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In this paper, a tomography approach aiming at reconstructing a phase-space structure is proposed. For the phase-space resolved diagnostic system, a signal must be decomposed in real-space, velocity-space, and time; therefore, it is challenging to obtain a sufficiently high signal intensity in a single detector bin. To overcome this difficulty, three different sets of data having different integration directions in real-space, velocity-space, and time are simultaneously used, and a reconstruction of the original structure in the phase-space is attempted by a tomographic manner. The proposed method is demonstrated using a synthetic dataset in the actual diagnostic setup in the Large Helical Device. Time evolution of a phase-space perturbation induced by the Landau damping, which is caused by energetic particle-driven magnetohydrodynamic bursts, is successfully reconstructed by this method. Robustness against realistic diagnostic noise is also presented.

In state-of-the-art stellarators, turbulence is a major cause of the degradation of plasma confinement. To maximize confinement, which eventually determines the amount of nuclear fusion reactions, turbulent transport needs to be reduced. Here we report the observation of a confinement regime in a stellarator plasma that is characterized by increased confinement and reduced turbulent fluctuations. The transition to this regime is driven by the injection of submillimetric boron powder grains into the plasma. With the line-averaged electron density being kept constant, we observe a substantial increase of stored energy and electron and ion temperatures. At the same time, the amplitude of the plasma turbulent fluctuations is halved. While lower frequency fluctuations are damped, higher frequency modes in the range between 100 and 200 kHz are excited. We have observed this regime for different heating schemes, namely with both electron and ion cyclotron resonant radio frequencies and neutral beams, for both directions of the magnetic field and both hydrogen and deuterium plasmas.

Doppler-backscattering (DBS) has been used in several fusion plasma devices because it can measure the perpendicular velocity of electron density perturbation v⊥, the radial electric field E r, and the perpendicular wavenumber spectrum S( k⊥) with high wavenumber and spatial resolution. In particular, recently constructed frequency comb DBS systems enable observation of turbulent phenomena at multiple observation points in the radial direction. A dual-comb microwave DBS system has been developed for the large helical device plasma measurement. Since it is desirable to control the gain of each frequency-comb separately, a frequency-comb DBS system was developed with a function to adjust the gain of the scattered signal intensity of each channel separately. A correction processing method was also developed to correct the amplitude ratio and the phase difference between the in-phase and quadrature-phase signals of the scattered signals. As a result, the error in Doppler-shift estimation required to observe vertical velocity and the radial electric field was reduced, which enables more precise measurements.

We present recent highlights from the most recent operation phases of Wendelstein 7-X, the most advanced stellarator in the world. Stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, have been maintained for tens of seconds. Pellet fueling allows for plasma phases with reduced ion-temperature-gradient turbulence, and during such phases, the overall confinement is so good (energy confinement times often exceeding 200 ms) that the attained density and temperature profiles would not have been possible in less optimized devices, since they would have had neoclassical transport losses exceeding the heating applied in W7-X. This provides proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X plasmas generally show good impurity screening and high plasma purity, but there is evidence of longer impurity confinement times during turbulence-suppressed phases. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 and 2019-2020 under Grant Agreement No. 633053. Peer Reviewed Article signat per 497 autors/es: Thomas Sunn Pedersen1,2,∗ , I. Abramovic3, P. Agostinetti4, M. Agredano Torres1, S. Äkäslompolo1, J. Alcuson Belloso1, P. Aleynikov1, K. Aleynikova1, M. Alhashimi1, A. Ali1, N. Allen5, A. Alonso6, G. Anda7, T. Andreeva1, C. Angioni8, A. Arkhipov8, A. Arnold1, W. Asad8, E. Ascasibar6, M.-H. Aumeunier9, K. Avramidis10, E. Aymerich11, S.-G. Baek3, J. Bähner1, A. Baillod12, M. Balden1, M. Balden8, J. Baldzuhn1, S. Ballinger3, M. Banduch1, S. Bannmann1, A. Banon Navarro8, A. Bañon Navarro ´ 1, T. Barbui13, C. Beidler1, C. Belafdil9, A. Bencze7, A. Benndorf1, M. Beurskens1, C. Biedermann1, O. Biletskyi14, B. Blackwell15, M. Blatzheim1, T. Bluhm1, D. Böckenhoff1, G. Bongiovi16, M. Borchardt1, D. Borodin17, J. Boscary8, H. Bosch1,18, T. Bosmann19, B. Böswirth8, L. Böttger1, A. Bottino8, S. Bozhenkov1, R. Brakel1, C. Brandt1, T. Bräuer1, H. Braune1, S. Brezinsek17, K. Brunner1, S. Buller1, R. Burhenn1, R. Bussiahn1, B. Buttenschön1, A. Buzás7, V. Bykov1, I. Calvo6, K. Camacho Mata1, I. Caminal20, B. Cannas11, A. Cappa6, A. Carls1, F. Carovani1, M. Carr21, D. Carralero6, B. Carvalho22, J. Casas20, D. Castano-Bardawil17, F. Castejon6, N. Chaudhary1, I. Chelis23, A. Chomiczewska24, J.W. Coenen13,17, M. Cole1, F. Cordella25, Y. Corre9, K. Crombe26, G. Cseh7, B. Csillag7, H. Damm1, C. Day10, M. de Baar27, E. De la Cal6, S. Degenkolbe1, A. Demby13, S. Denk3, C. Dhard1, A. Di Siena8,28, A. Dinklage12, T. Dittmar17, M. Dreval14, M. Drevlak1, P. Drewelow1, P. Drews17, D. Dunai7, E. Edlund3, F. Effenberg29, G. Ehrke1, M. Endler1, D.A. Ennis5, F.J. Escoto6, T. Estrada6, E. Fable8, N. Fahrenkamp1, A. Fanni11, J. Faustin1, J. Fellinger1, Y. Feng1, W. Figacz4, E. Flom13, O. Ford1, T. Fornal24, H. Frerichs13, S. Freundt1, G. Fuchert1, M. Fukuyama30, F. Füllenbach1, G. Gantenbein10, Y. Gao1, K. Garcia13, J.M. García Regaña6, I. García-Cortés6, J. Gaspar31, D.A. Gates29, J. Geiger1, B. Geiger13, L. Giudicotti32, A. González6, A. Goriaev26,33, D. Gradic1, M. Grahl1, J.P. Graves12, J. Green13, E. Grelier9, H. Greuner8, S. Groß1, H. Grote1, M. Groth34, M. Gruca24, O. Grulke1,35, M. Grün1, J. Guerrero Arnaiz1, S. Günter8, V. Haak1, M. Haas1, P. Hacker1, A. Hakola36, A. Hallenbert1, K. Hammond29, X. Han17,37, S.K. Hansen3, J.H. Harris38, H. Hartfuß1, D. Hartmann1, D. Hathiramani1, R. Hatzky8, J. Hawke39, S. Hegedus7, B. Hein8, B. Heinemann8, P. Helander12, S. Henneberg1, U. Hergenhahn8,40, C. Hidalgo6, F. Hindenlang8, M. Hirsch1, U. Höfel1, K.P. Hollfeld17, A. Holtz1, D. Hopf8, D. Höschen17, M. Houry9, J. Howard19, X. Huang41, M. Hubeny17, S. Hudson29, K. Ida9, Y. Igitkhanov10, V. Igochine8, S. Illy10, C. Ionita-Schrittwieser42, M. Isobe39, M. Jabłczynska ´ 24, S. Jablonski24, B. Jagielski1, M. Jakubowski1, A. Jansen van Vuuren1, J. Jelonnek10, F. Jenko8, F. Jenko8, T. Jensen35, H. Jenzsch1, P. Junghanns8, J. Kaczmarczyk24, J. Kallmeyer1, U. Kamionka1, M. Kandler8, S. Kasilov43, Y. Kazakov26, D. Kennedy1, A. Kharwandikar1, M. Khokhlov1, C. Kiefer8, C. Killer1, A. Kirschner17, R. Kleiber1, T. Klinger12, S. Klose1, J. Knauer1, A. Knieps17, F. Köchl44, G. Kocsis7, Ya.I. Kolesnichenko45, A. Könies1, R. König1, J. Kontula34, P. Kornejew1, J. Koschinsky, M.M. Kozulia14, A. Krämer-Flecken17, R. Krampitz1, M. Krause1, N. Krawczyk24, T. Kremeyer1, L. Krier10, D.M. Kriete5, M. Krychowiak1, I. Ksiazek46, M. Kubkowska24, M. Kuczynski1, G. Kühner1, A. Kumar15, T. Kurki-Suonio34, S. Kwak1, M. Landreman47, P.T. Lang8, A. Langenberg1, H.P. Laqua12, H. Laqua1, R. Laube1, S. Lazerson1, M. Lewerentz1, C. Li17, Y. Liang17, Ch. Linsmeier17, J. Lion1, A. Litnovsky17,48, S. Liu37, J. Lobsien1, J. Loizu12, J. Lore38, A. Lorenz1, U. Losada6, F. Louche26, R. Lunsford29, V. Lutsenko45, M. Machielsen12, F. Mackel8, J. Maisano-Brown3, O. Maj8, D. Makowski49, G. Manduchi50, E. Maragkoudakis6, O. Marchuk17, S. Marsen1, E. Martines4, J. Martinez-Fernandez6, M. Marushchenko1, S. Masuzaki41, D. Maurer5, M. Mayer8, K.J. McCarthy6, O. Mccormack4, P. McNeely1, H. Meister8, B. Mendelevitch8, S. Mendes1, A. Merlo1, A. Messian26, A. Mielczarek49, O. Mishchenko1, B. Missal1, R. Mitteau9, V.E. Moiseenko14, A. Mollen1, V. Moncada9, T. Mönnich1, T. Morisaki41, D. Moseev1, G. Motojima41, S. Mulas6, M. Mulsow1, M. Nagel1, D. Naujoks1, V. Naulin35, T. Neelis19, H. Neilson29, R. Neu8, O. Neubauer17, U. Neuner1, D. Nicolai17, S.K. Nielsen35, H. Niemann1, T. Nishiza1, T. Nishizawa1, T. Nishizawa8, C. Nührenberg1, R. Ochoukov8, J. Oelmann17, G. Offermanns17 K. Ogawa41, S. Okamura41, J. Ölmanns17, J. Ongena26, J. Oosterbeek1, M. Otte1, N. Pablant29, N. Panadero Alvarez6, N. Panadero Alvarez6, A. Pandey1, E. Pasch1, R. Pavlichenko14, A. Pavone1, E. Pawelec46, G. Pechstein1, G. Pelka24, V. Perseo1, B. Peterson41, D. Pilopp1, S. Pingel1, F. Pisano11, B. Plöckl8, G. Plunk1, P. Pölöskei1, B. Pompe2, A. Popov51, M. Porkolab3, J. Proll19, M.J. Pueschel19,27, M.-E. Puiatti52, A. Puig Sitjes1, F. Purps1, K. Rahbarnia1, M. Rasinski ´ 17, J. Rasmussen35, A. Reiman29, F. Reimold1, M. Reisner8, D. Reiter17, M. Richou9, R. Riedl8, J. Riemann1, K. Riße1, G. Roberg-Clark1, V. Rohde8, J. Romazanov17, D. Rondeshagen1, P. Rong1, L. Rudischhauser1, T. Rummel1, K. Rummel1, A. Runov1, N. Rust1, L. Ryc24, P. Salembier20, M. Salewski35, E. Sanchez6, S. Satake41, G. Satheeswaran17, J. Schacht1, E. Scharff1, F. Schauer8, J. Schilling1, G. Schlisio1, K. Schmid8, J. Schmitt5, O. Schmitz13, W. Schneider1, M. Schneider1, P. Schneider8, R. Schrittwieser42, T. Schröder1, M. Schröder1, R. Schroeder1, B. Schweer26, D. Schwörer1, E. Scott1, E. Scott8, B. Shanahan1, G. Sias11, P. Sichta29, M. Singer1, P. Sinha29, S. Sipliä34, C. Slaby1, M. Sleczka53, H. Smith1, J. Smoniewski54, E. Sonnendrücker8, M. Spolaore4, A. Spring1, R. Stadler8, D. Stanczak24, T. Stange1, I. Stepanov26, L. Stephey13, J. Stober8, U. Stroth8,55, E. Strumberger8, C. Suzuki41, Y. Suzuki41, J. Svensson1, T. Szabolics7, T. Szepesi7, M. Szücs7, F.L. Tabares6, N. Tamura41, A. Tancetti35, C. Tantos10, J. Terry3, H. Thienpondt6, H. Thomsen1, M. Thumm10, J.M. Travere9, P. Traverso5, J. Tretter8, E. Trier8, H. Trimino Mora1, T. Tsujimura41, Y. Turkin1, A. Tykhyi45, B. Unterberg17, P. van Eeten1, B.Ph. van Milligen6, M. van Schoor26, L. Vano1, S. Varoutis10, M. Vecsei7, L. Vela56, J.L. Velasco6, M. Vervier17, N. Vianello50, H. Viebke1, R. Vilbrandt1, G. Vogel8, N. Vogt1, C. Volkhausen1, A. von Stechow1, F. Wagner1, E. Wang17, H. Wang57, F. Warmer1, T. Wauters26, L. Wegener1, T. Wegner1, G. Weir1, U. Wenzel1, A. White3, F. Wilde1, F. Wilms1, T. Windisch1, M. Winkler1, A. Winter1, V. Winters1, R. Wolf118, A.M. Wright29, G.A. Wurden39, P. Xanthopoulos1, S. Xu17, H. Yamada58, H. Yamaguchi41, M. Yokoyama41, M. Yoshinuma41, Q. Yu8, M. Zamanov14, M. Zanini1, M. Zarnstorff29, D. Zhang1, S. Zhou17, J. Zhu1, C. Zhu29, M. Zilker8, A. Zocco1, H. Zohm8, S. Zoletnik7 and L. Zsuga7 // 1 Max Planck Institute for Plasma Physics, Garching and Greifswald, Germany: 2 University of Greifswald, Greifswald, Germany; 3 Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139, United States of America; 4 Consorzio RFX, Corso Stati Uniti, 4-35127 Padova, Italy; 5 Auburn University, Auburn, AL 36849, United States of America; 6 CIEMAT, Avenida Complutense, 40, 28040 Madrid, Spain; 7 Center for Energy Research, Konkoly-Thegeut 29-33, 1121 Budapest, Hungary; 8 Max-Planck-Institute for Plasma Physics, Boltzmannstraße 2, 85748 Garching bei München, Germany; 9 CEA Cadarache, 13115 Saint-Paul-lez-Durance, France; 10 Karlsruhe Institute of Technology, Kaiserstr. 12, 76131 Karlsruhe, Germany; 11 University of Cagliari, Via Universita, 40, 09124 Cagliari, Italy; 12 École Polytechnique Fédérale de Lausanne, Swiss Plasma Center, CH-1015 Lausanne, Switzerland; 13 University of Wisconsin–Madison, Engineering Drive, Madison, WI 53706, United States of America; 14 Institute of Plasma Physics, National Science Center ‘Kharkiv Institute of Physics and Technology’, Kharkiv, Ukraine; 15 The Australian National University, Acton ACT 2601, Canberra, Australia; 16 Department of Engineering, University of Palermo, Viale delle Scienze, Edificio 6, Palermo, 90128, Italy; 17 Forschungszentrum Jülich GmbH, Institut für Energie-und Klimaforschung—Plasmaphysik, 52425 Jülich, Germany; 18 Technical University of Berlin, Strasse des 17. Juni 135, 10623 Berlin, Germany; 19 Eindhoven University of Technology, 5600 MB Eindhoven, Netherlands; 20 Universitat Politècnica de Catalunya. BarcelonaTech, C. Jordi Girona, 31, 08034 Barcelona, Spain; 21 Culham Center for Fusion Energy, Abingdon OX14 3EB, United Kingdom; 22 Instituto de Plasmas e Fusao Nuclear, Av. Rovisco Pais, 1049-001 Lisboa, Portugal; 23 Department of Physics, National and Kapodistrian University of Athens, 15784 Athens, Greece; 24 Institute of Plasma Physics and Laser Microfusion, 23 Hery Str., 01-497 Warsaw, Poland; 25 ENEA—Centro Ricerche Frascati, Via Enrico Fermi, 45, 00044 Frascati RM, Italy; 26 Laboratory for Plasma Physics, LPP-ERM/KMS, TEC Partner, B-1000 Brussels, Belgium; 27 Dutch Institute for Fundamental Energy Research, PO Box 6336, 5600 HH Eindhoven, Netherlands; 28 University of Texas, Austin, TX, United States of America; 29 Princeton Plasma Physics Laboratory, Princeton, NJ 08543, United States of America; 30 Kyushu University, 744 Motooka Nishi-ku, Fukuoka 819-0395, Japan; 31 Aix-Marseille University, Jardin du Pharo, 58 Boulevard Charles Livon, 13007, Marseille, France; 32 Department of Physics and Astronomy, Padova University, Via Marzolo 8, 35131 Padova, Italy; 33 Department of Applied Physics, Ghent University, Sint-Pietersnieuwstraat 41 B4, 9000 Ghent, Belgium; 34 Aalto University, 02150 Espoo, Finland; 35 Department of Physics, Technical University of Denmark, Anker Engelunds Vej, 2800 Kgs Lyngby, Denmark; 36 VTT Technical Research Center of Finland Ltd., PO Box 1000, FI-02044 VTT, Finland; 37 Institute of Plasma Physics, Chinese Academy of Sciences, 230031 Hefei, Anhui, China; 38 Oak Ridge National Laboratory, 1 Bethel Valley Rd, Oak Ridge, TN 37830, United States of America; 39 Los Alamos National Laboratory, NM 87545, United States of America; 40 Fritz-Haber-Institut der Max-Planck-Gesellschaft, 14195 Berlin, Germany; 41 National Institute for Fusion Science, National Institutes of Natural Sciences, 322-6 Oroshi-cho, Toki, Gifu Prefecture 509-5292, Japan; 42 Institute for Ion Physics and Applied Physics, University of Innsbruck, Innsbruck, Austria; 43 Graz University of Technology, Rechbauerstraße 12, 8010 GRAZ, Austria; 44 Austrian Academy of Science, Doktor-Ignaz-Seipel-Platz 2, 1010 Wien, Austria; 45 Institute for Nuclear Research, prospekt Nauky 47, Kyiv 03028, Ukraine; 46 University of Opole, plac Kopernika 11a, 45-001 Opole, Poland; 47 University of Maryland, Paint Branch Drive, College Park, MA 20742, United States of America; 48 National Research Nuclear University MEPhI, 115409 Moscow, Russian Federation; 49 Department of Microelectronics and Computer Science, Lodz University of Technology, Wolczanska 221/223, 90-924 Lodz, Poland; 50 Consiglio Nazionale delle Ricerche, Piazzale Aldo Moro, 7, 00185 Roma, Italy; 51 Ioffe Physical-Technical Institute of the Russian Academy of Sciences, 26 Politekhnicheskaya, St Petersburg 194021, Russian Federation; 52 Istituto di Fisica del Plasma Piero Caldirola, Via Roberto Cozzi, 53, 20125 Milano, Italy; 53 University of Szczecin, 70-453, aleja Papieza Jana Pawła II 22A, Szczecin, Poland; 54 Lawrence University, 711 E Boldt Way, Appleton, WI 54911, United States of America; 55 Physik-Department E28, Technische Universität München, 85747 Garching, Germany; 56 Universidad Carlos III de Madrid, Av. de la Universidad, 30 Madrid, Spain; 57 Yale University, New Haven, CT 06520, United States of America; 58 University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa, Chhiab 277-0882, Japan Objectius de Desenvolupament Sostenible::7 - Energia Assequible i No Contaminant Objectius de Desenvolupament Sostenible::7 - Energia Assequible i No Contaminant::7.a - Per a 2030, augmentar la cooperació internacional per tal de facilitar l’accés a la investigació i a les tecnolo­gies energètiques no contaminants, incloses les fonts d’energia renovables, l’eficiència energètica i les tecnologies de combustibles fòssils avançades i menys contaminants, i promoure la inversió en infraestructures energètiques i tecnologies d’energia no contaminant

The preceding propagation of turbulence pulses has been observed for the first time in heat avalanche events during the collapse of the electron internal transport barrier (e-ITB) in the Large Helical Device. The turbulence and heat pulses are generated near the foot of the e-ITB and propagate to the peripheral region within a much shorter time than the diffusion timescale. The propagation speed of the turbulence pulse is approximately 10 km/s, which is faster than that of the heat pulse propagating at a speed of 1.5 km/s. The heat pulse propagates at approximately the same speed as that in the theoretical prediction, whereas the turbulence pulse propagates one order of magnitude faster than that in the prediction, thereby providing important insights into the physics of non-local transport.

Controlled particulate injections from the PPPL impurity powder dropper (IPD) into the Large Helical Device have demonstrated positive effects on the wall conditions on both an intra and inter-shot basis. Injections over a range of densities, input powers, pulse lengths, heating schemes, injection quantities and main ion species show conclusive evidence of improvement to plasma wall conditions. Successful injections are confirmed by both spectroscopic measurements as well as real-time visible camera signals. In 7 s long plasmas the responses include a reduction in wall recycling as well as a reduction in native impurity content as observed over the course of several discharges. For plasmas longer than 40 s, improvements to the recycling rate and increased impurity control are observed in real time as a consequence of the extended particulate injections. These experiments demonstrate the extended applicability of this solid particulate conditioning technique to the control and maintenance of the plasma wall conditions. In addition they are an important initial step in the development of the real-time boronization technique as a supplement to standard conditioning scenarios.

We assess the magnetic field configuration in modern fusion devices by comparing experiments with the same heating power, between a stellarator and a heliotron. The key role of turbulence is evident in the optimized stellarator, while neoclassical processes largely determine the transport in the heliotron device. Gyrokinetic simulations elucidate the underlying mechanisms promoting stronger ion scale turbulence in the stellarator. Similar plasma performances in these experiments suggests that neoclassical and turbulent transport should both be optimized in next step reactor designs.

A heating source with off-axis electron cyclotron heating (ECH) alone produced a plasma with a quasi-steady-state hollow electron-temperature profile in the Large Helical Device. The clear formation of this quasi-steady-state hollow electron-temperature profile can be explained by adding the outward heat convection term to the diffusion term, as a simple model to describe the electron heat flux, using the energy conservation equation. In addition, we directly observed the non-locality of the non-diffusive (convective) contribution in transient electron thermal transport in the condition that power-modulated on-axis ECH was applied to the plasma sustained by off-axis ECH. The experimentally evaluated flux-gradient relation shows two different positive values of the electron heat flux at zero temperature gradient by going back and forth between positive and negative temperature gradient regions in the transport hysteresis phenomenon.

In this study, the use of a spatial heterodyne spectrometer (SHS) to measure the toroidal flow velocity ( V f) and the ion temperature [Formula: see text] of the C6+ impurity ion by charge exchange spectroscopy was explored. The instrumental width (IW) of the SHS (aperture size = 16.77 mm2, etendue = 2.9867 mm2 sr) was extrapolated to be 0.09 nm, which is half of the 0.17 nm IW extrapolated for a conventionally used dispersive spectrometer (DS) (aperture size = 2.6 mm2, etendue = 0.2605 mm2 sr). The resulting V f and [Formula: see text] measurements were found to be in good agreement with those measured using the DS.

Density control is crucial for maintaining stable confined plasma. Divertor pumping, where neutral particles are compressed and exhausted in the divertor region, was developed for this task for the Large Helical Device. In this study, neutral particle pressure, which is related to recycling, was systematically scanned in the magnetic configuration by changing the magnetic axis position. High neutral particle pressure and compression were obtained in the divertor for a high plasma electron density and the inner magnetic axis configuration. Density control using divertor pumping with gas puffing was applied to electron cyclotron heated plasma in the inner magnetic axis configuration, which provides high neutral particle compression and exhaust in the divertor. Stable plasma density and electron temperature were maintained with divertor pumping. A heat analysis shows that divertor pumping did not affect edge electron heat conductivity, but it led to low electron heat conductivity in the core caused by electron-internal-transport-barrier-like formation.

In this study, we perform a comprehensive comparison of the transport hysteresis width in deuterium (D) plasmas, hydrogen (H) plasmas, and D-H mixed plasmas. The core focused modulation electron cyclotron resonance heating (MECH) is applied as the heat source perturbation, and the heat flux is evaluated using the energy conservation equation with the measured electron temperature response and the ECH deposition profile calculated by the ray-tracing scheme. Systematic density scan in plasmas with different ion mass reveals that there is no significant isotope effect in their hysteresis width. It is found that plasmas with heavier isotope mass can easily form the electron internal transport barrier. As the hysteresis width is insensitive to the isotope mass, the classical part of the diffusivity is considered to be responsible for the isotope effect in the transport barrier formation.

The internal transport barrier (ITB) appears in ion root plasmas with NBI heating, when the Te/Ti ratio is held below unity in LHD. The improvement of transport is observed only in the ion heat transport, not in the electron heat transport nor in the particle transport. During the formation of ion ITB, the region of low ion thermal diffusivity expands toward the plasma center and results in a wide weak ITB, which is in contrast to the narrow strong ITB typically observed in tokamaks (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

The neoclassical (NC) transport analyses have been performed to elucidate the plausible approaches towards higher ion-temperature heliotron plasmas. Avoidance of the ripple transport is the key issue, for which the neoclassical ambipolar radial electric field (Er) can be utilized. The ion-root scenario and the electron-root scenario are expected to be effective according to the experimental situation (especially, the temperature ratio between ions and electrons). The impact of the ion mass on the neoclassical ambipolar Er is also investigated to reveal the easier realization of electron-root Er in heavier ion plasmas (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

An improved confinement state by the formation of an edge transport barrier can be obtained in the compact helical system. A power threshold exists, as is the case for the H-mode in tokamaks. The threshold scales as . These dependences are similar to those in tokamaks. The threshold also depends on magnetic configuration parameters such as the position of the magnetic axis. The absolute value of the threshold is 1–3 times larger than the tokamak H-mode power scaling.

We study the effect of the net toroidal plasma current on the diamagnetic measurement for large toroidal current experiments in the Large Helical Device (LHD). In order to treat precisely the effect, we apply a new method that evaluates magnetic fluxes of magnetic measurements directly from plasma currents of a magnetohydrodynamic equilibrium state with net toroidal current. There is a large difference between the value calibrated by the new method and by a conventional method under the cylindrical and large aspect ratio plasma model. We calibrate the diamagnetic beta value using the net toroidal current distribution observed by motional Stark effect spectroscopy in a LHD discharge with a high net toroidal current, 40 kA T−1. We obtain the results that the temporal behaviour of the calibrated diamagnetic beta value is quite the same with the beta value based on the profile measurement by Thomson scattering and far-infrared interferometry.

The dependence of the Shafranov shift on magnetic field configuration, toroidicity and central rotational transform ι(0) in neutral beam heated plasma has been experimentally investigated in the Large Helical Device. The toroidicity of the plasma is controlled by the quadrupole field, while the central ι(0) is controlled by changing the distance of the current centre of the helical coil to the plasma. It is experimentally confirmed that both the lower toroidicity and the higher ι(0) contribute to the reduction of the Shafranov shift as predicted by the three-dimensional equilibrium code, VMEC.

On the Large Helical Device (LHD), low to high confinement (L–H) transition and edge transport barrier (ETB) formation were observed in the low beta regime (βdia 1.5%). In most of ETB plasmas electron density preferentially increases in the edge region without a substantial rise of the edge electron temperature. The ETB zone develops inside the ergodic field layer calculated in the vacuum field. The ETB formation strongly destabilizes edge coherent modes such as m/n = 2/3 or 1/2 (m, n: poloidal and toroidal mode numbers), because the plasma edge region is in the magnetic hill. The ETB is partially destroyed by the combination of these edge MHD modes and ELM-like activities. For a particular experimental condition, the forced generation of a sizable m/n = 1/1 magnetic island near the edge by application of external field perturbations facilitates the L–H transition at a lower electron density and suppresses edge MHD modes and ELM-like activities to lower levels.

In the Large Helical Device (LHD), two different divertor configurations, i.e. helical divertor (HD) and local island divertor (LID), are utilized to control the edge plasma. The HD with two X-points is an intrinsic divertor for heliotron devices, accompanied with a relatively thick ergodic layer outside the confinement region. Edge and divertor plasma behavior from low density to high density regimes is presented, referring to the divertor detachment. The effect of the ergodic layer on the edge transport is also discussed. On the other hand, the LID is an advanced divertor concept which realizes a high pumping efficiency by the combination of an externally induced magnetic island and a closed pumping system. Experimental results to confirm the fundamental divertor performance of the LID are presented.

On the Large Helical Device (LHD) where nested magnetic surfaces are surrounded by the ergodic field layer, edge transport barrier (ETB) was produced in neutral-beam-injection (NBI) heated plasmas through transition and non-transition processes. The former case is the ETB formation by L-H transition, where characteristics of L-H transition observed in a tokamak plasma are clearly recognized. The confinement improvement is the modest (~ 10%), compared with the ISS95 international stellarator scaling. The threshold power for the transition is comparable or slightly lower than the ITER scaling law established by tokamaks and compact tori. The ETB is formed inside the ergodic field layer of the vacuum field. The ETB formation destabilizes edge coherent modes such as m/n = 1/1, 2/3 and 1/2, of which rational surfaces are in the magnetic hill. The formed ETB is partially and transiently destroyed by these coherent edge MHD modes and edge localized modes (ELMs) typically observed in Hα signals. The latter ETB is observed in a plasma with large reversed NBI-driven current more than 100 kA at Bt = 1 T. In these plasmas, the edge magnetic shear is enhanced by the current and the rotational transform in the core region is expected to be appreciably reduced. Thus reduced rotational transform in the plasma central region will enhance outward heat and particle fluxes toward ergodic edge layer. The ETB with steep electron temperature gradient up to ~ 5 keV/m is formed by blocking enhanced outward heat flux.

A variety of electron density (ne) profiles have been observed in the Large Helical Device (LHD). The density profiles change dramatically with heating power and toroidal magnetic field (Bt). The particle transport coefficients, i.e. diffusion coefficient (D) and convection velocity (V) are experimentally obtained in the standard configuration from density modulation experiments. The values of D and V are estimated separately in the core and edge. The diffusion coefficients are found to be a function of electron temperature (Te), and vary with Bt. Edge diffusion coefficients are proportional to . Non-zero V is observed, and it is found that the electron temperature gradient can drive particle convection, particularly in the core region. The convection velocity both in the core and edge reverses direction from inward to outward as the Te gradient increases. However, the toroidal magnetic field also significantly affects the value and direction of V. The density fluctuation profiles are measured by a two-dimensional phase contrast interferometer. It was found that fluctuations which are localized in the edge propagate towards the ion diamagnetic direction in the laboratory frame, while the phase velocity of fluctuations around mid-radius is close to the plasma poloidal Er × Bt rotation velocity. The fluctuation level becomes larger as particle flux becomes larger in the edge region.

Edge transport barrier (ETB) formation was observed in a compact helical system. A sharp decrease in Hα emission indicates the quick transition of edge particle transport. The increase in the density gradient at the edge was measured by using various profile diagnostics. The transition process has a slowly developing pre-phase, and a quick transition for which the fastest case is less than 0.2 ms. There exists a heating power threshold which is roughly proportional to the density and the magnetic field. The transition and back transition is controlled by the heating power. The magnetic configuration effect on ETB formation was also studied. The local density measurement by beam emission spectroscopy shows intermittent bursts of low frequency fluctuations during the ETB formation phase. When the initial density profile is very hollow, ETB formation together with the electron temperature increase (electron internal transport barrier) in the core region, was observed for the NBI discharges without ECH.

Magnetic island formation is experimentally investigated in the Large Helical Device. The (m, n) = (1, 1) vacuum magnetic island is generated by using the local island diverter field, where m and n are the poloidal and toroidal mode numbers, respectively. The island width depends on plasma parameters (the electron temperature and the β) and the magnetic axis position. In the case of Rax = 3.53 m the magnetic island in the plasma is larger than that in the vacuum field. Here, Rax is the major radius of the magnetic axis. In the case of Rax = 3.6 m the magnetic island is not generated when the error field is less than the threshold which is increased as the β is increased. Evidence of island current is obtained when the magnetic island is formed due to a small error field. However, the mechanism that generates the island is not yet known.

In the Large Helical Device (LHD) programme, one of the key research issues is to enhance helical plasma performance through edge plasma control. For the first time in the LHD programme, the edge plasma control was performed with a local island divertor (LID) that is a closed divertor, utilizing an m/n = 1/1 island generated externally by which 20 small perturbation coils, and fundamental LID functions were demonstrated experimentally. It was found that the outward heat and particle fluxes crossing the island separatrix flow along the field lines to the rear of the divertor head, where carbon plates are placed to receive the heat and particle loads. Accordingly, high efficient pumping was demonstrated, which is considered to be the key in realizing high temperature divertor operations, resulting in an improvement in energy confinement. In this experiment, relatively good energy confinement is achieved in the high density regime at a magnetic axis position, Rax, of 3.75?m. Results of edge modelling are also presented by using the EMC3-EIRENE code.

Control of the radial electric field, Er, is considered to be important in helical plasmas, because the radial electric field and its shear are expected to reduce neoclassical and anomalous transport, respectively. In general, the radial electric field can be controlled by changing the collisionality, and positive or negative electric fields have been obtained by decreasing or increasing the electron density, respectively. Although the sign of the radial electric field can be controlled by changing the collisionality, modification of the magnetic field is required to achieve further control of the radial electric field, especially to produce a strong radial electric field shear. In the Large Helical Device (LHD) the radial electric field profiles are shown to be controlled by the modification of the magnetic field by (1) changing the radial profile of the effective helical ripples, ?h, (2) creating a magnetic island with an external perturbation field coil and (3) changing the local island divertor coil current.

A local island divertor (LID) experiment has begun on LHD, with the aims of controlling edge recycling and improving the plasma confinement. The fundamental divertor functions of the LID have been demonstrated in the recent experiments. From the particle flux profile measurements on the LID head it was found that the particles diffusing out from the core region are well guided along the island separatrix to the LID head. Owing to the closed configuration around the LID head, evidence of the high efficient pumping was observed, together with a strong capacity to screen impurities. The first results of edge modeling using the EMC3-EIRENE code are also presented.

Transition phenomena of the radial electric field are investigated using charge exchange spectroscopy with an improved time resolution in large helical device (LHD). The critical density, where the radial electric field changes from a positive to negative value, is observed to be higher in the outward shifted configuration because of a large ripple loss than that in the inward shifted configurations, which qualitatively agrees with the predictions of neoclassical theory.This article was due to appear in volume 45, issue 12A of Plasma Physics and Controlled Fusion. To access this special issue please follow this link www.iop.org/EJ/toc/0741-3335/45/12A.

Optimization has been carried out on the operational regime where neoclassical internal transport barriers (N-ITBs) are formed. The condition of the forming N-ITB is sensitive to the location of the electron cyclotron heating resonance zone. A high electron temperature is obtained for the central heating (−2 cm < Rresonance < 2 cm). The variation of the magnetic axis location affects the formation of the N-ITB. The electron and ion temperatures at the magnetic axis location (Rax) of 94.9 cm are higher than those at 92.1 and 97.4 cm. This result indicates that the neoclassical flux and the created radial electric field are influenced by the magnetic field structure.

We report recent progress in the study of neoclassical internal transport barriers (N-ITBs) in the compact helical system (CHS) experiment. N-ITBs have been observed in electron cyclotron (EC) heated NBI plasmas. The improved confinement region for electrons is expanded compared with that of previous N-ITB plasmas heated by only EC heating. Moreover, the ion temperature is found to be increased simultaneously by two to three times (Ti(0) ~ 400–500 eV), with a steep gradient region at ρ ~ 0.6. From measurement of the averaged peak energy intensity using a soft x-ray CCD camera, it is confirmed that the impurity confinement is also improved with N-ITBs. The radial electric field is observed to bifurcate into ion and electron root (Er ~ 15 kV m−1), and a rather large electric field shear (dEr/dr ~ −300 kV m−2) is produced in the layer between ion and electron root.

Standard diagnostics for fundamental plasma parameters and for plasma physics are routinely utilized for daily operation and physics studies in the large helical device (LHD) with high reliability. Diagnostics for steady-state plasma are under intensive development, especially for Te, ne (yttrium–aluminium garnet (YAG) laser Thomson, CO2 laser polarimeter), data acquisition in steady-state and heat-resistant probes. To clarify the plasma properties of the helical structure, two- or three-dimensional diagnostics are being aggressively developed: tangential cameras (fast SX TV, photon counting CCD, Hα TV); tomography (tangential SX CCD, bolometer); imaging (bolometer, ECE, reflectometer). Divertor and edge physics are important key issues for steady-state operation. Diagnostics for neutral flux (Hα array, Zeeman spectroscopy) and ne (fast scanning probe, Li beam probe, pulsed radar reflectometer) are also in advanced stages of development. In addition to these, advanced diagnostics are being intensively developed in LHD through domestic and international collaborations.

Significant progress has been made with ion cyclotron range of frequencies (ICRF) heating in the Large Helical Device. This is mainly due to better confinement of the helically trapped particles and less accumulation of impurities in the region of the plasma core. During the past two years, ICRF heating power has been increased from 1.35 to 2.7 MW. Various wave-mode tests were carried out using minority-ion heating, second-harmonic heating, slow-wave heating and high-density fast-wave heating at the fundamental cyclotron frequency. This fundamental heating mode extended the plasma density range of effective ICRF heating to a value of 1×1020 m−3. This use of the heating mode was its first successful application in large fusion devices. Using the minority-ion mode gave the best performance, and the stored energy reached 240 kJ using ICRF alone. This was obtained for the inward-shifted magnetic axis configuration. The improvement associated with the axis-shift was common for both bulk plasma and highly accelerated particles. For the minority-ion mode, high-energy ions up to 500 keV were observed by concentrating the heating power near the plasma axis. The confinement properties of high-energy particles were studied for different magnetic axis configurations, using the power-modulation technique. It confirmed that with the inward-shifted configuration the confinement of high-energy particles was better than with the normal configuration. By increasing the distance of the plasma to the vessel wall to about 2 cm, the impurity influx was sufficiently reduced to allow sustainment of the plasma with ICRF heating alone for more than 2 min.

Internal transport barriers with respect to electron thermal transport (eITB) were observed in the large helical device, when the electron cyclotron resonance heating (ECH) power was highly localized on the centre of a plasma sustained by neutral beam injection. The eITB is characterized by a high central electron temperature of 6–8 keV with an extremely steep gradient, as high as 55 keV m−1 and a low electron thermal diffusivity within a normalized average radius ρ≈0.3 as well as by the existence of clear thresholds for the ECH power and plasma collisionality.

The behaviour of high-energy ions accelerated by an ion cyclotron range of frequency (ICRF) electric field in the large helical device (LHD) is discussed. A better confinement performance of high-energy ions in the inward-shifted magnetic axis configuration was experimentally verified by measuring their energy spectrum and comparing it with the effective temperature determined by an electron slowing down process. In the standard magnetic axis configuration a saturation of the measured tail temperature was observed as the effective temperature was increased. The ratio between these two quantities is a measure of the quality of transfer efficiency from high-energy ions to a bulk plasma; when this efficiency was compared with Monte Carlo simulations the results agreed fairly well. The ratio of the stored energy of the high-energy ions to that of the bulk plasma was measured using an ICRF heating power modulation method; it was deduced from phase differences between total and bulk plasma stored energies and the modulated ICRF heating power. The measured high energy fraction agreed with that calculated using the injected ICRF heating power, the transfer efficiency determined in the experiment and the confinement scaling of the LHD plasma.

The experimentally optimized magnetic field configuration of the Large Helical Device [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)], where the magnetic axis is shifted inward by 15 cm from the early theoretical prediction, reveals 50% better global energy confinement than the prediction of the scaling law. This configuration has been investigated further from the viewpoints of high energy particle confinement and magnetohydrodynamic (MHD) stability. The confinement of high energy ions is improved as expected. The minority heating of ion cyclotron range of frequency was successful and the heating efficiency was improved by the inward shift. The confinement of passing particles by neutral beam injection was also improved under low magnetic field strength, and there could be obtained an almost steady high beta discharge up to 3% in volume average. This was a surprising result because the observed pressure gradient exceeded the Mercier unstable limit. The observed MHD activities became as high as beta but they did not grow enough to deteriorate the confinement of high energy ions or the performance of the bulk plasma, which was still 50% better than the scaling. According to these favorable results, better performance would be expected by increasing the heating power because the neoclassical transport can also be improved there.

Enhancement of the output power per gyrotron has been planned in the Large Helical Device (LHD). Three 77-GHz gyrotrons with an output power of more than 1 MW have been operated. In addition, a high power gyrotron with the frequency of 154 GHz (1 MW/5 s, 0.5 MW/CW) was newly installed in 2012 and the total injection power of ECRH reached 4.6 MW. The operational regime of ECRH plasma on the LHD has been extended due to the upgraded ECRH system such as the central electron temperature T{sub e0} = 13.5 keV with n{sub e} = 1×10{sup 19}m{sup −3}. In the LHD, an electron-internal-transport barrier (e-ITB) related to the production of high T{sub e} plasmas has been realized by strongly centre-focused ECRH. The electron thermal confinement clearly improved inside the e-ITB. The radial electric field was measured using the heavy ion beam probe. The formation of the positive E{sub r} was observed in the core region, which well agreed with the prediction of the neoclassical transport theory. The energy confinement characteristics have been investigated in the ECRH plasmas. It was found that higher plasma stored energy and lower radiation power was realized in the outward configuration. The plasma stored energy of 530more » kJ with n{sub e} = 3.2×10{sup 19}m{sup −3}, which is the 1.7 times larger than the previous record in the ECRH plasma in the LHD, has been successfully achieved.« less

Reversed-shear Alfven eigenmodes were observed for the first time in a helical plasma having negative q{sub 0}{sup ''} (the curvature of the safety factor q at the zero shear layer). The frequency is swept downward and upward sequentially via the time variation in the maximum of q. The eigenmodes calculated by ideal MHD theory are consistent with the experimental data. The frequency sweeping is mainly determined by the effects of energetic ions and the bulk pressure gradient. Coupling of reversed-shear Alfven eigenmodes with energetic ion driven geodesic acoustic modes generates a multitude of frequency-sweeping modes.

A spontaneous transition phenomena between two states of a plasma with an internal transport barrier (ITB) is observed in the steady-state phase of the magnetic shear in the negative magnetic shear plasma in the JT-60U tokamak. These two ITB states are characterized by different profiles of the second radial derivative of the ion temperature inside the ITB region (one has a weak concave shape and the other has a strong convex shape) and by different degrees of sharpness of the interfaces between the L mode and the ITB region, which is determined by the turbulence penetration into the ITB region.