Alcohol electrolysis using polymeric membrane electrolytes is a promising route for storing excess renewable energy in hydrogen, alternative to the thermodynamically limited water electrolysis. By properly choosing the ionic agent (i.e. H+ or OH) and the catalyst support, and by tuning the catalyst structure, we developed membrane-electrode-assemblies which are suitable for cost-effective and efficient alcohol electrolysis. Novel porous electrodes were prepared by Atomic Layer Deposition (ALD) of Pt on a TiO2-Ti web of microfibers and were interfaced to polymeric membranes with either H+ or OH conductivity. Our results suggest that alcohol electrolysis is more efficient using OH conducting membranes under appropriate operation conditions (high pH in anolyte solution). ALD enables better catalyst utilization while it appears that the TiO2-Ti substrate is an ideal alternative to the conventional carbon-based diffusion layers, due to its open structure. Overall, by using our developmental anodes instead of commercial porous electrodes, the performance of the alcohol electrolyser (normalized per mass of Pt) can be increased up to ~30 times.

VL - 44 IS - 21 U1 -MaSF

U2 -CEPEA

U5 - add57380c809f09666243428989840ed ER - TY - JOUR T1 - Operational characteristics of the superconducting high flux plasma generator Magnum-PSI JF - Fusion Engineering and Design Y1 - 2018 A1 - van de Pol, M. J. A1 - Alonso van der Westen, S. A1 - Aussems, D. U. B. A1 - van den Berg, M. A. A1 - Brons, S. A1 - van Eck, H. J. N. A1 - van Eden, G. G. A1 - Genuit, J. W. A1 - van der Meiden, H. J. A1 - Morgan, T. W. A1 - Scholten, J. A1 - Vernimmen, J. W. M. A1 - Vos, E. G. P. A1 - de Baar, M. R. KW - ELMs KW - ITER KW - Linear plasma device KW - Plasma-surface interactions AB -The interaction of intense plasma impacting on the wall of a fusion reactor is an area of high and increasing importance in the development of electricity production from nuclear fusion. In the Magnum-PSI linear device, an axial magnetic field confines a high density, low temperature plasma produced by a wall stabilized DC cascaded arc into an intense magnetized plasma beam directed onto a target. The experiment has shown its capability to reach conditions that enable fundamental studies of plasma-surface interactions in the regime relevant for fusion reactors such as ITER: 1023–1025 m−2s−1 hydrogen plasma flux densities at 1–5 eV for tens of seconds by using conventional electromagnets. Recently the machine was upgraded with a superconducting magnet, enabling steady-state magnetic fields up to 2.5 T, expanding the operational space to high fluence capabilities for the first time. Also the diagnostic suite has been expanded by a new 4-channel resistive bolometer array and ion beam analysis techniques for surface analysis after plasma exposure of the target. A novel collective Thomson scattering system has been developed and will be implemented on Magnum-PSI. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.

VL - 136 IS - Part A U1 -FP

U2 -FFI

U3 - FP75 U5 - 2016e06893f98e992bb6c70b5e7cd7ea ER - TY - JOUR T1 - Observation of enhanced ion particle transport in mixed H/D isotope plasmas on JET JF - Nuclear Fusion Y1 - 2018 A1 - Maslov, M. A1 - King, D. A1 - Viezzer, E. A1 - Keeling, D. L. A1 - Giroud, C. A1 - Tala, T. A1 - Salmi, A. A1 - Marin, M. A1 - Citrin, J. A1 - Bourdelle, C. A1 - Solano, E. R. A1 - JET Contributors AB -Particle transport in tokamak plasmas has been intensively studied in the past, particularly in relation to density peaking and the presence of anomalous inward particle convection in L- and H-modes. While in the L-mode case the presence of the anomalous inward pinch has previously been unambiguously demonstrated, particle transport in the H-mode was unclear. The main difficulty of such studies is that particle diffusion and convection could not be measured independently in steady-state conditions in the presence of a core particle flux. Therefore, it is usually not possible to separate the transport effect(inward convection), from the source effect (slow diffusion of particles introduced to the plasma core by neutral beam injection heating). In this work we describe experiments done on JET with mixtures of two hydrogenic isotopes: H and D. It is demonstrated that in the case of several ion species, convection and diffusion can be separated in a steady plasma without implementation of perturbative techniques such as gas puff modulation. Previous H-mode density peaking studies suggested that for this relatively high electron collisionality plasma scenario, the observed density gradient is mostly driven by particle source and low particle diffusivity D < 0.5 * χ eff. Transport coefficients derived from observation of the isotope profiles in the new experiments far exceed that value—ion particle diffusion is found to be as high as D ≥ 2 * χ eff, combined with a strong inward convection. Apparent disagreement with previous findings was explained by significantly faster transport of ion components with respect to the electrons, which could not be observed in a single main ion species plasma. This conclusion is confirmed by quasilinear gyrokinetic simulations.

VL - 58 IS - 7 U1 -FP

U2 -IMT

U5 - 811540f379fb24aad4246914a85abf67 ER - TY - JOUR T1 - Oxygen evolution reaction in nanoconfined carbon nanotubes JF - Physica E: Low-dimensional Systems and Nanostructures Y1 - 2018 A1 - Li, Y. A1 - Lu, X. A1 - Li, Y. F. A1 - Zhang, X. Q. KW - Carbon nanotube KW - CONFINEMENT KW - Overpotential KW - oxygen evolution reaction KW - water splitting AB - Improving oxygen electrochemistry through nanoscopic confinement has recently been highlighted as a promising strategy. In-depth understanding the role of confinement is therefore required. In this study, we simulate the oxygen evolution reaction (OER) on iron oxide nanoclusters under confinement of (7,7) and (8,8) armchair carbon nanotubes (CNTs). The free energies of the four proton coupled electron transfer (PCET) steps and the OER overpotentials are calculated. The Fe4O6 nanocluster confined in (7,7) CNT is found to be the most active for OER among the systems considered in this work. This leads to an increase in catalytic efficiency of OER compared to the hematite (110) surface, which was reported recently as an active surface towards OER. The calculated results show that the OER overpotential depends strongly on the magnetic properties of the iron oxide nanocluster. These findings are helpful for experimental design of efficient catalyst for water splitting applications. VL - 99 U1 - MaSF U2 - EMI U5 - 52805b73da067243ba85dbdcbbad155c ER - TY - JOUR T1 - Overview of the JET results in support to ITER JF - Nuclear Fusion Y1 - 2017 A1 - X. Litaudon A1 - Abduallev, S. A1 - Abhangi, M. A1 - Citrin, J. A1 - den Harder, N. A1 - Hogeweij, G. M. D. A1 - Jaulmes, F. A1 - Shumack, A. A1 - Tsalas, M. A1 - van Rooij, G. J. A1 - et al. AB - The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H = 1 at β N ~ 1.8 and n/n GW ~ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D–T campaign and 14 MeV neutron calibration strategy are reviewed. VL - 57 IS - 10 U1 - FP U2 - IMT U5 - 38b74a22f1f4822d531a5a047a2ecc1c ER - TY - JOUR T1 - Overview of ASDEX Upgrade results JF - Nuclear Fusion Y1 - 2017 A1 - Kallenbach, A. A1 - ASDEX-Upgrade Team A1 - EUROfusion MST1 Team AB - The ASDEX Upgrade (AUG) programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. Since 2015, AUG is equipped with a new pair of 3-strap ICRF antennas, which were designed for a reduction of tungsten release during ICRF operation. As predicted, a factor two reduction on the ICRF-induced W plasma content could be achieved by the reduction of the sheath voltage at the antenna limiters via the compensation of the image currents of the central and side straps in the antenna frame. There are two main operational scenario lines in AUG. Experiments with low collisionality, which comprise current drive, ELM mitigation/suppression and fast ion physics, are mainly done with freshly boronized walls to reduce the tungsten influx at these high edge temperature conditions. Full ELM suppression and non-inductive operation up to a plasma current of I p=0.8 could be obtained at low plasma density. Plasma exhaust is studied under conditions of high neutral divertor pressure and separatrix electron density, where a fresh boronization is not required. Substantial progress could be achieved for the understanding of the confinement degradation by strong D puffing and the improvement with nitrogen or carbon seeding. Inward/outward shifts of the electron density profile relative to the temperature profile effect the edge stability via the pressure profile changes and lead to improved/decreased pedestal performance. Seeding and D gas puffing are found to effect the core fueling via changes in a region of high density on the high field side (HFSHD).The integration of all above mentioned operational scenarios will be feasible and naturally obtained in a large device where the edge is more opaque for neutrals and higher plasma temperatures provide a lower collisionality. The combination of exhaust control with pellet fueling has been successfully demonstrated. High divertor enrichment values of nitrogen E N>=10 have been obtained during pellet injection, which is a prerequisite for the simultaneous achievement of good core plasma purity and high divertor radiation levels. Impurity accumulation observed in the all-metal AUG device caused by the strong neoclassical inward transport of tungsten in the pedestal is expected to be relieved by the higher neoclassical temperature screening in larger devices. VL - 57 IS - 10 U1 - PSI U2 - PSI-E U5 - b4f602f6528932aa5c7da3a1b079b2ae ER - TY - JOUR T1 - Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution JF - Nuclear Fusion Y1 - 2017 A1 - Meyer, H. A1 - Eich, T. A1 - Citrin, J. A1 - Classen, I. A1 - Hogeweij, D. A1 - Jaulmes, F. A1 - Kappatou, A. A1 - van den Brand, H. A1 - Vanovac, B. A1 - Vijvers, W. A. J. A1 - Westerhof, E. A1 - et al. AB - Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n = 2 RMP maintaining good confinement H H(98,y2) =approx 0.95. Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes. VL - 57 IS - 10 U1 - FP U2 - PEPD U5 - f781e58d912e0c330cdf9b05c806267b ER - TY - JOUR T1 - Overview of the TCV tokamak program: scientific progress and facility upgrades JF - Nuclear Fusion Y1 - 2017 A1 - Coda, S. A1 - Ahn, J. A1 - Albanese, R. A1 - Alberti, S. A1 - Alessi, E. A1 - Citrin, J. A1 - Hogeweij, D. A1 - Vijvers, W. A. J. A1 - EUROfusion MST1 Team A1 - et al. AB - The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation. VL - 57 IS - 10 U1 - FP U2 - PEPD U5 - 2a93b89d78ec281b0b5b9970ec417422 ER - TY - JOUR T1 - Oscillatory vapour shielding of liquid metal walls in nuclear fusion devices JF - Nature Communications Y1 - 2017 A1 - van Eden, G. G. A1 - Kvon, V. A1 - van de Sanden, M. C. M. A1 - Morgan, T. W. AB - Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically reduced over time due to progressing damage accumulation. Liquid metals, however, are now gaining interest in solving the challenge of extreme heat flux hitting the reactor walls. A key advantage of liquid metals is the use of vapour shielding to reduce the plasma exhaust. Here we demonstrate that this phenomenon is oscillatory by nature. The dynamics of a Sn vapour cloud are investigated by exposing liquid Sn targets to H and He plasmas at heat fluxes greater than 5 MW m−2. The observations indicate the presence of a dynamic equilibrium between the plasma and liquid target ruled by recombinatory processes in the plasma, leading to an approximately stable surface temperature. VL - 8 U1 - PSI U2 - PMI U5 - 1e2cf6044a6f47c20361bf3a9375d6e0 ER - TY - JOUR T1 - Optimization of ICRH for core impurity control in JET-ILW JF - Nuclear Fusion Y1 - 2016 A1 - Lerche, E. A1 - Goniche, M. A1 - Jacquet, P. A1 - Van Eester, D. A1 - Bobkov, V. A1 - Colas, L. A1 - Giroud, C. A1 - Monakhov, I. A1 - Casson, F. J. A1 - Tsalas, M. A1 - Rimini, F. A1 - Angioni, C. A1 - Baruzzo, M. A1 - Blackman, T. A1 - Brezinsek, S. A1 - Brix, M. A1 - Czarnecka, A. A1 - Crombe, K. A1 - Challis, C. A1 - Dumont, R. A1 - Eriksson, J. A1 - Fedorczak, N. A1 - Graham, M. A1 - Graves, J. P. A1 - Gorini, G. A1 - Hobirk, J. A1 - Joffrin, E. A1 - Johnson, T. A1 - Kazakov, Y. A1 - Kiptily, V. A1 - Krivska, A. A1 - Lennholm, M. A1 - Lomas, P. A1 - Maggi, C. A1 - Mantica, P. A1 - Mathews, G. A1 - Mayoral, M. A1 - Meneses, L. A1 - Mlynar, J. A1 - Monier-Garbet, P. A1 - Nave, M. F. A1 - Noble, C. A1 - Nocente, M. A1 - Nunes, I. A1 - Ongena, J. A1 - Petravich, G. A1 - Petrzilka, V. A1 - Putterich, T. A1 - Reich, M. A1 - Santala, M. A1 - Solano, E. R. A1 - Shaw, A. A1 - Sips, G. A1 - Stamp, M. A1 - Tardocchi, M. A1 - Valisa, M. A1 - JET Contributors AB -Ion cyclotron resonance frequency (ICRF) heating has been an essential component in the development of high power H-mode scenarios in the Jet European Torus ITER-like wall (JET-ILW). The ICRF performance was improved by enhancing the antenna-plasma coupling with dedicated main chamber gas injection, including the preliminary minimization of RF-induced plasma-wall interactions, while the RF heating scenarios where optimized for core impurity screening in terms of the ion cyclotron resonance position and the minority hydrogen concentration. The impact of ICRF heating on core impurity content in a variety of 2.5 MA JET-ILW H-mode plasmas will be presented, and the steps that were taken for optimizing ICRF heating in these experiments will be reviewed.

VL - 56 UR - http://www.euro-fusionscipub.org/wp-content/uploads/2015/09/WPJET1PR1528.pdf IS - 3 U1 -FP

U2 -TP

U5 - d602ccdf3e42dd82b551d41759691058 ER - TY - JOUR T1 - Observation of a helium ion energy threshold for retention in tungsten exposed to hydrogen/helium mixture plasma JF - Nuclear Fusion Y1 - 2016 A1 - Thompson, M. A1 - Deslandes, A. A1 - Morgan, T. W. A1 - Elliman, R. G. A1 - De Temmerman, G. A1 - Kluth, P. A1 - Riley, D. A1 - Corr, C. S. AB - Helium retention is measured in tungsten samples exposed to mixed H/He plasma in the Magnum-PSI linear plasma device. It is observed that there is very little He retention below helium ion impact energies of 9.0 +- 1.4 eV, indicating the existence of a potential barrier which must be overcome for implantation to occur. The helium retention in samples exposed to plasma at temperatures >1000 K is strongly correlated with nano-bubble formation measured using grazing incidence small-angle x-ray scattering. The diameters of nano-bubbles were not found to increase with increasing helium concentration, indicating that additional helium must be accommodated by increasing the bubble concentration or an increase in bubble pressure. For some samples pre-irradiation with heavy ions of 2.0 MeV energy is investigated to simulate the effects of neutron damage. It is observed that nano-bubble sizes are comparable between samples pre-irradiated with heavy-ions, and those without heavy-ion pre-irradiation. VL - 56 IS - 10 U1 - PSI U2 - PSI-E U3 - FP75 U5 - 03f63f3601160a3f8dae16f93290c382 ER - TY - JOUR T1 - Oxygen Evolution at Hematite Surfaces: The Impact of Structure and Oxygen Vacancies on Lowering the Overpotential JF - Journal of Physical Chemistry C, The Y1 - 2016 A1 - Zhang, X. A1 - Klaver, P. A1 - van Santen, R. A1 - van de Sanden, M. C. M. A1 - Bieberle, A. AB -Simulations of the oxygen evolution reaction (OER) are essential for understanding the limitations of water splitting. Most research has focused so far on the OER at flat metal oxide surfaces. The structure sensitivity of the OER has, however, recently been highlighted as a promising research direction. To probe the structure sensitivity, we investigate the OER at 11 hematite (Fe2O3) surfaces with density functional theory + Hubbard U (DFT+U) calculations. The results show that the O–O coupling (O–O bond formation via two adjacent terminal Os at dual site) OER mechanism at the (110) surface is competing with the mechanism of OOH formation at single site. We study the effects of surface orientation (110 vs 104), active surface sites (bridge vs terminal site), presence of surface steps and oxygen vacancy concentration on the OER and explore strategies to reduce the OER overpotential. It is found that the oxygen vacancy concentration is the most effective parameter in reducing the overpotential. In particular, an overpotential of as low as 0.47 V is obtained for the (110) surface with an oxygen vacancy concentration of 1.26 vacancies/nm2.

VL - 120 IS - 32 U1 -MaSF

U2 -EMI

U3 - FP00 U5 - 455304e46dd8843ab23bd71606c5f6ff ER - TY - JOUR T1 - Optimizing the parameter space for increased crystallinity of silicon nanoparticles grown in the gas phase JF - Physica Status Solidi A - Applications and Materials Science Y1 - 2016 A1 - Mohan, A. A1 - Schropp, R. E. I. A1 - Poulios, I. A1 - Goedheer, W. J. A1 - Rath, J. K. AB - see also: back cover VL - 213 IS - 7 U1 - FP U2 - TP U3 - FP120 U5 - 2e1b686fbc434471e93278dbc20dabc3 ER - TY - JOUR T1 - Orientation Sensitivity of Oxygen Evolution Reaction on Hematite JF - Journal of Physical Chemistry C, The Y1 - 2016 A1 - Zhang, X. Q. A1 - Cao, C. A1 - Bieberle, A. AB -The sensitivity of the surface orientation on photoelectrochemical water oxidation has recently been reported by experimental studies. However, a detailed theoretical understanding is still missing. Density functional theory + Hubbard U (DFT + U) calculations are therefore carried out in order to investigate the oxygen evolution reaction (OER) on hematite (Fe2O3) surfaces for five surface orientations, namely (100), (210), (101), (021), and (211). The free energies of four proton-coupled electron transfer steps and the OER overpotential were calculated, and the trend in activity was analyzed. For the (100) orientation, two adsorbate–adsorbate distances were studied. Interestingly, a very low overpotential of 0.52 V was found for the (100) surface with a bridge site (adsorbate on a bridge of two Fe atoms) configuration that benefited from adsorbate–adsorbate interactions.

VL - 120 IS - 50 U1 -MaSF

U2 -EMI

U5 - 00a64aabc44b05c83a65bd7d21a75fd2 ER - TY - JOUR T1 - Overview of MAST results JF - Nuclear Fusion Y1 - 2015 A1 - Chapman, I.T. A1 - Adamek, J. A1 - Akers, R. J. A1 - Allan, S. A1 - Appel, L. A1 - Asunta, O. A1 - Barnes, M. A1 - N. Ben Ayed A1 - Hawke, J. A1 - Bigelow, T. A1 - Boeglin, W. A1 - Bradley, J. A1 - Brünner, J. A1 - Cahyna, P. A1 - Carr, M. A1 - Caughman, J. A1 - Cecconello, M. A1 - Challis, C. A1 - Chapman, S. A1 - Chorley, J. A1 - Colyer, G. A1 - Conway, N. A1 - Cooper, W. A. A1 - Cox, M. A1 - Crocker, N. A1 - Crowley, B. A1 - Cunningham, G. A1 - Danilov, A. A1 - Darrow, D. A1 - Dendy, R. A1 - Diallo, A. A1 - Dickinson, D. A1 - Diem, S. A1 - Dorland, W. A1 - Dudson, B. A1 - Dunai, D. A1 - Easy, L. A1 - Elmore, S. A1 - Field, A. A1 - Fishpool, G. A1 - Fox, M. A1 - Fredrickson, E. A1 - Freethy, S. A1 - Garzotti, L. A1 - Ghim, Y. C. A1 - Gibson, K. A1 - Graves, J. A1 - Gurl, C. A1 - Guttenfelder, W. A1 - Ham, C. A1 - Harrison, J. A1 - Harting, D. A1 - Havlickova, E. A1 - Hawkes, N. A1 - Hender, T. A1 - Henderson, S. A1 - Highcock, E. A1 - Hillesheim, J. A1 - Hnat, B. A1 - Holgate, J. A1 - Horacek, J. A1 - Howard, J. A1 - Huang, B. A1 - Imada, K. A1 - Jones, O. A1 - S. Kaye A1 - Keeling, D. A1 - Kirk, A. A1 - Klimek, I. A1 - Kocan, M. A1 - Leggate, H. A1 - Lilley, M. A1 - Lipschultz, B. A1 - Lisgo, S. A1 - Liu, Y. Q. A1 - Lloyd, B. A1 - Lomanowski, B. A1 - Lupelli, I. A1 - Maddison, G. A1 - J. Mailloux A1 - Martin, R. A1 - McArdle, G. A1 - McClements, K. A1 - McMillan, B. A1 - Meakins, A. A1 - Meyer, H. A1 - Michael, C. A1 - Militello, F. A1 - Milnes, J. A1 - Morris, A. W. A1 - Motojima, G. A1 - Muir, D. A1 - Nardon, E. A1 - Naulin, V. A1 - Naylor, G. A1 - Nielsen, A. A1 - O'Brien, M. A1 - O'Gorman, T. A1 - Ono, Y. A1 - Oliver, H. A1 - Pamela, S. A1 - Pangioni, L. A1 - Parra, F. A1 - Patel, A. A1 - Peebles, W. A1 - Peng, M. A1 - Perez, R. A1 - Pinches, S. A1 - Piron, L. A1 - Podesta, M. A1 - Price, M. A1 - Reinke, M. A1 - Ren, Y. A1 - Roach, C. A1 - Robinson, J. A1 - Romanelli, M. A1 - Rozhansky, V. A1 - Saarelma, S. A1 - Sangaroon, S. A1 - Saveliev, A. A1 - Scannell, R. A1 - Schekochihin, A. A1 - Sharapov, S. A1 - Sharples, R. A1 - Shevchenko, V. A1 - Silburn, S. A1 - J. Simpson A1 - Storrs, J. A1 - Takase, Y. A1 - Tanabe, H. A1 - Tanaka, H. A1 - Taylor, D. A1 - Taylor, G. A1 - Thomas, D. A1 - Thomas-Davies, N. A1 - Thornton, A. A1 - Turnyanskiy, M. A1 - Valovic, M. A1 - Vann, R. A1 - Walkden, N. A1 - Wilson, H. A1 - Wyk, L. V. A1 - Yamada, T. A1 - Zoletnik, S. A1 - MAST Team A1 - MAST Upgrade Teams VL - 55 IS - 10 U1 - FP U2 - TP U5 - 9d7b191e90422e8ed8bcf2078b75987f ER - TY - JOUR T1 - Optimization of tungsten castellated structures for the ITER divertor JF - Journal of Nuclear Materials Y1 - 2015 A1 - Litnovsky, A. A1 - Hellwig, M. A1 - Matveev, D. A1 - Komm, M. A1 - van den Berg, M. A1 - De Temmerman, G. A1 - Rudakov, D. A1 - Ding, F. A1 - Luo, G. N. A1 - Krieger, K. A1 - Sugiyama, K. A1 - Pitts, R.A. A1 - Petersson, P. AB - In ITER, the plasma-facing components (PFCs) of the first wall and the divertor armor will be castellated to improve their thermo-mechanical stability and to limit forces due to induced currents. The fuel accumulation in the gaps may significantly contribute to the in-vessel fuel inventory. Castellation shaping may be the most straightforward way to minimize the fuel inventory and to alleviate the thermal loads onto castellations. A new castellation shape was proposed and comparative modeling of conventional (rectangular) and shaped castellation was performed for ITER conditions. Shaped castellation was predicted to be capable to operate under stationary heat load of 20 MW/m2. An 11-fold decrease of beryllium (Be) content in the gaps of the shaped cells alone with a 7-fold decrease of carbon content was predicted. In order to validate the predictive capabilities of modeling tools used for ITER conditions, the dedicated modeling with the same codes was made for existing tokamaks and benchmarked with the results of multi-machine experiments. For the castellations exposed in TEXTOR and DIII-D, the carbon amount in the gaps of shaped cells was 1.9–2.3 times smaller than that of rectangular ones. Modeling for TEXTOR conditions yielded to 1.5-fold decrease of carbon content in the gaps of shaped castellation outlining fair agreement with the experiment. At the same time, a number of processes, like enhanced erosion of molten layer yet need to be implemented in the codes in order to increase the accuracy of predictions for ITER. VL - 463 IS - Aug N1 - PLASMA-SURFACE INTERACTIONS 21 Proceedings of the 21st International Conference on Plasma-Surface Interactions in Controlled Fusion Devices Kanazawa, Japan May 26-30, 2014 U1 - PSI U2 - PSI-E U5 - 5720a3f9043818a3e5cc189dd1e70f73 ER - TY - JOUR T1 - The occurrence and damage of unipolar arcing on fuzzy tungsten JF - Journal of Nuclear Materials Y1 - 2015 A1 - Aussems, D. U. B. A1 - Nishijima, D. A1 - Brandt, C. A1 - van der Meiden, H. J. A1 - M. Vilémová A1 - J. Matějíček A1 - De Temmerman, G. A1 - Doerner, R. P. A1 - N.J. Lopes Cardozo AB - Abstract This research investigated whether unipolar arcing in the divertor of fusion reactors is a potential cause for enhanced wear of the divertor. It was found that 1 μm of nano-fuzz growth is sufficient to initiate arcing, mainly depending on the sheath potential drop and electron density. The average mass loss rate induced by the arc was determined from mass loss measurements and found to be consistent with the value estimated from the arc current. The average arc track erosion depth was estimated by using the measured mass loss and damaged surface area and was found to be one tenth of the fuzzy layer thickness. Due to melting of the fuzzy structures the actual depth is larger and some arc tracks occasionally appeared to even reach the bulk beyond the fuzzy layer. The conclusion of this study is therefore that arcing in the divertor of future tokamaks (e.g. ITER) potentially is an important cause for surface damage and plasma pollution. VL - 463 IS - Aug N1 - PLASMA-SURFACE İNTERACTIONS\} 21 Proceedings of the 21st International Conference on Plasma-Surface Interactions in Controlled Fusion Devices Kanazawa, Japan May 26-30, 2014 U1 - PSI U2 - PSI-E U5 - c3b2d2fa98a1df5755e7feedf7ed818c ER - TY - THES T1 - Optical boundary reconstruction for shape control of tokamak plasmas Y1 - 2014 A1 - Hommen, G. PB - Eindhoven University of Technology CY - Eindhoven, Netherlands VL - PhD SN - 9789088919251 UR - http://repository.tue.nl/780045 U1 - FP U2 - TP U5 - 3523557668f69380bf4503d0774a71a2 ER - TY - JOUR T1 - Operational characteristics of the high flux plasma generator Magnum-PSI JF - Fusion Engineering and Design Y1 - 2014 A1 - van Eck, H. J. N. A1 - Abrams, T. A1 - van den Berg, M. A. A1 - Brons, S. A1 - van Eden, G. G. A1 - Jaworski, M. A. A1 - Kaita, R. A1 - van der Meiden, H. J. A1 - Morgan, T. W. A1 - van de Pol, M.J. A1 - Scholten, J. A1 - Smeets, P. H. M. A1 - De Temmerman, G. A1 - de Vries, P. C. A1 - Zeijlmans van Emmichoven, P. A. KW - ELMs KW - ITER KW - Linear plasma device KW - Lithium coatings KW - Plasma–surface interactions AB - Abstract In Magnum-PSI (MAgnetized plasma Generator and \{NUMerical\} modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma surface interactions in the regime relevant for fusion reactors such as ITER: 1023-1025 m−2 s−1 hydrogen plasma flux densities at 1-5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented. VL - 89 IS - 9-10 U1 - PSI U2 - PSI-FI U5 - e5796d3bcf8a774844724716fd74aaad ER - TY - JOUR T1 - Operational status of the Magnum-PSI linear plasma device JF - Fusion Engineering and Design Y1 - 2013 A1 - Scholten, J. A1 - Zeijlmans van Emmichoven, P. A. A1 - van Eck, H. J. N. A1 - Smeets, P. H. M. A1 - De Temmerman, G. C. A1 - Brons, S. A1 - van den Berg, M. A. A1 - van der Meiden, H. J. A1 - van de Pol, M.J. A1 - M. F. Graswinckel A1 - Groen, P. W. C. A1 - Poelman, A. J. A1 - Genuit, J. W. KW - ITER KW - Magnum-PSI KW - plasma generator KW - Plasma surface interactions AB -The construction phase of the linear plasma generator Magnum-PSI at the FOM institute DIFFER has been completed and the facility has been officially opened in March 2012. The scientific program to gain more insight in the plasma–wall interactions relevant for ITER and future fusion reactors has started. In Magnum-PSI, targets of a wide range of materials and shapes can be exposed to high particle, high heat flux plasmas (>1024 ions m−2 s−1; >10 MW/m2). For magnetization of the plasma, oil-cooled electromagnets are temporarily installed to enable pulsed operation until the device is upgraded with a superconducting magnet. The magnets generate a field of up to 1.9 T close to the plasma source for a duration of 6 s. Longer exposure times are available for lower field settings. Plasma characterizations were done with a variety of gases (H, D, He, Ne and Ar) to determine the machine performance and prepare for subsequent scientific experiments. Thomson scattering and optical emission spectroscopy were used to determine the plasma parameters while infrared thermography and target calorimetry were used to determine the power loads to the surface. This paper reports on the status of Magnum-PSI and its diagnostic systems. In addition, an overview of the plasma parameters that can be achieved in the present state will be given.

VL - 88 N1 -PSI

U2 -PSI-FI

U5 - ec624221f1f1c3ac29d2d8860a12da09 ER - TY - JOUR T1 - Optimizing the current ramp-up phase for the hybrid ITER scenario JF - Nuclear Fusion Y1 - 2013 A1 - Hogeweij, G. M. D. A1 - Artaud, J. F. A1 - Casper, T. A. A1 - Citrin, J. A1 - Imbeaux, F. A1 - Kochl, F. A1 - X. Litaudon A1 - Voitsekhovitch, I. A1 - ITM-TF ITER Scenario Modelling Group AB -The current ramp-up phase for the ITER hybrid scenario is analysed with the CRONOS integrated modelling suite. The simulations presented in this paper show that the heating systems available at ITER allow, within the operational limits, the attainment of a hybrid q profile at the end of the current ramp-up. A reference ramp-up scenario is reached by a combination of NBI, ECCD (UPL) and LHCD. A heating scheme with only NBI and ECCD can also reach the target q profile; however, LHCD can play a crucial role in reducing the flux consumption during the ramp-up phase. The optimum heating scheme depends on the chosen transport model, and on assumptions of parameters like n e peaking, edge T e,i and Z eff . The sensitivity of the current diffusion on parameters that are not easily controlled, shows that development of real-time control is important to reach the target q profile. A first step in that direction has been indicated in this paper. Minimizing resistive flux consumption and optimizing the q profile turn out to be conflicting requirements. A trade-off between these two requirements has to be made. In this paper it is shown that fast current ramp with L-mode current overshoot is at the one extreme, i.e. the optimum q profile at the cost of increased resistive flux consumption, whereas early H-mode transition is at the other extreme.

VL - 53 U1 -FP

U2 -CPP-HT

U5 - 88a9a6c3198d406bdfbbe7a0eea1a68d ER - TY - JOUR T1 - Overview of ASDEX Upgrade results JF - Nuclear Fusion Y1 - 2013 A1 - Stroth, U. A1 - Adamek, J. A1 - Aho-Mantila, L. A1 - Akaslompolo, S. A1 - Amdor, C. A1 - Angioni, C. A1 - Balden, M. A1 - Bardin, S. A1 - L. Barrera Orte A1 - Behler, K. A1 - Belonohy, E. A1 - Bergmann, A. A1 - Bernert, M. A1 - Bilato, R. A1 - Birkenmeier, G. A1 - Bobkov, V. A1 - Boom, J. A1 - Bottereau, C. A1 - Bottino, A. A1 - Braun, F. A1 - Brezinsek, S. A1 - Brochard, T. A1 - M. Brüdgam A1 - Buhler, A. A1 - Burckhart, A. A1 - Casson, F. J. A1 - Chankin, A. A1 - Chapman, I. A1 - Clairet, F. A1 - Classen, I.G.J. A1 - Coenen, J. W. A1 - Conway, G. D. A1 - Coster, D. P. A1 - Curran, D. A1 - da Silva, F. A1 - P. de Marné A1 - D'Inca, R. A1 - Douai, D. A1 - Drube, R. A1 - Dunne, M. A1 - Dux, R. A1 - Eich, T. A1 - Eixenberger, H. A1 - Endstrasser, N. A1 - Engelhardt, K. A1 - Esposito, B. A1 - Fable, E. A1 - Fischer, R. A1 - H. Fünfgelder A1 - Fuchs, J. C. A1 - K. Gál A1 - M. García Muñoz A1 - Geiger, B. A1 - Giannone, L. A1 - T. Görler A1 - da Graca, S. A1 - Greuner, H. A1 - Gruber, O. A1 - Gude, A. A1 - Guimarais, L. A1 - S. Günter A1 - Haas, G. A1 - Hakola, A. H. A1 - Hangan, D. A1 - Happel, T. A1 - T. Härtl A1 - Hauff, T. A1 - Heinemann, B. A1 - Herrmann, A. A1 - Hobirk, J. A1 - H. Höhnle A1 - M. Hölzl A1 - Hopf, C. A1 - Houben, A. A1 - Igochine, V. A1 - Ionita, C. A1 - Janzer, A. A1 - Jenko, F. A1 - Kantor, M. A1 - C.-P. Käsemann A1 - Kallenbach, A. A1 - S. Kálvin A1 - Kantor, M. A1 - Kappatou, A. A1 - Kardaun, O. A1 - Kasparek, W. A1 - Kaufmann, M. A1 - Kirk, A. A1 - H.-J. Klingshirn A1 - Kocan, M. A1 - Kocsis, G. A1 - Konz, C. A1 - Koslowski, R. A1 - Krieger, K. A1 - Kubic, M. A1 - Kurki-Suonio, T. A1 - Kurzan, B. A1 - Lackner, K. A1 - Lang, P. T. A1 - Lauber, P. A1 - Laux, M. A1 - Lazaros, A. A1 - Leipold, F. A1 - Leuterer, F. A1 - Lindig, S. A1 - Lisgo, S. A1 - Lohs, A. A1 - Lunt, T. A1 - Maier, H. A1 - Makkonen, T. A1 - Mank, K. A1 - M.-E. Manso A1 - Maraschek, M. A1 - Mayer, M. A1 - McCarthy, P. J. A1 - McDermott, R. A1 - Mehlmann, F. A1 - Meister, H. A1 - Menchero, L. A1 - Meo, F. A1 - Merkel, P. A1 - Merkel, R. A1 - Mertens, V. A1 - Merz, F. A1 - Mlynek, A. A1 - Monaco, F. A1 - Müller, S. A1 - H.W. Müller A1 - M. Münich A1 - Neu, G. A1 - Neu, R. A1 - Neuwirth, D. A1 - Nocente, M. A1 - Nold, B. A1 - Noterdaeme, J. M. A1 - Pautasso, G. A1 - Pereverzev, G. A1 - B. Plöckl A1 - Podoba, Y. A1 - Pompon, F. A1 - Poli, E. A1 - Polozhiy, K. A1 - Potzel, S. A1 - Puschel, M. J. A1 - Putterich, T. A1 - Rathgeber, S. K. A1 - Raupp, G. A1 - Reich, M. A1 - Reimold, F. A1 - Ribeiro, T. A1 - Riedl, R. A1 - Rohde, V. A1 - van Rooij, G. J. A1 - Roth, J. A1 - Rott, M. A1 - Ryter, F. A1 - Salewski, M. A1 - Santos, J. A1 - Sauter, P. A1 - Scarabosio, A. A1 - Schall, G. A1 - Schmid, K. A1 - Schneider, P. A. A1 - Schneider, W. A1 - Schrittwieser, R. A1 - Schubert, M. A1 - Schweinzer, J. A1 - Scott, B. A1 - Sempf, M. A1 - Sertoli, M. A1 - Siccinio, M. A1 - Sieglin, B. A1 - Sigalov, A. A1 - Silva, A. A1 - Sommer, F. A1 - A. Stäbler A1 - Stober, J. A1 - Streibl, B. A1 - Strumberger, E. A1 - Sugiyama, K. A1 - Suttrop, W. A1 - Tala, T. A1 - Tardini, G. A1 - Teschke, M. A1 - Tichmann, C. A1 - Told, D. A1 - Treutterer, W. A1 - Tsalas, M. A1 - VanZeeland, M. A. A1 - Varela, P. A1 - Veres, G. A1 - Vicente, J. A1 - Vianello, N. A1 - Vierle, T. A1 - Viezzer, E. A1 - Viola, B. A1 - Vorpahl, C. A1 - Wachowski, M. A1 - Wagner, D. A1 - Wauters, T. A1 - Weller, A. A1 - Wenninger, R. A1 - Wieland, B. A1 - Willensdorfer, M. A1 - Wischmeier, M. A1 - Wolfrum, E. A1 - E. Würsching A1 - Yu, Q. A1 - Zammuto, I. A1 - Zasche, D. A1 - Zehetbauer, T. A1 - Zhang, Y. A1 - Zilker, M. A1 - Zohm, H. AB - The medium size divertor tokamak ASDEX Upgrade (major and minor radii 1.65 m and 0.5 m, respectively, magnetic-field strength 2.5 T) possesses flexible shaping and versatile heating and current drive systems. Recently the technical capabilities were extended by increasing the electron cyclotron resonance heating (ECRH) power, by installing 2 × 8 internal magnetic perturbation coils, and by improving the ion cyclotron range of frequency compatibility with the tungsten wall. With the perturbation coils, reliable suppression of large type-I edge localized modes (ELMs) could be demonstrated in a wide operational window, which opens up above a critical plasma pedestal density. The pellet fuelling efficiency was observed to increase which gives access to H-mode discharges with peaked density profiles at line densities clearly exceeding the empirical Greenwald limit. Owing to the increased ECRH power of 4 MW, H-mode discharges could be studied in regimes with dominant electron heating and low plasma rotation velocities, i.e. under conditions particularly relevant for ITER. The ion-pressure gradient and the neoclassical radial electric field emerge as key parameters for the transition. Using the total simultaneously available heating power of 23 MW, high performance discharges have been carried out where feed-back controlled radiative cooling in the core and the divertor allowed the divertor peak power loads to be maintained below 5 MW m −2 . Under attached divertor conditions, a multi-device scaling expression for the power-decay length was obtained which is independent of major radius and decreases with magnetic field resulting in a decay length of 1 mm for ITER. At higher densities and under partially detached conditions, however, a broadening of the decay length is observed. In discharges with density ramps up to the density limit, the divertor plasma shows a complex behaviour with a localized high-density region in the inner divertor before the outer divertor detaches. Turbulent transport is studied in the core and the scrape-off layer (SOL). Discharges over a wide parameter range exhibit a close link between core momentum and density transport. Consistent with gyro-kinetic calculations, the density gradient at half plasma radius determines the momentum transport through residual stress and thus the central toroidal rotation. In the SOL a close comparison of probe data with a gyro-fluid code showed excellent agreement and points to the dominance of drift waves. Intermittent structures from ELMs and from turbulence are shown to have high ion temperatures even at large distances outside the separatrix. VL - 53 UR - http://hdl.handle.net/11858/00-001M-0000-0026-E166-7 IS - 10 U1 - FP U2 - PDG U5 - 0b5b08fdc590c85cc01e6d1db1958848 ER - TY - JOUR T1 - Observations of orientation dependence of surface morphology in tungsten implanted by low energy and high flux D plasma JF - Journal of Nuclear Materials Y1 - 2013 A1 - Xu, H.Y. A1 - Zhang, Y. B. A1 - Yuan, Y. A1 - Fu, B. Q. A1 - Godfrey, A. A1 - De Temmerman, G. A1 - Liu, W. A1 - Huang, X. AB - Surface modification by formation of blistering and nanostructures with pronounced orientation dependence has been observed on surfaces of rolled tungsten and recrystallized tungsten after exposure to a low energy (38 eV) deuterium (D) plasma with a high flux of 1024 m-2 s -1. The correlation between blisters and nanostructures with grain orientation was examined on recrystallized tungsten to exclude the influence of defects introduced during plastic deformation on the pattern of surface modification. The amount of blistering changed from the most in grains oriented close to 〈1 1 1〉 to the least in grains oriented close to 〈0 0 1〉. Three kinds of typical nanostructures were observed, with a clear dependence on grain orientation. Triangular structures were observed on grains oriented near the 〈1 1 1〉 corner of the inverse pole figure, with lamellar structures formed for grains oriented near the 〈0 1 1〉 corner, and spongy structures for grains near the 〈0 0 1〉 corner. Possible reasons for the orientation dependence of both the blisters and nanostructures are discussed. © 2013 IOP Publishing Ltd. VL - 443 U1 - PSI U2 - PSI-E U5 - adedb936c6505bbff4eb22f589d6eda8 ER - TY - JOUR T1 - Observations on the W-transport in the core plasma of JET and ASDEX Upgrade JF - Plasma Physics and Controlled Fusion Y1 - 2013 A1 - Putterich, T. A1 - Dux, R. A1 - Neu, R. A1 - Bernert, M. A1 - Beurskens, M. N. A. A1 - Bobkov, V. A1 - Brezinsek, S. A1 - Challis, C. A1 - Coenen, J. W. A1 - Coffey, I. A1 - Czarnecka, A. A1 - Giroud, C. A1 - Jacquet, P. A1 - Joffrin, E. A1 - Kallenbach, A. A1 - Lehnen, M. A1 - Lerche, E. A1 - de la Luna, E. A1 - Marsen, S. A1 - Matthews, G. A1 - M-L Mayoral A1 - McDermott, R. M. A1 - Meigs, A. A1 - Mlynar, J. A1 - Sertoli, M. A1 - van Rooij, G. J. A1 - ASDEX-Upgrade Team A1 - JET-EFDA Contributors AB - The W-transport in the core plasma of JET is investigated experimentally by deriving the W-concentration profiles from the modelling of the signals of the soft x-ray cameras. For the case of pure neutral beam heating W accumulates in the core ( r / a < 0.3) approaching W-concentrations of 10 −3 in between the sawtooth crashes, which flatten the W-profile to a concentration of about 3 × 10 −5 . When central Ion cyclotron resonant heating is additionally applied the core W-concentration decays in phases that exhibit a changed mode activity, while also the electron temperature increases and the density profile becomes less peaked. The immediate correlation between the change of magnetohydrodymanic (MHD) and the removal of W from the plasma core supports the hypothesis that the change of the MHD activity is the underlying cause for the change of transport. Furthermore, a discharge from the ASDEX Upgrade is investigated. In this case the plasma profiles exhibit small changes only, while the most prominent change occurs in the W-content of the confined plasma caused by the reduction of the fuelling deuterium gas puff. Concomintantly, the W-concentration profiles in the core plasma r / a < 0.2 steepen up reminescent to the well-known connection between central radiation and transport during cases with strong, established W-accumulation, while in the present analysis such a causality between the two during the onset of W-accumulation could not be pinned down. Both case studies exemplify that small changes of the core parameters of a plasma my influence the W-transport in the plasma core drastically. VL - 55 UR - http://stacks.iop.org/0741-3335/55/i=12/a=124036 U1 - PSI U2 - PSI-E U5 - 2b2c6ee28e9368baf3d04cebb61bb5b5 ER - TY - JOUR T1 - Outline of optical design and viewing geometry for divertor Thomson scattering on MAST upgrade JF - Journal of Instrumentation Y1 - 2013 A1 - Hawke, J. A1 - Scannell, R. A1 - Harrison, J. A1 - Huxford, R. A1 - Bohm, P. AB -The super-X divertor on MAST Upgrade will be diagnosed by a Thomson scattering diagnostic. A preliminary design of the collection optics and calculations of the diagnostic's performance are discussed in this paper. As part of the design the location and size of the collection cell were optimized to minimize vignetting, especially in the region of interest close to the divertor strike point. The design process was complicated by the limited access available in the closed divertor geometry. In the study of the diagnostic's performance, the radial resolution, projection of the laser image onto the fiber bundle, and impact of depth of field with a multiple laser system were investigated. In this design there is a trade-off between the resolution of the system and the lifetime of the beam dump. For this reason the beam has its focal point at the start of the viewing region and diverges in width to approximately five millimeters near the divertor tile. The effect of this large variation in beam width is examined primarily at the two extremes by means of ray trace modeling. This model takes an object with dimensions of the beam width imaged onto the fiber bundle to investigate the effect of misalignment for a narrow or broad laser image. In a similar manner ray tracing was performed to determine the effects of depth of field for four and two laser systems. As the electron density of the system may be low, performance analysis considers firing multiple lasers simultaneously to improve photon statistics.

VL - 8 IS - 11 U1 -FP

U2 -PDG

U5 - bc432db4e4cd856d6e7b3a556b348d42 ER - TY - JOUR T1 - Overview of physics results from MAST towards ITER/DEMO and the MAST Upgrade JF - Nuclear Fusion Y1 - 2013 A1 - Meyer, H. A1 - Abel, I. G. A1 - Akers, R. J. A1 - Allan, A. A1 - Allan, S. Y. A1 - Appel, L. C. A1 - Asunta, O. A1 - Barnes, M. A1 - Barratt, N. C. A1 - N. Ben Ayed A1 - Bradley, J. W. A1 - Canik, J. A1 - Cahyna, P. A1 - Cecconello, M. A1 - Challis, C. D. A1 - Chapman, I.T. A1 - Ciric, D. A1 - Colyer, G. A1 - Conway, N. J. A1 - Cox, M. A1 - Crowley, B. J. A1 - Cowley, S. C. A1 - Cunningham, G. A1 - Danilov, A. A1 - Darke, A. A1 - de Bock, M. F. M. A1 - De Temmerman, G. A1 - Dendy, R. O. A1 - Denner, P. A1 - Dickinson, D. A1 - Dnestrovskij, A. Y. A1 - Dnestrovsky, Y. A1 - Driscoll, M. D. A1 - Dudson, B. A1 - Dunai, D. A1 - Dunstan, M. A1 - Dura, P. A1 - Elmore, S. A1 - Field, A. R. A1 - Fishpool, G. A1 - Freethy, S. A1 - Fundamenski, W. A1 - Garzotti, L. A1 - Ghim, Y. C. A1 - Gibson, K. J. A1 - Gryaznevich, M. P. A1 - Harrison, J. A1 - E. Havlíčková A1 - Hawkes, N. C. A1 - Heidbrink, W. W. A1 - Hender, T. C. A1 - Highcock, E. A1 - Higgins, D. A1 - Hill, P. A1 - Hnat, B. A1 - Hole, M. J. A1 - J. Horáček A1 - Howell, D. F. A1 - Imada, K. A1 - Jones, O. A1 - Kaveeva, E. A1 - Keeling, D. A1 - Kirk, A. A1 - M. Kočan A1 - Lake, R. J. A1 - Lehnen, M. A1 - Leggate, H. J. A1 - Liang, Y. A1 - Lilley, M. K. A1 - Lisgo, S. W. A1 - Liu, Y. Q. A1 - Lloyd, B. A1 - G. P. Maddison A1 - J. Mailloux A1 - Martin, R. A1 - McArdle, G. J. A1 - McClements, K. G. A1 - McMillan, B. A1 - Michael, C. A1 - Militello, F. A1 - Molchanov, P. A1 - Mordijck, S. A1 - Morgan, T. A1 - Morris, A. W. A1 - Muir, D. G. A1 - Nardon, E. A1 - Naulin, V. A1 - Naylor, G. A1 - Nielsen, A. H. A1 - O'Brien, M. R. A1 - O'Gorman, T. A1 - Pamela, S. A1 - Parra, F. I. A1 - Patel, A. A1 - Pinches, S. D. A1 - Price, M. N. A1 - Roach, C. M. A1 - Robinson, J. R. A1 - Romanelli, M. A1 - Rozhansky, V. A1 - Saarelma, S. A1 - Sangaroon, S. A1 - Saveliev, A. A1 - Scannell, R. A1 - Seidl, J. A1 - Sharapov, S. E. A1 - Schekochihin, A. A. A1 - Shevchenko, V. A1 - Shibaev, S. A1 - Stork, D. A1 - Storrs, J. A1 - Sykes, A. A1 - Tallents, G. J. A1 - Tamain, P. A1 - Taylor, D. A1 - Temple, D. A1 - Thomas-Davies, N. A1 - Thornton, A. A1 - Turnyanskiy, M. R. A1 - M. Valovič A1 - Vann, R. G. L. A1 - Verwichte, E. A1 - Voskoboynikov, P. A1 - Voss, G. A1 - Warder, S. E. V. A1 - Wilson, H. R. A1 - Wodniak, I. A1 - Zoletnik, S. A1 - Zagorski, R. A1 - MAST Team A1 - NBI Team AB - New diagnostic, modelling and plant capability on the Mega Ampère Spherical Tokamak (MAST) have delivered important results in key areas for ITER/DEMO and the upcoming MAST Upgrade, a step towards future ST devices on the path to fusion currently under procurement. Micro-stability analysis of the pedestal highlights the potential roles of micro-tearing modes and kinetic ballooning modes for the pedestal formation. Mitigation of edge localized modes (ELM) using resonant magnetic perturbation has been demonstrated for toroidal mode numbers n = 3, 4, 6 with an ELM frequency increase by up to a factor of 9, compatible with pellet fuelling. The peak heat flux of mitigated and natural ELMs follows the same linear trend with ELM energy loss and the first ELM-resolved T i measurements in the divertor region are shown. Measurements of flow shear and turbulence dynamics during L–H transitions show filaments erupting from the plasma edge whilst the full flow shear is still present. Off-axis neutral beam injection helps to strongly reduce the redistribution of fast-ions due to fishbone modes when compared to on-axis injection. Low- k ion-scale turbulence has been measured in L-mode and compared to global gyro-kinetic simulations. A statistical analysis of principal turbulence time scales shows them to be of comparable magnitude and reasonably correlated with turbulence decorrelation time. T e inside the island of a neoclassical tearing mode allow the analysis of the island evolution without assuming specific models for the heat flux. Other results include the discrepancy of the current profile evolution during the current ramp-up with solutions of the poloidal field diffusion equation, studies of the anomalous Doppler resonance compressional Alfvén eigenmodes, disruption mitigation studies and modelling of the new divertor design for MAST Upgrade. The novel 3D electron Bernstein synthetic imaging shows promising first data sensitive to the edge current profile and flows. VL - 53 UR - http://stacks.iop.org/0029-5515/53/i=10/a=104008 U1 - PSI U2 - PSI-E U5 - fee6f536ea06c1003255446f71a039bd ER - TY - JOUR T1 - Overview on plasma operation with a full tungsten wall in ASDEX Upgrade JF - Journal of Nuclear Materials Y1 - 2013 A1 - Neu, R. A1 - Kallenbach, A. A1 - Balden, M. A1 - Bobkov, V. A1 - Coenen, J. W. A1 - Drube, R. A1 - Dux, R. A1 - Greuner, H. A1 - Herrmann, A. A1 - Hobirk, J. A1 - H. Höhnle A1 - Krieger, K. A1 - M. Kočan A1 - Lang, P. A1 - Lunt, T. A1 - Maier, H. A1 - Mayer, M. A1 - H.W. Müller A1 - Potzel, S. A1 - Putterich, T. A1 - Rapp, J. A1 - Rohde, V. A1 - Ryter, F. A1 - Schneider, P. A. A1 - Schweinzer, J. A1 - Sertoli, M. A1 - Stober, J. A1 - Suttrop, W. A1 - Sugiyama, K. A1 - van Rooij, G. J. A1 - Wischmeier, M. AB - Abstract Operation with all tungsten plasma facing components has become routine in ASDEX Upgrade. The conditioning of the device is strongly simplified and short glow discharges are used only on a daily basis. The long term fuel retention was reduced by more than a factor of 5 as demonstrated in gas balance as well as in post mortem analyses. Injecting nitrogen for radiative cooling, discharges with additional heating power up to 23 MW have been achieved, providing good confinement (H98y2=1), divertor power loads around 5 MW m−2 and divertor temperatures below 10 eV. ELM mitigation by pellet ELM pacemaking or magnetic perturbation coils reduces the deposited energy during ELMs, but also keeps the W density at the pedestal low. As a recipe to keep the central W concentration sufficiently low, central (wave) heating is well established and low density H-Modes could be re-established with the newly available ECRH power of up to 4 MW. The ICRH induced W sources could be strongly reduced by applying boron coatings to the poloidal guard limiters. VL - 438, Supplement UR - http://www.sciencedirect.com/science/article/pii/S0022311513000147 N1 -Experiments in JET with carbon-based plasma-facing components have been carried out in preparation of the ITER-Like Wall with beryllium main chamber and full tungsten divertor. The preparatory work was twofold: (i) development of techniques, which ensure safe operation with the new wall and (ii) provision of reference plasmas, which allow a comparison of operation with carbon and metallic wall. (i) Compatibility with the W divertor with respect to energy loads could be achieved in N-2 seeded plasmas at high densities and low temperatures, finally approaching partial detachment, with only moderate confinement reduction of 10%. Strike-point sweeping increases the operational space further by re-distributing the load over several components. (ii) Be and C migration to the divertor has been documented with spectroscopy and QMBs under different plasma conditions providing a database which will allow a comparison of the material transport to remote areas with metallic walls. Fuel retention rates of 1.0-2.0 x 10(21) D s(-1) were obtained as references in accompanied gas balance studies. (C) 2010 Published by Elsevier B.V.

VL - 415 SN - 0022-3115 IS - 1 N1 - ISI Document Delivery No.: 862XTTimes Cited: 8Cited Reference Count: 1719th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI)MAY 24-28, 2010San Diego, CALawrence Livermore Natl LabS U1 -FP

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U5 - d47299eebfd541ed3d2532c0cfcec12f ER - TY - JOUR T1 - Overview of the second stage in the comprehensive mirrors test in JET JF - Physica Scripta Y1 - 2011 A1 - Rubel, M. A1 - Ivanova, D. A1 - Coad, J. P. A1 - De Temmerman, G. A1 - Likonen, J. A1 - Marot, L. A1 - Schmidt, A. A1 - Widdowson, A. KW - DEPOSITION KW - EROSION/DEPOSITION KW - ITER KW - RHODIUM FILMS AB -The first mirror test for ITER in JET with carbon walls has been completed. Thirty polycrystalline Mo mirrors including four coated with a 1 mu m rhodium (Rh) film were exposed to plasma in the divertor region and in the main chamber. The mirrors were installed in eight cassettes of pan-pipe shape. The reflectivity of all mirrors exposed in the divertor has been degraded by 80-90% because of the formation of thick (> 20 mu m) flaking co-deposits on surfaces. Only small reflectivity losses (5-10%) occurred on mirrors located at the channel mouth of the cassettes from the main chamber wall. This is due to the in situ removal of deposited species by charge exchange neutrals. Deuterium, (12)C and (9)Be are the main isotopes detected on surfaces, but other isotopes ((13)C) are also found in some locations, thus indicating differences in the material migration. Rhodium coatings with an initial reflectivity that is 30% better than that of pure Mo survived the test without detachment, but their resultant reflectivity was the same as that of the exposed Mo surfaces.

VL - T145 SN - 0031-8949 UR - http://www.euro-fusionscipub.org/wp-content/uploads/2014/11/EFDC110105.pdf N1 - ISI Document Delivery No.: 867SUTimes Cited: 0Cited Reference Count: 2113th International Workshop on Plasma-Facing Materials and Components for Fusion Applications (PFMC)/1st International Conference on Fusion Energy Materials Science (FEMaS)MAY 09-13, 2011Rosenheim, GERMANYIPP, European Commiss U1 -PSI

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U5 - 6de565f35f3c1ec8ca73331ffe013714 ER - TY - JOUR T1 - Optimizing ion-cyclotron resonance frequency heating for ITER: dedicated JET experiments JF - Plasma Physics and Controlled Fusion Y1 - 2011 A1 - Lerche, E. A1 - Van Eester, D. A1 - Ongena, J. A1 - Mayoral, M. L. A1 - Laxaback, M. A1 - Rimini, F. A1 - Argouarch, A. A1 - Beaumont, P. A1 - Blackman, T. A1 - Bobkov, V. A1 - Brennan, D. A1 - Brett, A. A1 - Calabro, G. A1 - Cecconello, M. A1 - Coffey, I. A1 - Colas, L. A1 - Coyne, A. A1 - Crombe, K. A1 - Czarnecka, A. A1 - Dumont, R. A1 - Durodie, F. A1 - Felton, R. A1 - Frigione, D. A1 - Johnson, M. G. A1 - Giroud, C. A1 - Gorini, G. A1 - Graham, M. A1 - Hellesen, C. A1 - Hellsten, T. A1 - Huygen, S. A1 - Jacquet, P. A1 - Johnson, T. A1 - Kiptily, V. A1 - Knipe, S. A1 - Krasilnikov, A. A1 - Lamalle, P. A1 - Lennholm, M. A1 - Loarte, A. A1 - Maggiora, R. A1 - Maslov, M. A1 - Messiaen, A. A1 - Milanesio, D. A1 - Monakhov, I. A1 - Nightingale, M. A1 - Noble, C. A1 - Nocente, M. A1 - Pangioni, L. A1 - Proverbio, I. A1 - Sozzi, C. A1 - Stamp, M. A1 - Studholme, W. A1 - Tardocchi, M. A1 - Versloot, T. W. A1 - Vdovin, V. A1 - Vrancken, M. A1 - Whitehurst, A. A1 - Wooldridge, E. A1 - Zoita, V. KW - DESIGN KW - ICRF ANTENNAS KW - MODE CONVERSION KW - PLASMAS KW - Sawtooth KW - SCENARIOS KW - SYSTEM KW - TOKAMAK AB -In the past years, one of the focal points of the JET experimental programme was on ion-cyclotron resonance heating (ICRH) studies in view of the design and exploitation of the ICRH system being developed for ITER. In this brief review, some of the main achievements obtained in JET in this field during the last 5 years will be summarized. The results reported here include important aspects of a more engineering nature, such as (i) the appropriate design of the RF feeding circuits for optimal load resilient operation and (ii) the test of a compact high-power density antenna array, as well as RF physics oriented studies aiming at refining the numerical models used for predicting the performance of the ICRH system in ITER. The latter include (i) experiments designed for improving the modelling of the antenna coupling resistance under various plasma conditions and (ii) the assessment of the heating performance of ICRH scenarios to be used in the non-active operation phase of ITER.

VL - 53 SN - 0741-3335 IS - 12 N1 - ISI Document Delivery No.: 870BLTimes Cited: 0Cited Reference Count: 43Part 1-2 U1 -FP

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U5 - 5271f643f9b6df31138d568a0bcdbc8b ER - TY - JOUR T1 - On the oxidation mechanism of microcrystalline silicon thin films studied by Fourier transform infrared spectroscopy JF - Journal of Non-Crystalline Solids Y1 - 2011 A1 - Bronneberg, A. C. A1 - Smets, A. H. M. A1 - Creatore, M. A1 - M. C. M. van de Sanden KW - ABSORPTION-BANDS KW - CRYSTALLINE KW - DEPOSITION KW - Fourier transform infrared KW - GROWTH KW - HYDROGENATED AMORPHOUS-SILICON KW - Microcrystalline silicon KW - oxidation KW - PLASMA KW - SI-H KW - SOLAR-CELLS KW - SPECTRA KW - SPECTROSCOPY KW - SURFACES AB - Insight into the oxidation mechanism of microcrystalline silicon thin films has been obtained by means of Fourier transform infrared spectroscopy. The films were deposited by using the expanding thermal plasma and their oxidation upon air exposure was followed in time. Transmission spectra were recorded directly after deposition and at regular intervals up to 8 months after deposition. The interpretation of the spectra is focused on the Si-H, stretching (2000-2100 cm(-1)), Si-O-Si (1000-1200 cm(-1)), and OxSi-Hy modes (2130-2250 cm(-1)).A short time scale (< 3 months) oxidation of the crystalline grain boundaries is observed, while at longer time scales, the oxidation of the amorphous tissue and the formation of O-H groups on the grain boundary surfaces play a role. The implications of this study on the quality of microcrystalline silicon exhibiting no post-deposition oxidation are discussed: it is not sufficient to merely passivate the surface of the crystalline grains and fill the gap between the grains with amorphous silicon. Instead, the quality of the amorphous silicon tissue should also be taken into account, since this oxidation can affect the passivating properties of the amorphous tissue on the surface of the crystalline silicon grains. (C) 2010 Elsevier B.V. All rights reserved. VL - 357 SN - 0022-3093 IS - 3 U1 - MaSF U2 - MaSF-E U5 - 46822b4c9ff718e0877361c1daabdb12 ER - TY - JOUR T1 - Objectives, physics requirements and conceptual design of an ECRH system for JET JF - Nuclear Fusion Y1 - 2011 A1 - Giruzzi, G. A1 - Lennholm, M. A1 - Parkin, A. A1 - Aiello, G. A1 - Bellinger, M. A1 - Bird, J. A1 - Bouquey, F. A1 - Braune, H. A1 - Bruschi, A. A1 - Butcher, P. A1 - Clay, R. A1 - de la Luna, E. A1 - Denisov, G. A1 - Edlington, T. A1 - Fanthome, J. A1 - Farina, D. A1 - Farthing, J. A1 - Figini, L. A1 - Garavaglia, S. A1 - Garcia, J. A1 - Gardener, M. A1 - Gerbaud, T. A1 - Granucci, G. A1 - Hay, J. A1 - Henderson, M. A1 - Hotchin, S. A1 - Ilyin, V. N. A1 - Jennison, M. A1 - Kasparek, W. A1 - Khilar, P. A1 - Kirneva, N. A1 - Kislov, D. A1 - Knipe, S. A1 - Kuyanov, A. A1 - X. Litaudon A1 - Litvak, A. G. A1 - Moro, A. A1 - Nowak, S. A1 - Parail, V. A1 - Plaum, B. A1 - Saibene, G. A1 - Sozzi, C. A1 - Späh, P. A1 - Strauss, D. A1 - Trukhina, E. A1 - Vaccaro, A. A1 - Vagdama, A. A1 - Vdovin, V. KW - CURRENT DRIVE KW - ITER KW - MODEL KW - PROGRESS KW - SAWTOOTH PERIOD AB -A study has been conducted to evaluate the feasibility of installing an electron cyclotron resonance heating (ECRH) and current drive system on the JET tokamak. The main functions of this system would be electron heating, sawtooth control, neoclassical tearing mode control to access high beta regimes and current profile control to access and maintain advanced plasma scenarios. This paper presents an overview of the studies performed in this framework by an EU-Russia project team. The motivations for this major upgrade of the JET heating systems and the required functions are discussed. The main results of the study are summarized. The usefulness of a 10 MW level EC system for JET is definitely confirmed by the physics studies. Neither feasibility issues nor strong limitations for any of the functions envisaged have been found. This has led to a preliminary conceptual design of the system.

VL - 51 SN - 0029-5515 IS - 6 N1 - ISI Document Delivery No.: 766MWTimes Cited: 2Cited Reference Count: 37 U1 -FP

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U5 - e451b1911c468bfd0a4f2e7f751d6c9f ER - TY - JOUR T1 - Overview on R&D and design activities for the ITER core charge exchange spectroscopy diagnostic system JF - Fusion Engineering and Design Y1 - 2011 A1 - Biel, W. A1 - Baross, T. A1 - Bourauel, P. A1 - Dunai, D. A1 - Durkut, M. A1 - Erdei, G. A1 - Hawkes, N. A1 - von Hellermann, M. A1 - Hogenbirk, A. A1 - R. Jaspers A1 - Kiss, G. A1 - Klinkhamer, F. A1 - Koning, J. F. A1 - Kotov, V. A1 - Krasikov, Y. A1 - Krimmer, A. A1 - Lischtschenko, O. A1 - Litnovsky, A. A1 - Marchuk, O. A1 - Neubauer, O. A1 - Offermanns, G. A1 - Panin, A. A1 - Patel, K. A1 - Pokol, G. A1 - Schrader, M. A1 - Snijders, B. A1 - Szabo, V. A1 - van der Valk, N. A1 - Voinchet, R. A1 - Wolters, J. A1 - Zoletnik, S. KW - Active spectroscopy KW - DIAGNOSTIC KW - engineering KW - ITER KW - Mechanical KW - Optical design KW - Upper port plug AB - The ITER core charge exchange recombination spectroscopy (core CXRS) diagnostic system is designed to provide experimental access to various measurement quantities in the ITER core plasma such as ion densities, temperatures and velocities. The implementation of the approved CXRS diagnostic principle on ITER faces significant challenges: First, a comparatively low CXRS signal intensity is expected, together with a high noise level due to bremsstrahlung, while the requested measurement accuracy and stability for the core CXRS system go far beyond the level commonly achieved in present-day fusion experiments. Second, the lifetime of the first mirror surface is limited due to either erosion by fast particle bombardment or deposition of impurities. Finally, the hostile technical environment on ITER imposes challenging boundary conditions for the diagnostic integration and operation, including high neutron loads, electromagnetic loads, seismic events and a limited access for maintenance. A brief overview on the R&D and design activities for the core CXRS system is presented here, while the details are described in parallel papers. (C) 2011 Elsevier B.V. All rights reserved. VL - 86 SN - 0920-3796 IS - 6-8 U1 - FP U2 - PDG U5 - 18eb9c792308e37ef35ec40b3031bfff ER - TY - JOUR T1 - Overview of the ITER EC H&CD system and its capabilities JF - Fusion Engineering and Design Y1 - 2011 A1 - Omori, T. A1 - Henderson, M. A. A1 - Albajar, F. A1 - Alberti, S. A1 - Baruah, U. A1 - Bigelow, T. S. A1 - Becket, B. A1 - Bertizzolo, R. A1 - Bonicelli, T. A1 - Brusch, A. A1 - Caughman, J. B. A1 - Chavan, R. A1 - Cirant, S. A1 - Collazos, A. A1 - Cox, D. A1 - Darbos, C. A1 - M.R. de Baar A1 - Denisov, G. A1 - Farina, D. A1 - Gandini, F. A1 - Gassmann, T. A1 - Goodman, T. P. A1 - Heidinger, R. A1 - Hogge, J. P. A1 - Illy, S. A1 - Jean, O. A1 - Jin, J. A1 - Kajiwara, K. A1 - Kasparek, W. A1 - Kasugai, A. A1 - Kern, S. A1 - Kobayashi, N. A1 - Kumric, H. A1 - Landis, J. D. A1 - Moro, A. A1 - Nazare, C. A1 - Oda, Y. A1 - Pagonakis, I. A1 - Piosczyk, B. A1 - Platania, P. A1 - Plaum, B. A1 - Poli, E. A1 - Porte, L. A1 - Purohit, D. A1 - Ramponi, G. A1 - Rao, S. L. A1 - Rasmussen, D. A. A1 - Ronden, D. M. S. A1 - Rzesnicki, T. A1 - Saibene, G. A1 - Sakamoto, K. A1 - Sanchez, F. A1 - Scherer, T. A1 - Shapiro, M. A. A1 - Sozzi, C. A1 - Spaeh, P. A1 - Strauss, D. A1 - Sauter, O. A1 - Takahashi, K. A1 - Temkin, R. J. A1 - Thumm, M. A1 - Tran, M. Q. A1 - Udintsev, V.S. A1 - Zohm, H. KW - DESIGN KW - Electron Cyclotron KW - gyrotron KW - ITER KW - launcher KW - MHD stabilization AB -The Electron Cyclotron (EC) system for the ITER tokamak is designed to inject >= 20 MW RF power into the plasma for Heating and Current Drive (H&CD) applications. The EC system consists of up to 26 gyrotrons (between 1 and 2 MW each), the associated power supplies, 24 transmission lines and 5 launchers. The EC system has a diverse range of applications including central heating and current drive, current profile tailoring and control of plasma magneto-hydrodynamic (MUD) instabilities such as the sawtooth and neoclassical tearing modes (NTMs). This diverse range of applications requires the launchers to be capable of depositing the EC power across nearly the entire plasma cross section. This is achieved by two types of antennas: an equatorial port launcher (capable of injecting up to 20 MW from the plasma axis to mid-radius) and four upper port launchers providing access from inside of mid radius to near the plasma edge. The equatorial launcher design is optimized for central heating, current drive and profile tailoring, while the upper launcher should provide a very focused and peaked current density profile to control the plasma instabilities. The overall EC system has been modified during the past 3 years taking into account the issues identified in the ITER design review from 2007 and 2008 as well as integrating new technologies. This paper will review the principal objectives of the EC system, modifications made during the past 2 years and how the design is compliant with the principal objectives. (C) 2011 ITER Organization. Published by Elsevier B.V. All rights reserved.

VL - 86 SN - 0920-3796 IS - 6-8 N1 - ISI Document Delivery No.: 853KZTimes Cited: 1Cited Reference Count: 2526th Symposium on Fusion Technology (SOFT)SEP 27-OCT 01, 2010Porto, PORTUGALInst Plasmas Fusao Nucl (IPFN), Commiss European Union, Inst Soldadura Qualidade (ISQ), Fundacao Ciencia Tecnologia (FCT), Univ Tecnica Lisboa (UTL), TAP, Andante U1 -FP

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U5 - c370f03eff7f1a27a15a35be0d59bc06 ER - TY - JOUR T1 - An overview of control system for the ITER electron cyclotron system JF - Fusion Engineering and Design Y1 - 2011 A1 - Purohit, D. A1 - Bigelow, T. A1 - Billava, D. A1 - Bonicelli, T. A1 - Caughman, J. A1 - Darbos, C. A1 - Denisov, G. A1 - Gandini, F. A1 - Gassmann, T. A1 - Henderson, M. A1 - Journeux, J. Y. A1 - Kajiwara, K. A1 - Kobayashi, N. A1 - Nazare, C. A1 - Oda, Y. A1 - Omori, T. A1 - Rao, S. L. A1 - Rasmussen, D. A1 - Ronden, D. A1 - Saibene, G. A1 - Sakamoto, K. A1 - Sartori, F. A1 - Takahashi, K. A1 - Temkin, R. KW - ECH KW - ECRH KW - I&C KW - ITER AB -The ITER electron cyclotron (EC) system having capability of up to 26 MW generated power at 170 GHz is being procured by 5 domestic agencies via 10 procurement arrangements. This implies diverse types of equipment and complex interface management. It also places a challenge on control system architecture to entertain the constraints of procurement slicing and meeting the overall functional requirement. The envisioned architecture is to use the local control units (supplied with each procurement) and a supervisory plant controller (by ITER). This offers a reliable control configuration for such delicate and complex EC plant system. The control system is envisioned to monitor the whole plant and perform automated tasks that are today performed via direct human intervention. For example, the automated gyrotron conditioning and active control of the EC plant to respond to requests from the plasma control system (PCS). This later aspect requires rapid shut down of the gyrotrons and power supplies, deviation of the actuators to direct the power from an equatorial to upper launcher and then restart of the power generation for rapid stabilization of the magneto hydrodynamic (MHD) instabilities that occur in high performance plasma operation. The plant controller will be designed for optimized performance with the PCS and the feedback control system used to actively control the power (with modulation capability up to 5 kHz) and launching direction for MHD stabilization. (C) 2011 ITER Organization. Published by Elsevier B.V. All rights reserved.

VL - 86 SN - 0920-3796 IS - 6-8 N1 - ISI Document Delivery No.: 853KZTimes Cited: 0Cited Reference Count: 626th Symposium on Fusion Technology (SOFT)SEP 27-OCT 01, 2010Porto, PORTUGALInst Plasmas Fusao Nucl (IPFN), Commiss European Union, Inst Soldadura Qualidade (ISQ), Fundacao Ciencia Tecnologia (FCT), Univ Tecnica Lisboa (UTL), TAP, Andante U1 -FP

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U5 - 623667772e2b71c98dbb863a866ba3e4 ER - TY - JOUR T1 - Optimization of the availability of the core CXRS diagnostics for ITER JF - Fusion Engineering and Design Y1 - 2011 A1 - Klinkhamer, F. A1 - Krimmer, A. A1 - Biel, W. A1 - Hawkes, N. A1 - Kiss, G. A1 - Koning, J. A1 - Krasikov, Y. A1 - Neubauer, O. A1 - Snijders, B. KW - Core CXRS KW - Mirror cleaning KW - mirror lifetime KW - mirrors KW - Port plug AB -New optical configurations for the ITER core CXRS system offer the possibility of longer ducts between the first mirror and the plasma. This has led to a renewed optimization of the availability. using a simple model of the degradation of the first mirror that starts with the conditions of (a) the required measurement performance and (b) the geometry of the port plug. It is found that for a fully passive system the design should strive for the longest duct length possible. Given known data, this will result in a diagnostic lifetime still substantially shorter than ITER lifetime. When an option of cleaning the first mirror is introduced (assuming this is a feasible option) the optimum is less straightforward, because the lifetime of the second mirror then also becomes important. The optimum then depends on the ratio between the cleaning interval and the ITER lifetime. Options are presented for various sets of assumptions. Finally practical limitations of supporting subsystems (cleaning system, shutter, calibration system) may influence the final design. Examples of such limitations with their impact are presented. (C) 2011 Published by Elsevier B.V.

VL - 86 SN - 0920-3796 IS - 6-8 N1 - ISI Document Delivery No.: 853KZTimes Cited: 1Cited Reference Count: 1526th Symposium on Fusion Technology (SOFT)SEP 27-OCT 01, 2010Porto, PORTUGALInst Plasmas Fusao Nucl (IPFN), Commiss European Union, Inst Soldadura Qualidade (ISQ), Fundacao Ciencia Tecnologia (FCT), Univ Tecnica Lisboa (UTL), TAP, Andante U1 -FP

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U5 - c3e1bf93f0a956948da3c7d3635f9e1e ER - TY - JOUR T1 - Overview of toroidal momentum transport JF - Nuclear Fusion Y1 - 2011 A1 - Peeters, A.G. A1 - Angioni, C. A1 - Bortolon, A. A1 - Camenen, Y. A1 - Casson, F. J. A1 - Duval, B. A1 - Fiederspiel, L. A1 - Hornsby, W. A. A1 - Idomura, Y. A1 - Hein, T. A1 - Kluy, N. A1 - Mantica, P. A1 - Parra, F. I. A1 - Snodin, A. P. A1 - Szepesi, G. A1 - Strintzi, D. A1 - Tala, T. A1 - Tardini, G. A1 - P. de Vries A1 - Weiland, J. KW - ALCATOR-C-MOD KW - ANGULAR-MOMENTUM KW - CYCLOTRON WAVE KW - FLOWS KW - INJECTION KW - ION TEMPERATURE KW - NEUTRAL-BEAM INJECTION KW - OFF-LAYER KW - OHMIC H-MODE KW - PLASMA ROTATION KW - RADIAL ELECTRIC-FIELD KW - TEMPERATURE-GRADIENT MODE AB -Toroidal momentum transport mechanisms are reviewed and put in a broader perspective. The generation of a finite momentum flux is closely related to the breaking of symmetry (parity) along the field. The symmetry argument allows for the systematic identification of possible transport mechanisms. Those that appear to lowest order in the normalized Larmor radius (the diagonal part, Coriolis pinch, E x B shearing, particle flux, and up-down asymmetric equilibria) are reasonably well understood. At higher order, expected to be of importance in the plasma edge, the theory is still under development.

VL - 51 SN - 0029-5515 IS - 9 N1 - ISI Document Delivery No.: 818DPTimes Cited: 1Cited Reference Count: 114SI U1 -FP

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U5 - 2efb8b4ed885eb95aa643b3d2cfd6e6b ER - TY - JOUR T1 - Overview of physics results from MAST JF - Nuclear Fusion Y1 - 2011 A1 - Lloyd, B. A1 - Akers, R. J. A1 - Alladio, F. A1 - Allan, S. A1 - Appel, L. C. A1 - Barnes, M. A1 - Barratt, N. C. A1 - N. Ben Ayed A1 - Breizman, B. N. A1 - Cecconello, M. A1 - Challis, C. D. A1 - Chapman, I.T. A1 - Ciric, D. A1 - Colyer, G. A1 - Connor, J. W. A1 - Conway, N. J. A1 - Cox, M. A1 - Cowley, S. C. A1 - Cunningham, G. A1 - Darke, A. A1 - De Bock, M. A1 - Delchambre, E. A1 - De Temmerman, G. A1 - Dendy, R. O. A1 - Denner, P. A1 - Driscoll, M. D. A1 - Dudson, B. A1 - Dunai, D. A1 - Dunstan, M. A1 - Elmore, S. A1 - Field, A. R. A1 - Fishpool, G. A1 - Freethy, S. A1 - Garzotti, L. A1 - Gibson, K. J. A1 - Gryaznevich, M. P. A1 - Guttenfelder, W. A1 - Harrison, J. A1 - Hastie, R. J. A1 - Hawkes, N. C. A1 - Hender, T. C. A1 - Hnat, B. A1 - Howell, D. F. A1 - Hua, M. D. A1 - Hubbard, A. A1 - Huysmans, G. A1 - Keeling, D. A1 - Kim, Y. C. A1 - Kirk, A. A1 - Liang, Y. A1 - Lilley, M. K. A1 - Lisak, M. A1 - Lisgo, S. A1 - Liu, Y. Q. A1 - G. P. Maddison A1 - Maingi, R. A1 - Manhood, S. J. A1 - Martin, R. A1 - McArdle, G. J. A1 - McCone, J. A1 - Meyer, H. A1 - Michael, C. A1 - Mordijck, S. A1 - Morgan, T. A1 - Morris, A. W. A1 - Muir, D. G. A1 - Nardon, E. A1 - Naylor, G. A1 - O'Brien, M. R. A1 - O'Gorman, T. A1 - Palenik, J. A1 - Patel, A. A1 - Pinches, S. D. A1 - Price, M. N. A1 - Roach, C. M. A1 - Rozhansky, V. A1 - Saarelma, S. A1 - Sabbagh, S. A. A1 - Saveliev, A. A1 - Scannell, R. A1 - Sharapov, S. E. A1 - Shevchenko, V. A1 - Shibaev, S. A1 - Stork, D. A1 - Storrs, J. A1 - Suttrop, W. A1 - Sykes, A. A1 - Tamain, P. A1 - Taylor, D. A1 - Temple, D. A1 - Thomas-Davies, N. A1 - Thornton, A. A1 - Turnyanskiy, M. R. A1 - Valovic, M. A1 - Vann, R. G. L. A1 - Voss, G. A1 - Walsh, M. J. A1 - Warder, S. E. V. A1 - Wilson, H. R. A1 - Windridge, M. A1 - Wisse, M. A1 - Zoletnik, S. KW - MODEL KW - TRANSPORT AB -Major developments on the Mega Amp Spherical Tokamak (MAST) have enabled important advances in support of ITER and the physics basis of a spherical tokamak (ST) based component test facility (CTF), as well as providing new insight into underlying tokamak physics. For example, L-H transition studies benefit from high spatial and temporal resolution measurements of pedestal profile evolution (temperature, density and radial electric field) and in support of pedestal stability studies the edge current density profile has been inferred from motional Stark effect measurements. The influence of the q-profile and E x B flow shear on transport has been studied in MAST and equilibrium flow shear has been included in gyro-kinetic codes, improving comparisons with the experimental data. H-modes exhibit a weaker q and stronger collisionality dependence of heat diffusivity than implied by IPB98(gamma, 2) scaling, which may have important implications for the design of an ST-based CTF. ELM mitigation, an important issue for ITER, has been demonstrated by applying resonant magnetic perturbations (RMPs) using both internal and external coils, but full stabilization of type-I ELMs has not been observed. Modelling shows the importance of including the plasma response to the RMP fields. MAST plasmas with q > 1 and weak central magnetic shear regularly exhibit a long-lived saturated ideal internal mode. Measured plasma braking in the presence of this mode compares well with neo-classical toroidal viscosity theory. In support of basic physics understanding, high resolution Thomson scattering measurements are providing new insight into sawtooth crash dynamics and neo-classical tearing mode critical island widths. Retarding field analyser measurements show elevated ion temperatures in the scrape-off layer of L-mode plasmas and, in the presence of type-I ELMs, ions with energy greater than 500 eV are detected 20 cm outside the separatrix. Disruption mitigation by massive gas injection has reduced divertor heat loads by up to 70%.

VL - 51 SN - 0029-5515 IS - 9 N1 - ISI Document Delivery No.: 818DPTimes Cited: 0Cited Reference Count: 60SI U1 -PSI

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U5 - 9e9434147fddee9a0003e22469980e60 ER - TY - JOUR T1 - Overview of ASDEX Upgrade results JF - Nuclear Fusion Y1 - 2011 A1 - Kallenbach, A. A1 - Adamek, J. A1 - Aho-Mantila, L. A1 - Akaslompolo, S. A1 - Angioni, C. A1 - Atanasiu, C. V. A1 - Balden, M. A1 - Behler, K. A1 - Belonohy, E. A1 - Bergmann, A. A1 - Bernert, M. A1 - Bilato, R. A1 - Bobkov, V. A1 - Boom, J. A1 - Bottino, A. A1 - Braun, F. A1 - Brudgam, M. A1 - Buhler, A. A1 - Burckhart, A. A1 - Chankin, A. A1 - Classen, I.G.J. A1 - Conway, G. D. A1 - Coster, D. P. A1 - de Marne, P. A1 - D'Inca, R. A1 - Drube, R. A1 - Dux, R. A1 - Eich, T. A1 - Endstrasser, N. A1 - Engelhardt, K. A1 - Esposito, B. A1 - Fable, E. A1 - Fahrbach, H. U. A1 - Fattorini, L. A1 - Fischer, R. A1 - Flaws, A. A1 - Funfgelder, H. A1 - Fuchs, J. C. A1 - Gal, K. A1 - Munoz, M. G. A1 - Geiger, B. A1 - Adamov, M. G. A1 - Giannone, L. A1 - Giroud, C. A1 - Gorler, T. A1 - da Graca, S. A1 - Greuner, H. A1 - Gruber, O. A1 - Gude, A. A1 - Gunter, S. A1 - Haas, G. A1 - Hakola, A. H. A1 - Hangan, D. A1 - Happel, T. A1 - Hauff, T. A1 - Heinemann, B. A1 - Herrmann, A. A1 - Hicks, N. A1 - Hobirk, J. A1 - Hohnle, H. A1 - Holzl, M. A1 - Hopf, C. A1 - Horton, L. A1 - Huart, M. A1 - Igochine, V. A1 - Ionita, C. A1 - Janzer, A. A1 - Jenko, F. A1 - Kasemann, C. P. A1 - Kalvin, S. A1 - Kardaun, O. A1 - Kaufmann, M. A1 - Kirk, A. A1 - Klingshirn, H. J. A1 - Kocan, M. A1 - Kocsis, G. A1 - Kollotzek, H. A1 - Konz, C. A1 - Koslowski, R. A1 - Krieger, K. A1 - Kurki-Suonio, T. A1 - Kurzan, B. A1 - Lackner, K. A1 - Lang, P. T. A1 - Lauber, P. A1 - Laux, M. A1 - Leipold, F. A1 - Leuterer, F. A1 - Lohs, A. A1 - N C Luhmann Jr. A1 - Lunt, T. A1 - Lyssoivan, A. A1 - Maier, H. A1 - Maggi, C. A1 - Mank, K. A1 - Manso, M. E. A1 - Maraschek, M. A1 - Martin, P. A1 - Mayer, M. A1 - McCarthy, P. J. A1 - McDermott, R. A1 - Meister, H. A1 - Menchero, L. A1 - Meo, F. A1 - Merkel, P. A1 - Merkel, R. A1 - Mertens, V. A1 - Merz, F. A1 - Mlynek, A. A1 - Monaco, F. A1 - Muller, H. W. A1 - Munich, M. A1 - Murmann, H. A1 - Neu, G. A1 - Neu, R. A1 - Nold, B. A1 - Noterdaeme, J. M. A1 - Park, H. K. A1 - Pautasso, G. A1 - Pereverzev, G. A1 - Podoba, Y. A1 - Pompon, F. A1 - Poli, E. A1 - Polochiy, K. A1 - Potzel, S. A1 - Prechtl, M. A1 - Puschel, M. J. A1 - Putterich, T. A1 - Rathgeber, S. K. A1 - Raupp, G. A1 - Reich, M. A1 - Reiter, B. A1 - Ribeiro, T. A1 - Riedl, R. A1 - Rohde, V. A1 - Roth, J. A1 - Rott, M. A1 - Ryter, F. A1 - Sandmann, W. A1 - Santos, J. A1 - Sassenberg, K. A1 - Sauter, P. A1 - Scarabosio, A. A1 - Schall, G. A1 - Schmid, K. A1 - Schneider, P. A. A1 - Schneider, W. A1 - Schramm, G. A1 - Schrittwieser, R. A1 - Schweinzer, J. A1 - Scott, B. A1 - Sempf, M. A1 - Serra, F. A1 - Sertoli, M. A1 - Siccinio, M. A1 - Sigalov, A. A1 - Silva, A. A1 - Sips, A.C.C. A1 - Sommer, F. A1 - Stabler, A. A1 - Stober, J. A1 - Streibl, B. A1 - Strumberger, E. A1 - Sugiyama, K. A1 - Suttrop, W. A1 - Szepesi, T. A1 - Tardini, G. A1 - Tichmann, C. A1 - Told, D. A1 - Treutterer, W. A1 - Urso, L. A1 - Varela, P. A1 - Vincente, J. A1 - Vianello, N. A1 - Vierle, T. A1 - Viezzer, E. A1 - Vorpahl, C. A1 - Wagner, D. A1 - Weller, A. A1 - Wenninger, R. A1 - Wieland, B. A1 - Wigger, C. A1 - Willensdorfer, M. A1 - Wischmeier, M. A1 - Wolfrum, E. A1 - Wursching, E. A1 - Yadikin, D. A1 - Yu, Q. A1 - Zammuto, I. A1 - Zasche, D. A1 - Zehetbauer, T. A1 - Zhang, Y. A1 - Zilker, M. A1 - Zohm, H. KW - PHYSICS KW - REFLECTOMETRY KW - TOKAMAK AB - The ASDEX Upgrade programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. After the finalization of the tungsten coating of the plasma facing components, the re-availability of all flywheel-generators allowed high-power operation with up to 20 MW heating power at I(p) up to 1.2 MA. Implementation of alternative ECRH schemes (140 GHz O2- and X3-mode) facilitated central heating above n(e) = 1.2 x 10(20) m(-3) and low q(95) operation at B(t) = 1.8 T. Central O2-mode heating was successfully used in high P/R discharges with 20 MW total heating power and divertor load control with nitrogen seeding. Improved energy confinement is obtained with nitrogen seeding both for type-I and type-III ELMy conditions. The main contributor is increased plasma temperature, no significant changes in the density profile have been observed. This behaviour may be explained by higher pedestal temperatures caused by ion dilution in combination with a pressure limited pedestal and hollow nitrogen profiles. Core particle transport simulations with gyrokinetic calculations have been benchmarked by dedicated discharges using variations of the ECRH deposition location. The reaction of normalized electron density gradients to variations of temperature gradients and the T(e)/T(i) ratio could be well reproduced. Doppler reflectometry studies at the L-H transition allowed the disentanglement of the interplay between the oscillatory geodesic acoustic modes, turbulent fluctuations and the mean equilibrium E x B flow in the edge negative E(r) well region just inside the separatrix. Improved pedestal diagnostics revealed also a refined picture of the pedestal transport in the fully developed H-mode type-I ELM cycle. Impurity ion transport turned out to be neoclassical in between ELMs. Electron and energy transport remain anomalous, but exhibit different recovery time scales after an ELM. After recovery of the pre-ELM profiles, strong fluctuations develop in the gradients of n(e) and T(e). The occurrence of the next ELM cannot be explained by the local current diffusion time scale, since this turns out to be too short. Fast ion losses induced by shear Alfven eigenmodes have been investigated by time-resolved energy and pitch angle measurements. This allowed the separation of the convective and diffusive loss mechanisms. VL - 51 SN - 0029-5515 IS - 9 N1 - ISI Document Delivery No.: 818DPTimes Cited: 1Cited Reference Count: 45SI U1 - FP U2 - PDG U5 - a193177a90d5b600862ca1e40bcc67af ER - TY - JOUR T1 - Optical boundary reconstruction of tokamak plasmas for feedback control of plasma position and shape JF - Review of Scientific Instruments Y1 - 2010 A1 - Hommen, G. A1 - de M. Baar A1 - Nuij, P. A1 - McArdle, G. A1 - Akers, R. A1 - Steinbuch, M. KW - ASDEX KW - CONFINEMENT KW - EDGE KW - H-mode KW - MAST AB - A new diagnostic is developed to reconstruct the plasma boundary using visible wavelength images. Exploiting the plasma's edge localized and toroidally symmetric emission profile, a new coordinate transform is presented to reconstruct the plasma boundary from a poloidal view image. The plasma boundary reconstruction is implemented in MATLAB and applied to camera images of Mega-Ampere Spherical Tokamak discharges. The optically reconstructed plasma boundaries are compared to magnetic reconstructions from the offline reconstruction code EFIT, showing very good qualitative and quantitative agreement. Average errors are within 2 cm and correlation is high. In the current software implementation, plasma boundary reconstruction from a single image takes 3 ms. The applicability and system requirements of the new optical boundary reconstruction, called OFIT, for use in both feedback control of plasma position and shape and in offline reconstruction tools are discussed. (C) 2010 American Institute of Physics. [doi:10.1063/1.3499219] VL - 81 SN - 0034-6748 UR -