the channel (ii) reflectivity loss by 70-90% is measured for mirrors placed in the divertor: outer, inner andbase; (iii) deuterium and carbon are the main elements detected on all mirror surfaces and the presence ofberyllium is also found; (iv) thick deposits show rough columnar structure and thickness is 1-20 μm; (v) bubblelike structures are detected in deposits; (vi) the deposition in channels in the divertor cassettes is pronounced atthe very entrance; (vii) photonic cleaning with laser removes deposits but the surface is damaged by laser pulses.In summary, reflectivity of all tested mirrors is degraded either by erosion with CX neutrals or by the formationof thick deposits. The implications of results obtained for first mirrors in next-step device are discussed andcritical assessment of various methods for in-situ cleaning of mirrors is presented. The conclusion is thatengineering solutions should be developed in order to install shutters or to implement a cassette with mirrors toreplace periodically the degraded ones

  • 10.
    Rubel, Marek J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Bailescu, V.
    Coad, J. P.
    Hirai, T.
    Likonen, J.
    Linke, J.
    Lungu, C. P.
    Matthews, G. F.
    Pedrick, L.
    Riccardo, V.
    Sundelin, Per
    Villedieu, E.
    Beryllium plasma-facing components for the ITER-Like Wall Project at JET2008In: PROCEEDINGS OF THE 17TH INTERNATIONAL VACUUM CONGRESS/13TH INTERNATIONAL CONFERENCE ON SURFACE SCIENCE/INTERNATIONAL CONFERENCE ON NANOSCIENCE AND TECHNOLOGY, 2008, Vol. 100, no PART 6Conference paper (Refereed)
    Abstract [en]

    ITER-Like Wall Project has been launched at the JET tokamak in order to study a tokamak operation with beryllium components on the main chamber wall and tungsten in the divertor. To perform this first comprehensive test of both materials in a thermonuclear fusion environment, a broad program has been undertaken to develop plasma-facing components and assess their performance under high power loads. The paper provides a concise report on scientific and technical issues in the development of a beryllium first wall at JET.

  • 11.
    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.

  • 12.
    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.

  • 13.
    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)
  • 14.
    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)
  • 15. 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.

  • 16.
    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.

  • 17.
    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.

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