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  • 1. Batistoni, P.
    et al.
    Likonen, J.
    Bekris, N.
    Brezinsek, S.
    Coad, P.
    Horton, L.
    Matthews, G.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Sips, G.
    Syme, B.
    Widdowson, A.
    The JET technology program in support of ITER2014In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 89, no 7-8, p. 896-900Article in journal (Refereed)
    Abstract [en]

    This paper presents an overview of the current and planned technological activities at JET in support of ITER operation and safety. The scope is very broad and it ranges from analysis of components from the ITER-like Wall (ILW) to determine material erosion and deposition, dust generation and fuel retention to neutronics measurements and analyses. Preliminary results are given of the post-mortem analyses of samples exposed to JET plasmas during the first JET-ILW operation in 2011-2012, and retrieved during the following in-vessel intervention. JET is the only fusion machine capable of producing significant neutron yields, up to nearly 10(19) n/s (14.1 MeV) in DT operations. Recently, the technological potential of a new DT campaign at JET in support of ITER has been explored and the outcome of this assessment is presented. The expected 14 MeV neutron yield, the use of tritium, the preparation and implementation of safety measures will provide a unique occasion to gain experience in several ITER relevant technological areas. A number of projects and experiments to be conducted in conjunction with the DT operation have been identified and they are described in this paper.

  • 2. Cavinato, M.
    et al.
    Gregoratto, D.
    Marchiori, G.
    Paccagnella, R.
    Brunsell, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Yadikin, Dmitriy
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Comparison of strategies and regulator design for active control of MHD modes2005In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 74, no 1-4, p. 549-553Article in journal (Refereed)
    Abstract [en]

    A system of evenly spaced poloidal arrays of saddle coils was recently installed on the reversed field pinch device EXTRAP T2R to perform experiments on the active control of MHD modes. The implementation of different control strategies, such as "intelligent shell" and "mode control", was made possible by a flexible digital control system. After giving some results on the performances of the innermost coil current control loop, two versions of "mode control" recently tested on the machine are presented. In the "wise shell" approach, equilibrium related modes are ruled out and a systematic increase of the pulse length is obtained. In a second, more model based, approach, a mode estimator/controller is designed aiming at a full state feedback by including modes, which are not directly measurable due to the limited number of available real-time signals.

  • 3. Coad, J. P.
    et al.
    Esser, H. -G
    Likonen, J.
    Mayer, M.
    Neill, G.
    Philipps, V.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Vince, J.
    Diagnostics for studying deposition and erosion processes in JET2005In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 74, no 1-4, p. 745-749Article in journal (Refereed)
    Abstract [en]

    Estimates of erosion, deposition and H-isotope retention in JET from previous divertor campaigns have relied on analysis of in-vessel components removed at shutdowns. The components analysed have also provided an incomplete coverage of the vessel. In 2004, new diagnostics are being installed to give a more complete picture (such as smart tiles) and to provide some time resolution. The latter includes further quartz microbalances (QMB), following the successful operation of a prototype in 2002-2004 [H.-G. Esser, G. Neill, P. Coad, G.F. Matthews, D. Jolovic, D. Wilson, M. Freisinger, V. Philipps, Quartz microbalance: a time-resolved diagnostic to measure material deposition in JET, Fusion Eng. Des. 66-68 (2003) 855-860; H.-G. Esser, V. Philipps, M. Freisinger, G.F. Matthews, J.P. Coad, G.F. Neill, JET EFDA Contributors, Effect of plasma configuration on carbon migration measured in the inner divertor of JET using quartz microbalance, J. Nucl. Mater. 337-339 (2005) 84-87], which will also have temperature control. Other diagnostics include rotating collectors and deposition monitors [M. Mayer, V. Rohde, P. Coad, P. Wienhold, ASDEX Upgrade Team, JET EFDA Contributors, Carbon erosion and migration in fusion devices, Phys. Scr. T111 (2004) 55-59]. Units are also being installed to provide information on mirrors for ITER.

  • 4. Grisolia, C.
    et al.
    Counsell, G.
    Dinescu, G.
    Semerok, A.
    Bekris, N.
    Coad, P.
    Hopf, C.
    Roth, J.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Widdowson, A.
    Tsitrone, E.
    Treatment of ITER plasma facing components: Current status and remaining open issues before ITER implementation2007In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 82, no 15-24, p. 2390-2398Article in journal (Refereed)
    Abstract [en]

    The in-vessel tritium inventory control is one of the most ITER challenging issues which has to be resolved to fulfil safety requirements. This is due mainly to the presence of carbon as a constituent of plasma facing components (PFCs) which leads to a high fuel permanent retention. For several years now, physics studies and technological developments have been undertaken worldwide in order to develop reliable techniques which could be used in ITER severe environment (magnetic field, vacuum, high temperature) for in situ tritium recovery. The scope of this contribution is to review the present status of these achievements and define the remaining work to be done in order to propose a dedicated work program. Different treatment techniques (chemical treatments, photonic cleaning) will be reviewed. In the frame of ITER, they will be compared in terms of fuel removal rate as well as surface accessibility, type of production (gas or particulates), ability to clean mixed material. And lastly, consequences of bulk trapping observed in tokamak on the techniques currently under development will be addressed.

  • 5. Grisolia, C.
    et al.
    Rosanvallon, S.
    Coad, P.
    Bekris, N.
    Braet, J.
    Brennan, D.
    Brichard, B.
    Counsell, G.
    Day, C.
    Likonen, J.
    Piazza, G.
    Poletiko, C.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Semerok, A.
    JET contributions to ITER technology issues2006In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 81, no 07-jan, p. 149-154Article in journal (Refereed)
    Abstract [en]

    The Joint European Torus (JET) fusion machine is the only device capable of operation with tritium and of handling Be and therefore is best suited to the study of tritium and fusion-related issues. A large variety of activities are performed within the JET fusion technology task force (FT-TF). In this paper, some topics such as erosion/deposition and material transport, characterisation of flakes and detritiation techniques are highlighted. Recent examples of results obtained on waste management studies are also given. Data on some ITER-relevant components that have been tested at JET, such as a pumping cryopanel and hardened optics fibers, are presented. In all fields, the work to be addressed in future JET work programmes is discussed.

  • 6. Hirai, T.
    et al.
    Linke, J.
    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.
    Coad, J. P.
    Likonen, J.
    Lungu, C. P.
    Matthews, G. F.
    Philipps, V.
    Wessel, E.
    Thermal load testing of erosion-monitoring beryllium marker tile for the ITER-Like Wall Project at JET2008In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 83, no 7-9, p. 1072-1076Article in journal (Refereed)
    Abstract [en]

    ITER-Like Wall Project has been launched at JET in order to perform a fully integrated test of plasma-facing materials. During the next major shutdown a full metal wall will be installed: tungsten in the divertor and beryllium in the main chamber. Beryllium erosion is one of key issues to be addressed. Special marker tiles have been designed for this purpose. Test coupons of such markers have been manufactured and examined. The performance test under high power deposition was carried in the electron beam facility JUDITH. The results of material characterization before and after high heat flux loads are presented. The samples survived, without macroscopic damage, power loads of up to 4.5 MW/m(2) for 10s (surface temperature similar to 650 degrees C) and 50 cyclic loads at 3.5 MW/m(2) lasting 10s each (surface temperature similar to 600 degrees C).

  • 7. Hirai, T.
    et al.
    Maier, H.
    Rubel, Marek J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Mertens, Ph
    Neu, R.
    Gauthier, E.
    Likonen, J.
    Lungu, C.
    Maddaluno, G.
    Matthews, G. F.
    Mitteau, R.
    Neubauer, O.
    Piazza, G.
    Philipps, V.
    Riccardi, B.
    Ruset, C.
    Uytdenhouwen, I.
    R&D on full tungsten divertor and beryllium wall for JET ITER-like wall project2007In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 82, no 15-24, p. 1839-1845Article in journal (Refereed)
    Abstract [en]

    The ITER reference materials have been tested separately in tokamaks, plasma simulators, ion beams and high heat flux test beds. 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. As a result of R&D projects in 2005-2006, bulk tungsten tiles are foreseen at the outer horizontal target and tungsten coating at the other divertor tiles. In some regions of the main chamber, beryllium coated Inconel tiles and bulk beryllium tiles are utilised which include marker tiles as erosion diagnostics. This paper gives an overview of the R&D carried out in the frame of the ITER-like Wall Project on the development of an inertially cooled bulk tungsten tile design and the characterization of tungsten and beryllium coating technologies.

  • 8. Leontyev, A.
    et al.
    Semerok, A.
    Farcage, D.
    Thro, P. -Y
    Grisolia, C.
    Widdowson, A.
    Coad, P.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Theoretical and experimental studies on molybdenum and stainless steel mirrors cleaning by high repetition rate laser beam2011In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 86, no 9-11, p. 1728-1731Article in journal (Refereed)
    Abstract [en]

    Our studies were aimed to determine the damage threshold of molybdenum (Mo) and stainless steel (SS) mirrors to provide the maximum fluence which the mirror surfaces could withstand without affecting their reflectivity properties. A high repetition rate ytterbium fiber laser (20 kHz, 1.06 mu m, 120 ns) was applied. The experimental single-pulse and multiple-pulse damage thresholds were obtained. To calculate damage thresholds, a 1D analytical model which takes into account the temperature dependent absorptance and multiple-pulse damage based on plastic deformations accumulation was applied. The experimental damage thresholds and the theoretical ones are in a good agreement. Cleaning tests with the contaminated mirrors exposed in JET have been performed.

  • 9. Louche, F.
    et al.
    Wauters, T.
    Ragona, R.
    Moeller, S.
    Durodie, F.
    Litnovsky, A.
    Lyssoivan, A.
    Messiaen, A.
    Ongena, J.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brezinsek, S.
    Linsmeier, Ch.
    Van Schoor, M.
    Design of an ICRF system for plasma-wall interactions and RF plasma production studies on TOMAS2017In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 123, p. 317-320Article in journal (Refereed)
    Abstract [en]

    Ion cyclotron wall conditioning (ICWC) is being developed for ITER and W7-X as a baseline conditioning technique in which the ion cyclotron heating and current drive system will be employed to produce and sustain the currentless conditioning plasma. The TOMAS project (TOroidal MAgnetized System, operated at the FZ-juelich, Germany) proposes to explore several key aspects of ICWC. For this purpose we have designed an ICRF system made of a single strap antenna within a metallic box, connected to a feeding port and a pre-matching system. We discuss the design work of the antenna system with the help of the commercial electromagnetic software CST Microwave Studio (R). The simulation results for a given geometry provide input impedance matrices for the two-port system. These matrices are afterwards inserted into various circuit models to assess the accessibility of the required frequency range. The sensitivity of the matching system to uncertainties on plasma loading and capacitance values is notably addressed. With a choice of three variable capacitors we show that the system can cope with such uncertainties. We also demonstrate that the system can cope as well with the high reflected power levels during the short breakdown phase of the RF discharge, but at the cost of a significantly reduced coupled power.

  • 10. Murari, A.
    et al.
    Edlington, T.
    Brzozowski, Jerzy H.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    De la Luna, E.
    Andrew, P.
    Arnoux, G.
    Cecil, F. E.
    Cupido, L.
    Darrow, D.
    Kiptily, V.
    Fessey, J.
    Gauthier, E.
    Hacquin, S.
    Hill, K.
    Huber, A.
    Loarer, T.
    McCormicki, K.
    Reichi, M.
    JET new diagnostic capability on the route to ITER2007In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 82, no 14-maj, p. 1161-1166Article in journal (Refereed)
    Abstract [en]

    The JET scientific programme is directed towards the development of ITER relevant scenarios. In support of this, significant effort has been made to develop diagnostics to better characterise the power deposition on the plasma facing components, to investigate in more detail the radiation losses particularly in the divertor region and to better detect Magneto Hydrodynamic Modes (MHD) instabilities and their effects on fast ion confinement. A new wide-angle infrared camera provides for the first time the opportunity to perform infrared thermography in the JET main chamber, even during fast events like ELMs and disruptions. A completely new bolometric system, with better spatial resolution particularly in the divertor, is now used to investigate the total radiation losses and their influence on the ELM behaviour. A new set of microwave waveguides has improved by 20 dB the signal to noise ratio of the JET X-mode reflectometers, that are now routinely used to detect MHD instabilities and in particular to localise the location of Alfven Eigenmodes. This improved diagnostic capability to monitor MHD instabilities is complemented by two new diagnostics to detect lost fast particles. Both the new scintillator probe and a poloidal array of Faraday cups have already shown clear correlations between MHD activity and ion losses at the edge.

  • 11.
    Olofsson, Erik
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brunsell, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Controlled magnetohydrodynamic mode sustainment in the reversed-field pinch: Theory, design and experiments2009In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 84, no 7-11, p. 1455-1459Article in journal (Refereed)
    Abstract [en]

    A novel control system design for magnetohydrodynamic (MHD) resistive-wall mode (RWM) stabilization is developed from the viewpoint of process control. The engineering approach assumed consists of system identification, selection of feedback interconnections, and subsequently, associated feedback gain tuning. A design for general output tracking is devised, implemented and experimentally verified to be capable of sustaining MHD modes in the reversed-field pinch (RFP) machine EXTRAP-T2R. In principle, by active feedback. the plasma column boundary is forced to 'user-specified' helicities of prescribed amplitudes and phases. Experimental success is mainly attributed to careful identification of local magnetic field diffusion time-constants, and individual actuator channel peripheral dynamics. Addition of functionality and key features of this new MHD feedback system software might provide a versatile tool for experimental plasma dynamics and innovative MHD stability research.

  • 12.
    Olofsson, K. Erik J.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brunsell, Per R.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Drake, James R.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Measurements of the vacuum-plasma response in EXTRAP T2R using generic closed-loop subspace system identification2012In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 87, no 12, p. 1926-1929Article in journal (Refereed)
    Abstract [en]

    A multibatch formulation of a multi-input multi-output closed-loop subspace system identification method is employed for the purpose of obtaining control-relevant models of the vacuum-plasma response in the magnetic confinement fusion experiment EXTRAP T2R. The accuracy of the estimate of the plant dynamics is estimated by computing bootstrap replication statistics of the dataset. It is seen that the thus identified models exhibit both predictive capabilities and physical spectral properties.

  • 13. Oya, Y.
    et al.
    Masuzaki, S.
    Tokitani, M.
    Azuma, K.
    Oyaidzu, M.
    Isobe, K.
    Asakura, N.
    Widdowson, A. M.
    Heinola, K.
    Jachmich, S.
    Rubel, Marek
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Contributors, JET
    Correlation of surface chemical states with hydrogen isotope retention in divertor tiles of JET with ITER-Like Wall2018In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 132, p. 24-28Article in journal (Refereed)
    Abstract [en]

    To understand the fuel retention mechanism correlation of surface chemical states and hydrogen isotope retention behavior determined by XPS (X-ray photoelectron spectroscopy) and TDS (Thermal desorption spectroscopy), respectively, for JET ITER-Like Wall samples from operational period 2011–2012 were investigated. It was found that the deposition layer was formed on the upper part of the inner vertical divertor area. At the inner plasma strike point region, the original surface materials, W or Mo, were found, indicating to an erosion-dominated region, but deposition of impurities was also found. Higher heat load would induce the formation of metal carbide. At the outer horizontal divertor tile, mixed material layer was formed with iron as an impurity. TDS showed the H and D desorption behavior and the major D desorption temperature for the upper part of the inner vertical tile was located at 370 °C and 530 °C. At the strike point region, the D desorption temperature was clearly shifted toward higher release temperatures, indicating the stabilization of D trapping by higher heat load.

  • 14. Romanelli, Francesco
    et al.
    Laxåback, Martin
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Durodie, Frederic
    Horton, Lorne
    Lehnen, Michael
    Murari, Andrea
    Rimini, Fernanda
    Sips, George
    Zastrow, Klaus-Dieter
    The role of JET for the preparation of the ITER exploitation2011In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 86, no 6-8, p. 459-464Article in journal (Refereed)
    Abstract [en]

    The JET programme is devoted to the consolidation of ITER design choices and the qualification of ITER integrated regimes of operation. During the experimental campaigns carried out in 2008 and 2009 attention focussed on the test of the ITER-like ICRH antenna, the ITER scenario preparation, the verification of the adequacy of the ITER poloidal field coil design and the test of disruption mitigation methods such as massive gas injection. From 2011 the new ITER-like wall with all beryllium and tungsten plasma facing components, the neutral beam power upgrade and the enhanced control and diagnostic capability will allow key questions on plasma-wall interactions, fuel retention and plasma impurity control with the foreseen ITER wall materials to be addressed. Finally, feasibility studies have confirmed the option of installing an ITER-technology based 170 GHz/10 MW electron cyclotron resonance heating system for the control of MHD activity and the development of advanced tokamak scenarios, and 32 in-vessel coils for ELM control capable of producing magnetic perturbation spectra with a Chirikov parameter above unity for plasma currents up to 5 MA. During the ITER construction phase, JET will be the only device of its class in operation and will therefore play a key role in the preparation of ITER operations - saving time and reducing risk from the ITER programme.

  • 15.
    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.
    Ivanova, Darya
    KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Philipps, V.
    Zlobinski, M.
    Huber, A.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Schweer, B.
    Efficiency of fuel removal techniques tested on plasma-facing components from the TEXTOR tokamak2012In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 87, no 5-6, p. 935-940Article in journal (Refereed)
    Abstract [en]

    An overview of several techniques considered for fuel and co-deposits removal is given. The methods were tested both on plasma-facing components from the TEXTOR tokamak and on laboratory-prepared layers: (a) chemical approach based on oxidative or nitrogen-assisted plasma; (b) photonic methods with laser-induced fuel desorption or ablation of co-deposits; (c) thermal desorption in vacuum or under oxidative conditions at a broad range of temperatures. The emphasis is on outstanding issues associated with every technique aiming at the reduction of fuel content: the efficiency of fuel and co-deposit removal, the surface state of PFC following the treatment and dust generation.

  • 16.
    Rubel, Marek J.
    et al.
    KTH, Superseded Departments, Alfvén Laboratory.
    Brunsell, Per R.
    Duwe, R.
    Linke, J.
    Molybdenum limiters for Extrap-T2 upgrade: surface properties and high heat flux testing2000In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 49, p. 323-329Article in journal (Refereed)
    Abstract [en]

    A vacuum vessel, a conductive copper shell and plasma-facing components of the Extrap-T2 device, a medium size reversed field pinch, are under major rebuild. The new machine is equipped with an array of 180 molybdenum limiters, which will be exposed to power loads of about 30 MW m(-2), but higher loads can not be excluded. Prior to the limiters' installation in the vessel, they were tested under high heat loads in the JUDITH electron beam facility in order to assess the possible damage to the surface and to the bulk of the material. A test limiter was irradiated with a beam of increasing power density from 15 to 1500 MW m(-2). Surface characterization was performed before and after the irradiation using electron and optical microscopy, energy dispersive X-ray spectroscopy, enhanced proton scattering and laser profilometry. Metallography studies were performed for the irradiated areas. The irradiation induced the change in surface morphology, e.g. surface melting and re-crystallization of grains, only following the 10 ms long pulses with the absorbed power density approaching 1500 MW m(-2).

  • 17.
    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.

  • 18.
    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.

  • 19.
    Rubel, Marek
    et al.
    KTH.
    Widdowson, A.
    Culham Sci Ctr, Culham Ctr Fus Energy, Abingdon OX14 3DB, Oxon, England..
    Grzonka, J.
    Warsaw Univ Technol, PL-02507 Warsaw, Poland.;Inst Elect Mat Technol, PL-01919 Warsaw, Poland..
    Fortuna-Zalesna, E.
    Warsaw Univ Technol, PL-02507 Warsaw, Poland..
    Moon, Sunwoo
    KTH.
    Petersson, Per
    KTH.
    Ashikawa, N.
    Natl Inst Fus Sci, Toki, Gifu 5095292, Japan..
    Asakura, N.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Hamaguchi, D.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Hatano, Y.
    Toyama Univ, Hydrogen Isotope Res Ctr, Gofuku 3190, Toyama 9308555, Japan..
    Isobe, K.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Masuzaki, S.
    Natl Inst Fus Sci, Toki, Gifu 5095292, Japan..
    Kurotaki, H.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Oya, Y.
    Shizuoka Univ, Suruga Ku, 836 Ohya, Shizuoka 4228529, Japan..
    Oyaidzu, M.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Tokitani, M.
    Natl Inst Fus Sci, Toki, Gifu 5095292, Japan..
    Dust generation in tokamaks: Overview of beryllium and tungsten dust characterisation in JET with the ITER-like wall2018In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 136, p. 579-586Article in journal (Refereed)
    Abstract [en]

    Operation of the JET tokamak with beryllium and tungsten ITER-like wall provides unique opportunity for detailed studies on dust generation: quantity, morphology, location, etc. The programme carried out in response to ITER needs for safety assessment comprises: (i) remotely controlled vacuum cleaning of the divertor; (ii) local sampling of loosely bound matter from plasma-facing components (PFC); (iii) collection of mobilized dust on various erosion-deposition probes located in the divertor and in the main chamber. Results of comprehensive analyses performed by a number of complementary techniques, e.g. a range of microscopy methods, electron and ion spectroscopy, liquid scintillography and thermal desorption, are summarized by following points: (a) Total amount of dust collected by vacuum cleaning after three campaigns is about 1-1.4 g per campaign (19.1-23.5 h plasma operation), i.e. over 100 times smaller than in JET operated with carbon walls (i.e. in JET-C). (b) Two major categories of Be dust are identified: flakes of co-deposits formed on PFC and droplets (2-10 mu m in diameter). Small quantifies, below 1 g, of Be droplets and splashes are associated mainly with melting of beryllium limiters.

  • 20.
    Setiadi, A. C.
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brunsell, Per R.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Villone, F.
    Mastrostefano, S.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Gray-box modeling of resistive wall modes with vacuum-plasma separation and optimal control design for EXTRAP T2R2017In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 121, p. 245-255Article in journal (Refereed)
    Abstract [en]

    This paper presents a graybox methodology to model the resistive wall mode instability by combining first principle approach and system identification technique. In particular we propose a separate vacuum and plasma modeling with cascade interconnection. The shell is modeled using CARIDDI code which solves the 3D integral formulation of eddy current problem, whereas the plasma response is obtained empirically by system identification. Furthermore the resulting model is used to design an optimal feedback control. The model and feedback control is validated experimentally in EXTRAP T2R reversed-field pinch, where RWMs stabilization and non-axisymmetric mode sustainment is considered. 

  • 21. Suttrop, W.
    et al.
    Gruber, O.
    Guenter, S.
    Hahn, D.
    Herrmann, A.
    Rott, M.
    Vierle, T.
    Seidel, U.
    Sempf, M.
    Streibl, B.
    Strumberger, E.
    Yadikin, D.
    Neubauer, O.
    Unterberg, B.
    Gaio, E.
    Toigo, V.
    Brunsell, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    In-vessel saddle coils for MHD control in ASDEX Upgrade2009In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 84, no 2-6, p. 290-294Article in journal (Refereed)
    Abstract [en]

    A set of 24 in-vessel saddle coils is planned for MHD control experiments in ASDEX Upgrade. These coils can produce static and alternating error fields for suppression of Edge Localised Modes, locked mode rotation control and, together with additional conducting wall elements, resistive wall mode excitation and feedback stabilisation experiments. All of these applications address critical physics issues for the operation of ITER. This extension is implemented in several stages, starting with two poloidally separated rings of eight toroidally distributed saddle coils above and below the outer midplane. In stages 2 and 3, eight midplane coils around the large vessel access ports and 12 AC Power converters are added, respectively. Finally (stage 4), the existing passive stabilising loop (PSL), a passive conductor for vertical growth rate reduction, will be complemented by wall elements that allow helical Current patterns to reduce the RWM growth rate for active control within the accessible bandwidth. The system is capable of producing error fields with toroidal mode number n = 4 for plasma edge ergodisation with core island width well below the neo-classical tearing mode seed island width even without rotational shielding. Phase variation between the three toroidal coil rings allows to create or avoid resonances with the plasma safety factor profile, in order to test the importance of resonances for ELM suppression.

  • 22. 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

  • 23. 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.

  • 24.
    Tokitani, M.
    et al.
    Natl Inst Fus Sci, 322-6 Oroshi, Toki, Gifu 5095292, Japan..
    Miyamoto, M.
    Shimane Univ, Matsue, Shimane 6908504, Japan..
    Masuzaki, S.
    Natl Inst Fus Sci, 322-6 Oroshi, Toki, Gifu 5095292, Japan..
    Sakamoto, R.
    Natl Inst Fus Sci, 322-6 Oroshi, Toki, Gifu 5095292, Japan..
    Oya, Y.
    Shizuoka Univ, Shizuoka 4228529, Japan..
    Hatano, Y.
    Univ Toyama, Toyama 9308555, Japan..
    Otsuka, T.
    Kindai Univ, Higashiosaka, Osaka 5778502, Japan..
    Oyaidzu, M.
    QST, Aomori 0393212, Japan..
    Kurotaki, H.
    QST, Aomori 0393212, Japan..
    Suzuki, T.
    QST, Aomori 0393212, Japan..
    Hamaguchi, D.
    QST, Aomori 0393212, Japan..
    Isobe, K.
    QST, Aomori 0393212, Japan..
    Asakura, N.
    QST, Aomori 0393212, Japan..
    Widdowson, A.
    EUROfus Consortium, Culham Sci Ctr, JET, Abingdon OX14 3DB, Oxon, England..
    Heinola, K.
    Univ Helsinki, POB 64, FI-00560 Helsinki, Finland..
    Rubel, Marek
    KTH. Royal Inst Technol KTH, S-10044 Stockholm, Sweden..
    Plasma-wall interaction on the divertor tiles of JET ITER-like wall from the viewpoint of micro/nanoscopic observations2018In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 136, p. 199-204Article in journal (Refereed)
    Abstract [en]

    Micro/nanoscopic observations on the surface of the divertor tiles used in the first campaign (2011-2012) of the JET tokamak with ITER-like Wall (JET ILW) have been carried out by means of several material analysis techniques. Previous results from the inner divertor were reported for a single poloidal section of the tile numbers 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. The formation of the thick stratified mixed-material deposition layer on tiles 1 and 4, and erosion on tile 3 were identified. This study is mostly focused on the outer divertor: tiles 6, 7 and 8. In contrast to the inner tile, remarkable surface modifications have not been observed on the vertical target (tiles 7 and 8) where sputtering erosion and impurity deposition would have been almost balanced. Only a specific part of tile 6 (horizontal target) located near the exhaust channel was covered with a stratified ("geological-like") mixed-material deposition layer which mainly included Be and Ni with the thickness of similar to 2 mu m. Special feature of this mixed layer was that a certain amount of nitrogen (N) was clearly detected in the layer. Since the concentration of N varied with the depth position, it could be depended on the amount of that gas puffed for plasma edge cooling during the JET experimental campaign. In addition to the outer divertor tiles, a very interesting feature of the local erosion and deposition effects is reported in this paper.

  • 25.
    Tolias, Panagiotis
    KTH, School of Electrical Engineering and Computer Science (EECS), Space and Plasma Physics.
    Lifshitz calculations of Hamaker constants for fusion relevant materials2018In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 133, p. 110-116Article in journal (Refereed)
    Abstract [en]

    The determination of the Hamaker constant is necessary for the quantification of the van der Waals force and thus of dust-wall adhesion. Here Lifshitz theory is employed for the calculation of the non-retarded Hamaker constants of ten common dust-wall material combinations. Extended-in-frequency reliable dielectric data are employed and two independent computational methods are considered for the calculation of the dielectric function at the imaginary Matsubara frequencies. The Hamaker constant for tungsten-on-tungsten is the largest calculated, which implies the strongest adhesion. The Hamaker constant for graphite-on-graphite is much smaller than tungsten-on-tungsten and even beryllium-on-beryllium. Copper, chromium and especially aluminium are identified to be proper adhesive proxies of beryllium.

  • 26. Vizvary, Z.
    et al.
    Bourdel, B.
    Garcia Carrasco, Alvaro
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Lam, N.
    Leipold, F.
    Pitts, R. A.
    Reichle, R.
    Riccardo, V.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    De Temmerman, G.
    Thompson, V.
    Widdowson, A.
    Engineering design and analysis of an ITER-like first mirror test assembly on JET2017In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 123, p. 1054-1057Article in journal (Refereed)
    Abstract [en]

    The ITER first mirrors are the components of optical diagnostic systems closest to the plasma. Deposition may build up on the surfaces of the mirror affecting their ability to fulfil their function. However, physics modelling of this layer growth is fraught with uncertainty. A new experiment is underway on JET, under contract to ITER, with primary objective to test if, under realistic plasma and wall material conditions and with ITER-like first mirror aperture geometry, deposits do grow on first mirrors. This paper describes the engineering design and analysis of this mirror test assembly. The assembly was installed in the 2014-15 shutdown and will be removed in the 2016-17 shutdown.

  • 27. Weinzettl, V.
    et al.
    Matejicek, J.
    Ratynskaia, Svetlana V.
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Tolias, Panagiotis
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    De Angeli, M.
    Riva, G.
    Dimitrova, M.
    Havlicek, J.
    Adamek, J.
    Seidl, J.
    Tomes, M.
    Cavalier, J.
    Imrisek, M.
    Havranek, A.
    Panek, R.
    Peterka, M.
    Dust remobilization experiments on the COMPASS tokamak2017In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 124, p. 446-449Article in journal (Refereed)
    Abstract [en]

    Dust remobilization is one of the not yet fully understood mechanisms connected to the prompt erosion of material from plasma facing surfaces in fusion devices. As a part of a newly initiated cross-machine study, dust remobilization experiments have been performed on the COMPASS tokamak. Tungsten samples with well-defined deposited tungsten dust grains, prepared using a recently developed controlled pre-adhesion method, have been exposed to ELMy H-mode discharges as well as L-mode discharges with forced disruptions. Here we report on the technical aspects of the experiment realization as well as on the experimental results of dust remobilization. The latter is discussed in the light of data from other machines and a physical interpretation is suggested for the observed spatial localization of the dust remobilization activity. Evidence of rearrangement of isolated dust into clusters and strings is also presented.

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

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

1 - 28 of 28
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